Posttest Analysis of the NUPEC/NRC 1:4 Scale Prestressed Concrete Containment Vessel Model (NUREG/CR-6809, SAND2003-0839P, ANA-01-0330)
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- Cover through page xxvi (Front Matter) (PDF - 374 KB)
- Chapter 1, Introduction (PDF - 6.7 MB)
- Chapter 2, Final Pretest Analysis (PDF - 3.7 MB)
- Chapter 3, Test Measurements (PDF - 414 KB)
- Chapter 4, Comparisons of Pretest Analysis Results with the Test (PDF - 24.9 MB)
- Chapter 5, Global Axisymmetric Analysis (PDF - 16.1 MB)
- Chapter 6, 3DCM Model Posttest Analysis (PDF - 10.9 MB)
- Chapter 7, Revised Local Penetration Posttest Model Analyses (PDF - 19.5 MB)
- Chapter 8, Liner Seam and "Rat-Hole" Detail Analysis
- Chapter 8.1 - Chapter 8.7, Figures 8-1 through 8-11 (PDF - 10.6 MB)
- Chapter 8.7, Figures 8-12 through 8-23 (PDF - 10.7 MB)
- Chapter 8.7, Figures 8-24 through 8-31 (PDF - 11.1 MB)
- Chapter 8.7, Figures 8-32 through 8-38 (PDF - 9.67 MB)
- Chapter 8.7, Figures 8-39 through 8-63 (PDF - 10.8 MB)
- Chapter 8.7, Figures 8-64 through 8-86 (PDF - 11.0 MB)
- Chapter 8.7, Figures 8-87 through 8-114 (PDF - 8.20 MB)
- Chapter 9, Global SFMT Posttest Analysis (PDF - 42.6 MB)
- Chapter 10, Conclusions and Lessons Learned(PDF - 109 KB)
- Chapter 11, References (PDF - 78 KB)
Manuscript Completed: March 2003
Date Published: March 2003
R.A. Dameron, B.E. Hansen, D.R. Parker, Y.R. Rashid
5435 Oberlin Drive
San Diego, California 92121
Under Sandia Contract No. AO-5464
M. Hessheimer, Sandia Project Monitor
Sandia National Laboratories
P.O. Box 5800
Albuquerque, New Mexico 87185-0744
Operated by Sandia Corporation
for the U.S. Department of Energy
J.F. Costello, NRC Project Manager
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
NRC Job Code Y6131
The Nuclear Power Engineering Corporation of Japan and the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, are cosponsoring and jointly funding a Cooperative Containment Research Program at Sandia National Laboratories (SNL) in Albuquerque, New Mexico. As a part of the program, a prestressed concrete containment vessel (PCCV) model was subjected to a series of overpressurization tests at SNL beginning in July 2000 and culminating in a functional failure mode or Limit State Test (LST) in September 2000 and a Structural Failure Mode Test (SFMT) in November 2001. The PCCV model, uniformly scaled at 1:4, represents the containment structure of an actual Pressurized Water Reactor (PWR) plant (OHI-3) in Japan. The objectives of the internal pressurization tests were to obtain measurement data on the structural response of the model to pressure loading beyond design basis accident in order to validate analytical modeling, find pressure capacity of the model, and observe its failure mechanisms.
This report compares results of pretest analytical studies of the PCCV model to the PCCV high pressure test measurements and describes results of posttest analytical studies. These analyses were performed by ANATECH Corp. under contract with SNL. The posttest analysis represents the third phase of a comprehensive PCCV analysis effort. The first phase consisted of preliminary analyses to determine what finite element models would be necessary for the pretest prediction analyses, and the second phase consisted of the pretest prediction analyses.
The principal objectives of the posttest analyses were: (1) to provide insights to improve the analytical methods for predicting the structural response and failure modes of a prestressed concrete containment, and (2) to evaluate by analysis any phenomena or failure mode observed during the test that had not been explicitly predicted by analysis. The posttest activities documented herein also include reviewing the effects of and "correcting" the test data for external factors that were not explicitly considered in the analyses, such as ambient temperature variations and artificial response data created by the instrumentation.
In addition to documenting the comparisons between measured behavior and predicted behavior of the liner, concrete, rebar, and tendons, a variety of failure modes and locations were investigated. Global analysis helped identify possible modes; other analyses investigated localized failure modes or modes specifically associated with 3D behavior. Liner tearing failure at the midheight of the cylinder near penetrations and a shear/bending failure at the base of the cylinder wall were both found to be competing failure modes. More detailed modeling of these locations placed a higher likelihood of failure on the liner tearing mode at the cylinder midheight near a major penetration. The most likely location for the liner tearing failure was near the Equipment Hatch at the ending point of a vertical T-anchor, near where the liner is attached to the thickened liner insert plate. The pressure at which the local analysis computed liner strains that reached the failure limits (indicating tearing and leakage) was 3.2 times the design pressure (Pd) of 0.39 MPa or 1.27 MPa. During the LST, liner tearing and leakage failure was first detected at a pressure of 2.4–2.5 Pd, and subsequent increase in pressure to 3.3 Pd resulted in further tearing at many strain concentration locations and increasing leakage. This report compares measured strains near as many of these strain concentrations as possible to the predictions from the global and local penetration analyses. The report also describes reanalysis of existing models and new analysis of new models, including representation of typical liner seam details aimed at simulating some local as-built conditions that existed in the test.
The LST resulted in liner tearing and leakage, but not in a structural failure. Structural damage was limited to concrete cracking and the overall structural response (displacements, rebar and tendon strains, etc.) was only slightly beyond yield. (Global hoop strains at the midheight of the cylinder only reached 0.4%, approximately twice the yield strain in steel.) In order to provide additional structural response data for comparison with in-elastic response conditions, the PCCV model was resealed, filled nearly full with water, and repressurized during the SFMT to a maximum pressure of 3.6 Pd when a catastrophic rupture occurred. A comparison of pretest and post-LST analysis results to the SFMT data and additional analyses, to provide some insight into the mechanisms leading to the structural failure, are also included in this report.
The report closes with summary and conclusions on the accuracy and adequacy of the pretest prediction analysis. The summary attempts to also draw lessons learned from previous containment research and highlight the new and unique lessons learned from the 1:4 scale PCCV project, such as the modeling and behavior of prestressing and some unique liner seam details. These conclusions are then used to establish guidelines for containment analysis. The relevance of this research to U.S. plants is also discussed.