Abstract

In the design study of an advanced sodium-cooled fast reactor (advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called fuel assembly with an inner duct structure (FAIDUS) has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the coolant temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named thermal-hydraulic analysis of asymmetrical flow in reactor elements (ASFRE) has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE had been validated by comparisons with experimental data, in the previous study. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mockup experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house computational fluid dynamics (CFD) code named single-phase thermal-hydraulic analysis in an irregular rod array layout (SPIRAL) were applied to make further discussion on applicability of ASFRE to the thermal hydraulics analysis in FAIDUS. Thermal hydraulic analyses of a typical FA and FAIDUS at high and low flowrate conditions were conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism on appearance of the specific temperature distributions between the numerical results by ASFRE and those by SPIRAL. In addition, the necessity of modification on the empirical constants in numerical model of ASFRE to improve the predictive accuracy was indicated.

References

1.
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
.10.5516/NET.2007.39.3.171
2.
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
.10.1080/18811248.2011.9711720
3.
Ohshima
,
H.
,
Narita
,
H.
, and
Ninokata
,
H.
,
1997
, “
Thermal-Hydraulic Analysis of Fast Reactor Fuel Subassembly With Porous Blockages
,”
Proceedings of the Fourth International Seminar on Subchannel Analysis (ISSCA-4)
,
Tokyo, Japan
, Sept. 25–26, pp.
323
333
.
4.
Kikuchi
,
N.
,
Imai
,
Y.
,
Yoshikawa
,
R.
,
Doda
,
N.
,
Tanaka
,
M.
, and
Ohshima
,
H.
,
2019
, “
Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor
,”
ASME J. Nucl. Eng. Radiat. Sci.
,
5
(
2
), p.
021001
.10.1115/1.4042191
5.
Ohshima
,
H.
, and
Imai
,
Y.
,
2017
, “
Numerical Simulation Method of Thermal-Hydraulics in Wire-Wrapped Fuel Pin Bundle of Sodium-Cooled Fast Reactor
,”
Proceedings of the Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17)
,
IAEA
,
Yekaterinburg, Russian Federation
, June 26–29, Paper No. IAEACN245-453, p.
9
.
6.
Meijerink
,
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
(
1
), pp.
134
155
.10.1016/0021-9991(81)90041-3
7.
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
.10.1016/0029-5493(87)90306-2
8.
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
.10.13182/NT72-A31065
9.
Waltar
,
A. E.
,
Todd
,
D. R.
, and
Tsvetkov
,
P. V.
,
2012
,
Fast Spectrum Reactors
,
Springer US
,
New York
, pp.
257
258
.
10.
Brooks
,
A.
, and
Hughes
,
T. J. R.
,
1982
, “
Streamline Upwind/Petrov-Galerkin Formulations for Convection Dominated Flows With Particular Emphasis on the Incompressible Navier-Stokes Equations
,”
Comput. Methods Appl. Mech. Eng.
,
32
(
1–3
), pp.
199
259
.10.1016/0045-7825(82)90071-8
11.
Abe
,
K.
,
Kondoh
,
T.
, and
Nagano
,
Y.
,
1994
, “
A New Turbulence Model for Predicting Fluid Flow and Heat Transfer in Separating and Reattaching flows - I. Flow Field Calculations
,”
Int. J. Heat Mass Transfer
,
37
(
1
), pp.
139
151
.10.1016/0017-9310(94)90168-6
12.
Ohshima
,
H.
, and
Imai
,
Y.
,
2005
, “
Validation Study of Thermal-Hydraulic Analysis Program “SPIRAL” for Fuel Pin Bundle of Sodium-Cooled Fast Reactor
,”
Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-11)
,
Avignon, France
, Oct. 2–6, Paper No. 423, p.
14
.
13.
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
.10.1016/j.nucengdes.2015.04.007
You do not currently have access to this content.