The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970's using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The United States Nuclear Regulatory Commission (US NRC) developed leak-before-break (LBB) procedures in standard review plan (SRP) 3.6.3 Revision 1 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987. The US NRC is currently developing a draft Regulatory Guide on what is called the transition break size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work (Wilkowski et al., “Robust LBB Analysis for Atucha II Nuclear Plant,” 2011 ASME PVP Conference, July 17–21, Baltimore, MD) on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1 × 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94% of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33% around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1 × 10−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

References

References
1.
Olson
,
R.
,
Scott
,
P.
, and
Wilkowski
,
G.
,
1992
, “
Application of a Nonlinear-Spring Element to Analysis of Circumferentially Cracked Pipe Under Dynamic Loading
,”
ASME Pressure Vessels and Piping Conference
, June 21–25,
New Orleans, LA
.
2.
Olson
,
R.
,
Wolterman
,
R.
,
Scott
,
P.
,
Krishnaswamy
,
P.
, and
Wilkowski
,
G.
,
1994
, “
The Next Generation Analysis Methodology for Cracked Pipe Systems Subjected to Dynamic Loads
,”
ASME Pressure Vessels and Piping Conference
, June 19–23,
Minneapolis, MN
.
3.
ABAQUS 6.10-1, Dassault Systemes Simulia Corp, 2010.
4.
Zhang
,
T.
,
Brust
,
F. W.
,
Shim
,
D. J.
,
Wilkowski
,
G.
,
Nie
,
J.
, and
Hofmayer
,
C.
,
2010
, “
Analysis of JNES Seismic Tests on Degraded Piping
,” NRC NUREG/CR-7015 (BNL-NUREG-91346-2010) Report, July.
5.
Zhang
,
T.
,
Brust
,
F. W.
,
Wilkowski
,
G.
,
Shim
,
D. J.
,
Nie
,
J.
,
Hofmayer
,
C.
, and
Ali
,
S.
,
2012
, “
Numerical Analysis of JNES Seismic Tests on Degraded Combined Piping System
,”
ASME J. Pressure Vessel Technol.
,
134
, p.
011801
.10.1115/1.4005055
6.
Suzuki
,
K.
,
Kawauchi
,
H.
, and
Abe
,
H.
,
2006
, “
Test Programs For Degraded Core Shroud and BWR Piping (Simulated Crack Models and Input Seismic Waves For Shaking Test)
,”
ASME Pressure Vessels and Piping Conference
, July 23–27, Vancouver, BC, Canada.
7.
Suzuki
,
K.
, and
Kawauchi
,
H.
,
2008
, “
Test Programs for Degraded Core Shroud and PLR System Piping (Seismic Test Results and Discussion on JSME Rules Application)
,”
2008 ASME PVP Conference
, July 27–31,
Chicago, IL
.
8.
Wilkowski
,
G.
,
Brust
,
F.
,
Zhang
,
T.
,
Hattery
,
G.
,
Kalyanam
,
S.
,
Shim
,
D.
,
Kurth
,
E.
,
Hioe
,
H.
,
Uddin
,
M.
,
Johnson
,
J.
,
Maslenikov
,
O.
,
Gurpinar
,
A.
,
Asfura
,
A.
,
Sumodobila
,
B.
,
Betervide
,
A.
, and
Mazzantini
,
O.
,
2011
, “
Robust LBB Analysis for Atucha II Nuclear Plant
,”
2011 ASME PVP Conference
, July 17–21,
Baltimore, MD
.
9.
Zhang
,
T.
,
Brust
,
F.
, and
Wilkowski
,
G.
,
2011
, “
Weld Residual Stress in Large Diameter Nuclear Nozzles
,”
2011 ASME Pressure Vessels and Piping Division Conference
, July 17–21,
Baltimore, MD
.
10.
Rudland
,
D.
,
Zhang
,
T.
,
Wilkowski
,
G.
, and
Csontos
,
A.
,
2008
, “
Welding Residual Stress Solutions for Dissimilar Metal Surge Line Nozzles Welds
,”
2008 ASME Pressure Vessels and Piping Division Conference
, July 27–31,
Chicago, IL
.
11.
Rudland
,
D.
,
Chen
,
Y.
,
Zhang
,
T.
,
Wilkowski
,
G.
,
Broussard
,
J.
, and
White
,
G.
,
2007
, “
Comparison of Welding Residual Stress Solutions for Control Rod Drive Mechanism Nozzles
,”
2007 ASME Pressure Vessels and Piping Division Conference
, July 22–26,
San Antonio, TX
.
12.
Zhang
,
T.
,
Brust
,
F.
,
Wilkowski
,
G.
,
Rudland
,
D.
, and
Csontos
,
A.
,
2009
, “
Welding Residual Stress and Multiple Flaw Evaluation for Reactor Pressure Vessel Head Replacement Welds With Alloy 52
,”
2009 ASME Pressure Vessels and Piping Division Conference
, July 26–30,
Prague, Czech Republic
.
13.
Ansys 11.0, Ansys, Inc.,
2007
.
You do not currently have access to this content.