0
Research Papers

Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

[+] Author and Article Information
Norihiro Kikuchi

Japan Atomic Energy Agency,
4002 Narita-cho,
Oarai 311-1393, Ibaraki, Japan
e-mail: kikuchi.norihiro@jaea.go.jp

Yasutomo Imai

NDD Corporation,
1-1-6 jonan,
Mito 310-0803, Ibaraki, Japan
e-mail: y.imai@nddhq.co.jp

Ryuji Yoshikawa

Japan Atomic Energy Agency,
4002 Narita-cho,
Oarai 311-1393, Ibaraki, Japan
e-mail: yoshikawa.ryuji@jaea.go.jp

Norihiro Doda

Japan Atomic Energy Agency,
4002 Narita-cho,
Oarai 311-1393, Ibaraki, Japan
e-mail: doda.norihiro@jaea.go.jp

Masaaki Tanaka

Mem. ASME
Japan Atomic Energy Agency,
4002 Narita-cho,
Oarai 311-1393, Ibaraki, Japan
e-mail: tanaka.masaaki@jaea.go.jp

Hiroyuki Ohshima

Japan Atomic Energy Agency,
4002 Narita-cho,
Oarai 311-1393, Ibaraki, Japan
e-mail: ohshima.hiroyuki@jaea.go.jp

Manuscript received October 31, 2017; final manuscript received November 27, 2018; published online March 15, 2019. Assoc. Editor: Yanping Huang.

ASME J of Nuclear Rad Sci 5(2), 021001 (Mar 15, 2019) (12 pages) Paper No: NERS-17-1280; doi: 10.1115/1.4042191 History: Received October 31, 2017; Revised November 27, 2018

In the design study of advanced loop-type sodium-cooled fast reactor in Japan, a specific fuel assembly (FA) called FA with inner duct structure (FAIDUS) is expected to enhance reactor safety during a core-disruptive accident. Evaluating the thermal-hydraulics in FAIDUS under various operating conditions is required for its design. This study is the first step toward confirming the design feasibility of FAIDUS; the thermal-hydraulics in FAIDUS are investigated with an in-house subchannel analysis code called asymmetrical flow in reactor elements (ASFRE), which can be applied to a wire-wrapped fuel pin bundle in conjunction with the distributed resistance model (DRM) and the turbulence-mixing model of the Todreas–Turi correlation model (TTM). Before simulating the thermal-hydraulics in FAIDUS, a few validations of DRM and TTM are conducted by comparing the numerical results of the pressure drop coefficients or temperature distribution obtained using ASFRE with the experimental data obtained using an apparatus with water or sodium for simulated FAs. After these validations, thermal-hydraulic analyses of FAIDUS and a typical FA are conducted for comparison. The numerical results indicate that, at a high flow rate simulating rated operation condition, no significant asymmetric temperature and velocity distribution occur in FAIDUS compared to the distribution in the typical FA. In addition, at a low flow rate simulating natural circulation condition in decay heat removal, the temperature distribution in FAIDUS is similar to that in the typical FA. This is because the local flow acceleration and the flow redistribution due to buoyancy force may occur in FAIDUS and the typical FA.

Copyright © 2019 by ASME
Your Session has timed out. Please sign back in to continue.

References

Ichimiya, M. , Mizuno, T. , and Kotake, S. , 2007, “ A Next Generation Sodium-Cooled Fast Reactor Concept and Its R&D Program,” Nucl. Eng. Technol., 39(3), pp. 171–186. [CrossRef]
Aoto, K. , Uto, N. , Sakamoto, Y. , Ito, Y. , Toda, M. , and Kotake, S. , 2011, “ Design Study and R&D Progress on Japan Sodium-Cooled Fast Reactor,” J. Nucl. Sci. Technol., 48(4), pp. 463–471. [CrossRef]
Okano, Y. , Ohshima, H. , and Okubo, T. , 2011, “ Sub-Channel Analysis of Innovative Fuel Assembly Concept: FAIDUS for Sodium-Cooled Fast Reactor,” Trans. Am. Nucl. Soc., 104(1), pp. 1049–1051. http://www.ans.org/pubs/transactions/https://jopss.jaea.go.jp/search/servlet/search?5029029&language=1
Ohshima, H. , Narita, H. , and Ninokata, H. , 1997, “ Thermal-Hydraulic Analysis of Fast Reactor Fuel Subassembly With Porous Blockages,” Fourth International Seminar on Subchannel Analysis (ISSCA-4), Tokyo, Japan, Sept. 25–26, pp. 323–333.
Ninokata, H. , Efthimiadis, A. , and Todreas, N. E. , 1987, “ Distributed Resistance Modeling of Wire-Wrapped Rod Bundles,” Nucl. Eng. Des., 104(1), pp. 93–102. [CrossRef]
Todreas, N. E. , and Turi, J. A. , 1972, “ Interchannel Mixing in Wire Wrapped Liquid Metal Fast Reactor Fuel Assemblies,” Nucl. Technol., 13(1), pp. 36–52. [CrossRef]
Liles, D. R. , and Reed, W. H. , 1978, “ A Semi-Implicit Method for Two-Phase Fluid Dynamics,” J. Comput. Phys., 26(3), pp. 390–407. [CrossRef]
Meiherink, J. A. , and Van Der Vorst, H. A. , 1981, “ Guidelines for the Usage of Incomplete Decompositions in Solving Sets of Linear Equations as They Occur in Practical Problems,” J. Comput. Phys., 44, pp. 134–155. [CrossRef]
Rehme, K. , 1973, “ Simple Method of Prediction Friction Factors of Turbulent Flow in Non-Circular Channels,” Int. J. Heat Mass Transfer, 16(5), pp. 933–950. [CrossRef]
Gunter, A. Y. , and Shaw, W. A. , 1945, “ A General Correlation of Friction Factors of Various Types of Surfaces in Crossflow,” Trans. ASME, 67(8), pp. 643–660.
Narita, H. , and Ohshima, H. , 1996, “ Improvement of Single-Phase Subchannel Analysis Code ASFRE-III—Modification of Fuel Pin Heat Conduction Model,” Japan Atomic Energy Agency, Oarai, Japan, No. PNC-TN9410 96-116 (in Japanese).
Tang, Y. S. , Coffield , R. D., Jr. , and Markley, R. A. , 1978, Thermal Analysis of Liquid-Metal Fast Breeder Reactors, American Nuclear Society, Washington, DC, pp. 198–200.
Satoh, K. , Kogawa, T. , Miyaguchi, K. , and Iguchi, T. , 1981, “ ‘JOYO’ MK-II Fuel Assembly Flow Test (V)—Flow Resistance Characteristics of the Fourth Mock-Up Core Fuel Assembly,” Japan Atomic Energy Agency, Oarai, Japan, No. PNC-SN941 81-62 (in Japanese).
Sato, K. , Kogawa, T. , Miyaguchi, K. , and Iguchi, T. , 1980, “ Prototype LMFBR ‘MONJU’ Fuel Assembly Hydraulic Simulation Test (VII) Flow Resistance of the Fifth Mock-Up Fuel Assembly,” Japan Atomic Energy Agency, Oarai, Japan, No. PNC-SN941 80-24 (in Japanese).
Cheng, S. K. , and Todreas, N. E. , 1986, “ Hydrodynamic Models and Correlations for Bare and Wire-Wrapped Hexagonal Rod Bundles—Bundle Friction Factors, Subchannel Friction Factors and Mixing Parameters,” Nucl. Eng. Des., 92(2), pp. 227–251. [CrossRef]
Chen, S. K. , Todreas, N. E. , and Nguyen, N. T. , 2014, “ Evaluation of Existing Correlations for the Prediction of Pressure Drop in Wire-Wrapped Hexagonal Array Pin Bundles,” Nucl. Eng. Des., 267, pp. 109–131. [CrossRef]
Kabir, M. E. , and Hayafune, H. , 1992, “ Study of Thermo-Hydraulic Behavior Within the Fuel Bundle Under a Loss of Flow Condition,” Japan Atomic Energy Agency, Oarai, Japan, No. PNC-TN9410 92-018.
Kamide, H. , Hayashi, K. , and Toda, S. , 1998, “ An Experimental Study of Inter-Subassembly Heat Transfer During Natural Circulation Decay Heat Removal in Fast Breeder Reactors,” Nucl. Eng. Des., 183(1–2), pp. 97–106. [CrossRef]
Otaka, M. , Ohshima, H. , Ninokata, H. , and Narita, H. , 1996, “ Validation of Single-Phase Subchannel Analysis Code ASFRE-III,” Japan Atomic Energy Agency, Oarai, Japan, No. PNC-TN9410 96-212 (in Japanese).
Murakami, T. , Eguchi, Y. , Oyama, K. , and Watanabe, O. , 2015, “ Reduced-Scale Water Test of Natural Circulation for Decay Heat Removal in Loop-Type Sodium-Cooled Fast Reactor,” Nucl. Eng. Des., 288, pp. 220–231. [CrossRef]
Rolfo, S. , Péniguel, C. , Guillaud, M. , and Laurence, D. , 2012, “ Thermal-Hydraulic Study of a Wire-Spacer Fuel Assembly,” Nucl. Eng. Des., 243, pp. 251–262. [CrossRef]
Pointer, W. D. , Thomas, J. , Fanning, T. , Fischer, P. , Siegel, A. , Smith, J. , and Tokuhiro, A. , 2009, “ RANS-Based CFD Simulations of Sodium Fast Reactor Wire-Wrapped Pin Bundles,” International Conference on Mathematics, Computational Methods and Reactor Physics, New York, May 5–7, p. 13.
Gajapathy, R. , Velusamy, K. , Selvaraj, P. , and Chellapandi, P. , 2015, “ CFD Investigation of Effect of Helical Wire-Wrapped Parameters on the Thermal Hydraulic Performance of 217 Fuel Pin Bundle,” Ann. Nucl. Energy, 77, pp. 498–513. [CrossRef]
Ohshima, H. , and Imai, Y. , 2017, “ Numerical Simulation Method of Thermal-Hydraulics in Wire-Wrapped Fuel Pin Bundle of Sodium-Cooled Fast Reactor,” FR17, Yekaterinburg, Russian Federation, June 26–29, No. IAEA-CN245-453.

Figures

Grahic Jump Location
Fig. 1

Illustration of FAIDUS

Grahic Jump Location
Fig. 2

Subchannel mesh arrangement (ex., 127-pin bundle analysis)

Grahic Jump Location
Fig. 3

Control volumes for ASFRE: (a) mass and energy equation and (b) momentum equation

Grahic Jump Location
Fig. 4

Pressure loss coefficient of 127-pin bundle (experimental data are referred from Ref. [13])

Grahic Jump Location
Fig. 5

Pressure loss coefficient of 169-pin bundle (experimental data are referred from Ref. [14])

Grahic Jump Location
Fig. 6

Overview of PLANDTL facility: (a) flow diagram of experimental setup and (b) horizontal cross section of simulated FA

Grahic Jump Location
Fig. 7

Sodium temperature distributions in 37-pin bundle on middle cross section of heated region along (a) line M1 and (b) line M2, as shown in Fig. 6(b) (experimental data are referred from Ref. [19])

Grahic Jump Location
Fig. 8

Sodium temperature distributions in 37-pin bundle on top cross section of heated region along (a) line T1 and (b) line T2, as shown in Fig. 6(b) (experimental data are referred from Ref. [19])

Grahic Jump Location
Fig. 9

Overview of CCTL-CFR facility: (a) flow diagram of experimental setup and (b) horizontal cross section of simulated core

Grahic Jump Location
Fig. 10

Temperature distribution of 61-pin bundle on middle cross section of heated region along line A, as shown in Fig.9(b) (experimental data are referred from Ref. [19])

Grahic Jump Location
Fig. 11

Temperature distribution of 61-pin bundle on top cross section of heated region along (a) line A and (b) line B, as shown in Fig. 9(b) (experimental data are referred from Ref.[19])

Grahic Jump Location
Fig. 12

Traverse lines in (a) 271-pin bundle FA, (b) 255-pin bundle FAIDUS, and (c) axial region of models

Grahic Jump Location
Fig. 13

Temperature distribution in 271-pin bundle FA on top cross section of heated region along (a) line A and (b) line B, as shown in Fig. 12(a)

Grahic Jump Location
Fig. 14

Velocity distribution of 271-pin bundle FA on top cross section of heated region along (a) line A and (b) line B, as shown in Fig. 12(a)

Grahic Jump Location
Fig. 15

Temperature distribution of 255-pin bundle FAIDUS on top cross section of heated region along (a) line A and (b) line B, as shown in Fig. 12(b)

Grahic Jump Location
Fig. 16

Velocity distribution of 255-pin bundle FAIDUS on top cross section of heated region along (a) line A and (b) line B, as shown in Fig. 12(b)

Grahic Jump Location
Fig. 17

Temperature comparison on top and middle cross sections of heated region along line A, as shown in Figs. 12(a) and 12(b): (a) under high flow rate condition and (b) under low flow rate condition

Grahic Jump Location
Fig. 18

Velocity comparison on top and middle cross sections of heated region, along line A, as shown in Figs. 12(a) and 12(b): (a) under high flow rate condition and (b) under low flow rate condition

Grahic Jump Location
Fig. 19

Temperature comparison on top cross section of heated region along peripheral line

Tables

Errata

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In