ML19309F013
| ML19309F013 | |
| Person / Time | |
|---|---|
| Site: | Vallecitos File:GEH Hitachi icon.png |
| Issue date: | 04/23/1980 |
| From: | Gilliland D GENERAL ELECTRIC CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8004280311 | |
| Download: ML19309F013 (12) | |
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G E N E R A L () E LE CT R I C
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ENGINEERING GENERAL ELECTRIC COMPANY, P.O. BOX 460, PLEAsANTON,. CALIFORNIA 94566 DIVISION April 23,1980 Mr. Darrell G. Eisenhut, Director Division of Project Management Office of Nuclear Reactor Regulations U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Eisenhut:
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Attached is a partial response (Part I) to structural questions raised by the Nuclear Regulatory Commission. The re:caining responses (Part II) will be sent as soon as now-in-progress analysis activities are complete.
Very truly yours, TFFICIEL S"AL
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AFFIRMATION The General Electric Company hereby submits the attached Response to NRC Questions - Structural Issues - General Electric Test Reactor, Docket 50-73.
To the best of my knowledge and belief, the information contained herein is accurate.
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Submitted and sworn before me this twenty-fourth day of April,1980, o f,,,,, 8/fe<<ym>
, Notary Public in and for the County of Alameda,
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State of California.
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PART I RESPONSE TO NRC QUESTIONS STRUCTURAL ISSUES GENERAL ELECTRIC TEST REACTOR 1
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Request No. 1 You state that containment shell stressas exceed critical buckling stresses on the first floor level. Specify the extent of local as well as overall failure.
Response to Request No. 1 The description, location, and extent of buckling stresses in the containment shell due to vibratory ground motion and surface rupture offset are discussed in submittals to the NRC in response to requests for information (Ref.1 and Ref. 2).
It is stated in Reference 1 that containment integrity is not necessary to assure public health and safety during and following a maximum postu-lated seismic event. Because of the complex nature of buckling stresses and deformations and the minimal importance of the containment shell, no attempt has been made to investigate whether failure of the shell will occur due to buckling.
Consequently, it is assumed that the integrity of the shell will not be maintained if a maximum seismic event occurs.
REFERENCES 1.
General Electric Company, " Response to NRC Request for Additional Information on the Phase 2 Report," submitted to NRC July 26, 1978.
2.
General Electric Company, " Response to NRC Request for Information Based on GETR Site Visit Held 18 June 1979," submitted to NRC July 9,1979.
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Question 7 For surface offset cases 2 and 3 (Figure 3-2 of the Phase 2 Report),
quantify the maximum possible cantilever and unsupported lengths.
Provide bases justifying these lengths, and verify that they were used in the offset analyses.
Response to Question 7 The maximum unsupported length (Case 2) was found to be 20 feet. This length was determined by distributing the weight of the reactor building on an assumed minimum size foundation area. The weight of the building divided by the maximum bearing pressure of 20 ksf determined the size of the support area, which was distributed symmetrically on each side of a line through the center of gravity parallel to the hypothetical fault orientation. The shortest distance from the edge of the assumed stress " block" to the outside face of the exterior perimeter wall was found to be 20 feet. The offset analysis was performed using this length as it was more conservative than Case 3.
In the offset analysis, the support springs from the finite element model were located inside the foundation bearing area, which insured that the unsupported cantilever length was conservatively simulated in the finite element model.
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Currently analyses are being conducted based on realistic values of surface rupture offset combined with vibratory ground motion. The values were selected in a rational manner, noting that it is unrealistic to combine a 20' unsupported cantilever length with any significant vibratory ground motion.
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Request No. 8 A shear strength of 6 / f' representing the first crack threshold strength c
indicates the onset of major cracking and failure. Justify why some allowable strength less than this should not be used to provide an acceptable level of safety.
Response to Request No. 8 The basis for selecting the allowable shear strength of 6 ] is documented in Reference 1.
The data which supports the use of this value are given in submittals to the NRC in response to requests for information (Ref. 2 and Ref. 3.
As discussed in the above-cited references, 6 /72-- is an appropriate and conservative threshold value for the initiation of cracking of the reactor building walls for the following reasons:
First, calculated principal stresses, rather than average stresses which are lower, were compared to the criterion value of 6 / f'. This criterion value was the principal tensile stress found at the initiation of cracking for the tests reported in Reference 4.
Thus, consistent calculated and criterion stresses were compared in the GETR analysis. This approach is different from analysis procedures used for conventional engineering structures where average stresses are normally compared to allowable values which are also based on average stresses.
Second, the authors of Reference 4 (who suggested the cracking threshold value 6 /Tr-), suggest a new equation for the 1971 ACI Building Code.
Based on that squation 6.75 T would be used as an average stress criterion for walls with a height to width ratio of 0.5, which corresponds to the GETR reactor building walls. This value reflects additional capacity above the threshold of cracking for unreinforced wall and exceeds the conservative criterion used for
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the GETR analysis.
Third, the selected shear capacity of 6 /TI-is based on no steel reinforcement.
Even though on a percentage basis the amount of reinforcement in the GETR reactor building walls is small, the presence of the many number 9 and 10 bars will provide additional capacity above the cracking threshold.
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i REFERENCES 1.
Engineering-Decison Analysis Company, Inc., " Seismic Analysis of Reactor Building, General Electric Test Reactor, Phase 1," prepared for General Electric Company, San Jose, California, EDAC-117-217.02, 3 February 1978.
2.
. General Electric Company, " Response to NRC Request for Additional Information on the Phase 2 Report," submitted to NRC July 26, 1978.
3.
General Electric Company, " Response to NRC Request for Additional Information," dated July 9,1979, submitted to NRC September 5,1979.
4.
Felix Barda, John M. Hanson, and W. Gene Corley, " Shear Strength of Low-Rise Walls with Boundary Elements," Portland Cement Association, Research and Development Bulletin RD043 010, preprinted with permission from ACI Symposium, " Reinforced Concrete Structures in Seismic Zones," American Concrete Institute, Detroit, Michigan, 1976.
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Question No. 9 Vertical excitation should be included in the sliding and uplift analyses.
In addition, sliding between the basement slab and foundation mat must be accounted for in the reactor building model.
Response to Question No. 9 Vertical excitation was included in the linear elastic dynamic analyses and the resulting stress analyses. For the nonlinear sliding and uplift analyses vertical excitation was not included. These analyses were conservative however, for the following reasons. The analyses did not include the resisting effects of the 20 ft. embedment. As shown in Table 2-7 of Reference 1, if the embedment resistance to sliding were included, the building would not slide regardless of the inclusion of vertical excitation.
Similarly, if the resisting effects of embedment were incluted in the overturning (rocking) analysis, the vertical displacements would be smaller due to the frictional restraining force between the reactor building and the surrounding soil. Thus the effect of the vertical excitation would be largely compensated for by the resisting effect of embedment.
It has been shown in Reference 2 that the inclusion of vertical excitation effects for nonlinear sliding analyses can either reduce or increase the resulting displacements. However, it is also shown that in either case, the change in the displacements is very minor due to the inclusion of vertical excitation for various acceleration levels and coefficients of friction.
As shown in Reference 1, the effect of including nonlinearities due to sliding and uplift would be to significantly reduce the stresses in the concrete structure.
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There are significant safety margins available against sliding and instability due to potential uplift.
For uplift, a maximum angle of rotation of 0.09 degrees was obtained from the Nonlinear Analysis A (Table 2-6, Reference 1) which is roughly 1/300 of the theoretical angle at which the building could "tip over".
(It should be noted that the concept of tipping is one which is based on static design procedures and is overly conservative when applied to dynamic oscillatory loadings.) Similarly, as shown in Table 2-7 of Reference 1, the force required to slide the building on the foundation mat is more than twice the induced seismic force obtained from the conservative linear elastic analysis.
The building would therefore be stable even in the unlikely event that the nonlinear effects were higher due to vertical excitation.
Furthermore, there is sufficient flexibility in the Fuel Flooding System (FFS) lines (in the fonn of several feet of slack) to accommodate large displacements.
It was found on the basis of preliminary computations (Table 2-9, Reference 1) that there will not be any sliding at the interface between the basement slab and the foundation mat.
It was therefore judged not necessary to perform detailed nonlinear analyses to take such sliding into consideration.
Even if there were sliding at this interface, the effect would be to reduce the forces in the concrete structure similar to the hypothetical sliding at the interface of foundation mat and soil which resulted in the reduction of forces by about 20 to 30 percent (Table 2-8, Page 2-11 of Reference 1).
The displacements associated with such sliding would also not have any effect on the stability of the structure or the integrity of the Fuel Flooding System (FFS).
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therefore conservative to ignore the effects of such sliding at the interface i
of the basement slab and the foundation mat even if there were a possibility of such sliding.
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Thus, consideration of the constraining effects of embedment, reduction of response due to nonlinear effects, adequacy of overall building stability, sufficient flexibility of Fuel Flooding System lines, and the results of other studies such as those described in Reference 2, demonstrate that the results of the linear and nonlinear analyses performed for horizontal excitation are conservative.
References 1.
Engineering Decision Analysis Company, Inc., " Seismic Analysis of Reactor Building, General Electric Test Reactor - Phase 2,"
EDAC-117-217.03, prepared for General Electric Company,1 June 1978.
2.
Aslam, M., Godden, W.G., and Scalise, D.
T., " Sliding Response of Rigid Bodies to Earthquake Motions," Report No. LBL-3868, UC-11, under Contract TID-4500-R63, Lawrence Berkeley Laboratory, University of California, Berkeley, California, September 1975, bJ
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Question 11 Discuss the findings of your program to verify that the piping restraints and anchors are in the correct locations.
Indicate all deviations.
Response to Question 11 Certain of the piping restraints and anchors have not been installed. The progrhm for verification of correct location for all the piping restraints and hangers will be performed after these few remaining piping restraints and hangers are completed. The results of the verification program will be forwarded to the NRC at that time.
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4 Question 12 What type of detail.was used in modeling the reactor core in vibratory motion analysis?
Response to Question 12 The reactor. pressure vessel (RPV) and RPV internals were evaluated for the criterion 0.89 effective peak ground acceleration earthquake. The RPV (and its internals) were modeled and analyzed as part of the primary cooling system. The RPV shell and various core components were modeled using three-dimensional beam elements.
The model of the core represented essential components such as the inlet flow deflector, guide tube, guide tube and poison section and guide tube and fuel follower section.
The details of the modeling, analysis, and analytical results are presented in Reference 1.
References (1) EDAC, " Seismic Analysis of Primary Cooling System and Reactor Pressure Vessel (RPV)," prepared for GE, EDAC-ll7-217.05, dated June 1978.
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