Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix

Atsushi Ui, Yoshiro Kudo, Masahiro Furuya

Research output: Contribution to conferencePaper

Abstract

In order to gain the reliability of subchannel analysis on three-fluid two-phase flow in nuclear fuel assemblies, implemented models are expected to describe the detailed three-dimensional two-phase flow in fuel assemblies. Especially, crossflow model limits the prediction performance of subchannel analysis codes, it is important to develop a model that can analyze the phenomenon appropriately. In this study, CTF results are validated against the NUPEC BWR Full-Size Fine-mesh Bundle Test (BFBT). In BFBT, void fraction distribution across 8×8 rod bundles was measured to confirm the effects of radial/axial power distribution and unheated rods. Moreover, uncertainties of void fraction were quantified. In order to evaluate the prediction performance of the CTF code for the BFBT, bundle-averaged void fraction difference and the residual void fraction difference for each subchannel were defined. Subchannels were classified into several groups considering the grid spacer pressure loss coefficients set for each channel and the characteristics of the subchannel considering location, such as corner subchannel, adjacent corner subchannel, etc. Sensitivity parameters affecting void fraction and/or cross flow were selected with the Kriging method, and the response surface model represented by these sensitivity parameters was created. the simulation-driven MCDA method using the alternative model was applied for optimizing sensitivity parameters by data assimilation, and a set of parameters to accurately calculate the bundle-averaged void fraction difference was identified with the Metropolis method. CTF analysis with the parameter set identified by the data assimilation was conducted, and it was confirmed that the average value of the bundle-averaged void fraction improved with the parameter set by the data assimilation so that the predicted value would match the experimental value.

Original languageEnglish
Pages5052-5063
Number of pages12
Publication statusPublished - 2019 Jan 1
Event18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, United States
Duration: 2019 Aug 182019 Aug 23

Conference

Conference18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019
CountryUnited States
CityPortland
Period19/8/1819/8/23

Fingerprint

Void fraction
assimilation
bundles
voids
mesh
matrices
performance prediction
two phase flow
Two phase flow
assemblies
sensitivity
rods
kriging
cross flow
nuclear fuels
Nuclear fuels
spacers
grids
Fluids
fluids

Keywords

  • CTF
  • Data assimilation
  • Subchannel analysis
  • Void fraction

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Instrumentation

Cite this

Ui, A., Kudo, Y., & Furuya, M. (2019). Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix. 5052-5063. Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.

Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix. / Ui, Atsushi; Kudo, Yoshiro; Furuya, Masahiro.

2019. 5052-5063 Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.

Research output: Contribution to conferencePaper

Ui, A, Kudo, Y & Furuya, M 2019, 'Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix', Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States, 19/8/18 - 19/8/23 pp. 5052-5063.
Ui A, Kudo Y, Furuya M. Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix. 2019. Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.
Ui, Atsushi ; Kudo, Yoshiro ; Furuya, Masahiro. / Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix. Paper presented at 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, Portland, United States.12 p.
@conference{fc8a5eddcfa44b7680b19dbf87133333,
title = "Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix",
abstract = "In order to gain the reliability of subchannel analysis on three-fluid two-phase flow in nuclear fuel assemblies, implemented models are expected to describe the detailed three-dimensional two-phase flow in fuel assemblies. Especially, crossflow model limits the prediction performance of subchannel analysis codes, it is important to develop a model that can analyze the phenomenon appropriately. In this study, CTF results are validated against the NUPEC BWR Full-Size Fine-mesh Bundle Test (BFBT). In BFBT, void fraction distribution across 8×8 rod bundles was measured to confirm the effects of radial/axial power distribution and unheated rods. Moreover, uncertainties of void fraction were quantified. In order to evaluate the prediction performance of the CTF code for the BFBT, bundle-averaged void fraction difference and the residual void fraction difference for each subchannel were defined. Subchannels were classified into several groups considering the grid spacer pressure loss coefficients set for each channel and the characteristics of the subchannel considering location, such as corner subchannel, adjacent corner subchannel, etc. Sensitivity parameters affecting void fraction and/or cross flow were selected with the Kriging method, and the response surface model represented by these sensitivity parameters was created. the simulation-driven MCDA method using the alternative model was applied for optimizing sensitivity parameters by data assimilation, and a set of parameters to accurately calculate the bundle-averaged void fraction difference was identified with the Metropolis method. CTF analysis with the parameter set identified by the data assimilation was conducted, and it was confirmed that the average value of the bundle-averaged void fraction improved with the parameter set by the data assimilation so that the predicted value would match the experimental value.",
keywords = "CTF, Data assimilation, Subchannel analysis, Void fraction",
author = "Atsushi Ui and Yoshiro Kudo and Masahiro Furuya",
year = "2019",
month = "1",
day = "1",
language = "English",
pages = "5052--5063",
note = "18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 ; Conference date: 18-08-2019 Through 23-08-2019",

}

TY - CONF

T1 - Data assimilation with subchannel analysis code CTF on NUPEC BWR BFBT test matrix

AU - Ui, Atsushi

AU - Kudo, Yoshiro

AU - Furuya, Masahiro

PY - 2019/1/1

Y1 - 2019/1/1

N2 - In order to gain the reliability of subchannel analysis on three-fluid two-phase flow in nuclear fuel assemblies, implemented models are expected to describe the detailed three-dimensional two-phase flow in fuel assemblies. Especially, crossflow model limits the prediction performance of subchannel analysis codes, it is important to develop a model that can analyze the phenomenon appropriately. In this study, CTF results are validated against the NUPEC BWR Full-Size Fine-mesh Bundle Test (BFBT). In BFBT, void fraction distribution across 8×8 rod bundles was measured to confirm the effects of radial/axial power distribution and unheated rods. Moreover, uncertainties of void fraction were quantified. In order to evaluate the prediction performance of the CTF code for the BFBT, bundle-averaged void fraction difference and the residual void fraction difference for each subchannel were defined. Subchannels were classified into several groups considering the grid spacer pressure loss coefficients set for each channel and the characteristics of the subchannel considering location, such as corner subchannel, adjacent corner subchannel, etc. Sensitivity parameters affecting void fraction and/or cross flow were selected with the Kriging method, and the response surface model represented by these sensitivity parameters was created. the simulation-driven MCDA method using the alternative model was applied for optimizing sensitivity parameters by data assimilation, and a set of parameters to accurately calculate the bundle-averaged void fraction difference was identified with the Metropolis method. CTF analysis with the parameter set identified by the data assimilation was conducted, and it was confirmed that the average value of the bundle-averaged void fraction improved with the parameter set by the data assimilation so that the predicted value would match the experimental value.

AB - In order to gain the reliability of subchannel analysis on three-fluid two-phase flow in nuclear fuel assemblies, implemented models are expected to describe the detailed three-dimensional two-phase flow in fuel assemblies. Especially, crossflow model limits the prediction performance of subchannel analysis codes, it is important to develop a model that can analyze the phenomenon appropriately. In this study, CTF results are validated against the NUPEC BWR Full-Size Fine-mesh Bundle Test (BFBT). In BFBT, void fraction distribution across 8×8 rod bundles was measured to confirm the effects of radial/axial power distribution and unheated rods. Moreover, uncertainties of void fraction were quantified. In order to evaluate the prediction performance of the CTF code for the BFBT, bundle-averaged void fraction difference and the residual void fraction difference for each subchannel were defined. Subchannels were classified into several groups considering the grid spacer pressure loss coefficients set for each channel and the characteristics of the subchannel considering location, such as corner subchannel, adjacent corner subchannel, etc. Sensitivity parameters affecting void fraction and/or cross flow were selected with the Kriging method, and the response surface model represented by these sensitivity parameters was created. the simulation-driven MCDA method using the alternative model was applied for optimizing sensitivity parameters by data assimilation, and a set of parameters to accurately calculate the bundle-averaged void fraction difference was identified with the Metropolis method. CTF analysis with the parameter set identified by the data assimilation was conducted, and it was confirmed that the average value of the bundle-averaged void fraction improved with the parameter set by the data assimilation so that the predicted value would match the experimental value.

KW - CTF

KW - Data assimilation

KW - Subchannel analysis

KW - Void fraction

UR - http://www.scopus.com/inward/record.url?scp=85073717378&partnerID=8YFLogxK

UR - http://www.scopus.com/inward/citedby.url?scp=85073717378&partnerID=8YFLogxK

M3 - Paper

AN - SCOPUS:85073717378

SP - 5052

EP - 5063

ER -