Adoption of double-wall straight-tube steam generators (SGs) made of Mod.9Cr-1Mo steel is planned for next-generation fast breeder reactors (FBRs) in Japan. One of the major concerns with the SG is the structural integrity of the tubesheet. During a transient event, a maximum thermal stress may be induced by the temperature distribution in the tubesheet, and the magnitude of the stress depends on the configuration of the tubesheet. Therefore, the stress generation mechanism of a tubesheet was studied through finite element (FE) analysis. Semispherical tubesheet models were investigated for the first survey of the thermal stress mechanism. The calculated results of the semispherical tubesheet model indicated an extensive peak stress around the outermost hole. The recognized thermal stress mechanism of a semispherical tubesheet is as follows: (1) The dominant thermal stress is hoop stress caused by the temperature difference between the perforated and surrounding regions. (2) The thermal stress is insensitive to the size of the specific portion, although it is dominated by an interaction mechanism between the perforated and surrounding regions. (3) The stress concentration around the edge of the holes generates a peak stress. (4) The amplitude of the peak stress depends on the tubesheet penetration angle, and the stress concentration becomes greatest near the outermost hole. Based on the above stress generation mechanism, we proposed a stress-mitigated tubesheet, a center-flattened spherical tubesheet (CFST), as an improved configuration. The calculated peak stress of the CFST was smaller than that of the semispherical tubesheet. Further investigation revealed the detailed stress generation mechanism of the CFST during a thermal transient. There were, in fact, two different comparable thermal peak stress mechanisms observed for the CFST. Both the location and magnitude of the maximum peak stress depended on the steam temperature histories during the thermal transient. The radial stress caused by structural discontinuity, which was located at the outermost hole, depended on the rate (dT/dt) of the steam temperature change. The hoop stress caused by the interaction between the perforated and surrounding regions, which occurred at the first inner layer hole (with respect to the outermost layer holes) depended on the range (ΔT) of the steam temperature change.

References

References
1.
Sagayama
,
Y.
,
Okada
,
K.
, and
Nagata
,
T.
,
2009
, “
Progress on Reactor System Technology in the FaCT Project Toward the Commercialization of Fast Reactor Cycle System
,”
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR2009)
, Paper No. IAEA-CN-176-01-02.
2.
Aoto
,
K.
,
Kotake
,
S.
,
Uto
,
N.
,
Ito
,
T.
, and
Toda
,
M.
,
2009
, “
JSFR Design Study and R&D Progress in the FaCT Project
,”
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR2009)
, Paper No. IAEA-CN-176-01-07.
3.
Sakai
,
T.
,
Kotake
,
S.
,
Aoto
,
K.
,
Ito
,
T.
,
Kamishima
,
Y.
, and
Oshima
,
J.
,
2010
, “
Conceptual Design Study Toward the Demonstration Reactor of JSFR
,”
Proceedings of ICAPP’10
, Paper No. 10285.
4.
Griffin
,
D. S.
,
1985
, “
Elevated-Temperature Structural Design Evaluation Issue in LMFBR Licensing
,”
Nucl. Eng. Des.
,
90
, pp.
299
306
.10.1016/0029-5493(85)90017-2
5.
Chellapandi
P.
,
Chetal
S. C.
, and
Bhoje
S. B.
,
1991
Tube Sheet Design for PFBR Steam Generator
,”
11th International Confernce on Structural Mechanics in Reactor Technology
, Paper No. E03/2.
6.
Suresh Kumar
,
R.
,
Srinivasan
,
R.
,
Chellapandi
,
P.
, and
Chetal
,
S. C.
,
2007
, “
Theoretical Explanation for Acceptability of Steam Generator Tubesheet Shell Junction Under Manufacturing Deviation
,”
19th International Conference on Structural Mechanics in Reactor Technology
, Paper No. F04/4.
7.
Kasahara
,
N.
,
Kawasaki
,
N.
,
Wakai
,
T.
, and
Takasho
,
H.
,
2007
, “
A General Determination Method of Non-Linear Equivalent Material Properties for Perforated Plate
,”
Transaction of 19th International Conference on Structural Mechanics in Reactor Technology
, Paper No. F02/2.
8.
Webb
,
D. C.
,
Kormi
,
K.
, and
Al-Hassani
,
S. T. S.
,
1995
, “
Use of FEM in Performance Assessment of Perforated Plate Subject to General Loading Condition
,”
Int. J. Pressure Piping
,
64
, pp.
137
152
.10.1016/0308-0161(94)00078-W
9.
Kasahara
,
N.
,
Takasho
,
H.
,
Kawasaki
,
N.
, and
Ando
,
M.
,
2008
Effective Stress Ratio of Triangular Pattern Perforated Plates
,”
Proceedings of ASME PVP 2008
, Paper No. PVP2008-61458.
10.
Pratt
,
R.
, and
Pritchard
,
O. J.
, “
The Design of Dished Tubeplates Using the Equivalent Plate Models
,”
Transaction of 7th International Conf. on Structural Mechanics in Reactor Technology
, Paper No. E3/8.
11.
International code, 2007 ASME Boiler & Pressure Vessel code, 2007, Section III, Rules for Construction of Nuclear Facility Components, Division 1, Appendices, Article A-8000, Stress in perforated flat plate, American Society of Mechanical Engineers, New York.
12.
International code, 2007 ASME Boiler & Pressure Vessel code, 2007, Section III, Rules for Construction of Nuclear Facility Components, Division 1, Subsection NH, Class 1 components in elevated temperature service, American Society of Mechanical Engineers, New York.
13.
JSME S NC2-2005, 2005, Code for Nuclear Power Generation Facilities, Rules on Design and Construction for Nuclear Power Plants, Section II Fast Reactor Codes, Japan Society of Mechanical Engineers (In Japanese), Japan.
14.
Japan Atomic Energy Agency and Itochu Techno-Solutions
,
2008
, FINAS User's Manual Ver. 19.0 (in Japanese).
15.
Kurome, K., Date, S., Sukekawa, M., Takakura, K., Kawasaki, N., and Tanaka, Y.,
1999
, “
Material Strength Code of 316FR Stainless Steel and Modified 9Cr-1Mo Steel
,”
Proceedings of ASME PVP 1999
, Paper No. PVP-Vol.391, P47-54.
16.
Wakai
,
T.
,
Aoto
,
K.
,
Sukekawa
,
M.
,
Date
,
S.
, and
Shibamoto
,
H.
,
2008
, “
Present Status of Development of High Chromium Steel for Japanese FBR Components
,”
Nucl. Eng. Des.
,
238
, pp.
399
407
.10.1016/j.nucengdes.2006.09.023
17.
International code, 2007 ASME Boiler & Pressure Vessel code, 2007, Section II, Materials, Part D, Properties (Customary), American Society of Mechanical Engineers, New York.
18.
Hazama
,
O.
, and
Araya
,
F.
,
2007
, “
Thermal Stress Evaluation of Tubesheet Structures for Double-Wall-Tube Steam Generators of FBRs (2) Large-Scale Thermal Stress Analysis
,”
Proceedings of JSME Annual Meeting 2007
, Paper No. 2805 (in Japanese).
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