ML19094A275
| ML19094A275 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 01/07/1971 |
| From: | Newmark N Nathan M. Newmark Consulting Engineering Services |
| To: | Case E US Atomic Energy Commission (AEC) |
| References | |
| Download: ML19094A275 (10) | |
Text
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.I N A T H A N M. N E W M A R K CONSULTING ENGINEERING SERVICES Mr. Edson G. Case, Director Division of Reactor Standards U. S. Atomic Energy Commission Washington, D.C.
20545 flle Cy. _
1114 CIVIL ENGINEERING BUILDING URBANA, ILLINOIS 61801 7 January 1971 Re:
Contract No. AT(49-5)-2667
Dear Mr. Case:
Surry Power Station Units l and 2 Virginia Electr~ower Company_
AEC Dockets No.~and 50-281 We are transmitting herewith 7 copies of a draft of our report concerriing the Operating License Review for the Surry Power Station Units 1 and 2.
A number of points remain to be clarified before our report is finalized.
We shall be pleased to discuss these items with your staff as appropriate.
bj p Enclosures cc:
W. J. Ha 11 Sincerely yours, PJ,1 ~\\ ~~I.A)~ o--J~
N. M. Newmark
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e Jfile C11 N A T H A N M. N E W M A R K CONSULTING ENGINEERING SERVICES 1114 CIVIL ENGINEERING BUILDING URBANA, ILLINOIS 61801 D R A F T REPORT TO THE AEC.REGULATORY STAFF STRUCTURAL ADEQUACY OF SURRY POWER STATION UNITS 1 AND 2 Virginia Electric and Power Company AEC Dockets No. ~*
and 50-281
- by N. M. Newmark and W. J. Ha 11 Urbana, Illinois 8 January 1971
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e REPORT TO THE AEC REGULATORY STAFF STRUCTURAL ADEQUACY OF SURRY POWER STATION UNITS INTRODUCTION AND 2 e
This report is concerned with the structural adequa.cy of the contain-ment structures, piping, equipment and other critical components for the Surry Power Station Units 1 and 2 for which application for a construction permit and an operating license has been made to the U. S, Atomic Energy Commission by the Virginia Ele~tric and Power Company.
The facility is located in Surry County, Virginia on the south side of the James River, approximately 30 miles NW of Norfolk, Virginia and 7 miles S of Williamsburg, Virginia.
This report is based on a review of the Final Safety Analysis Report and supplements (Ref. 1) and supplementary material (Ref. 2).
The report also is based in part on discussions held with the applicant and his consultants at the time of a visit to the Surry Power Station Site,on 20 October 1970 by W. J. Hall With D. Muller and R~ Lee (D~L/AEC).
A discussion of the adequacy of the structural criteria presented in the Preliminary Safety Analysis Report is contained in our report of March 1968 (Ref. 3), and unless otherwise noted no comment will be made in this report concerning points covered therein.
The design criteria for Class I components f6r this plant called for a design to withstand a Design Basis Earthquake of O. 15g maximum horizontal ground acceleration, coupled with other appropriate.loadings to provide for containment and safe shutdown.
The plant also was designed for an Operating Basis Earthquake of 0.07g maximum horizontal ground acceleration acting simulta~eously with other appropriate loadings forming the basis of containment design.
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COMMENTS ON ADEQUACY OF DESIGN Foundations The foundation conditions for the various Surry Power Station structures were considered in detail at the time of the PSAR review.
At the time of the site visit it was ascertained that, with Unit of the plant at approximately 80% completion, no unexpected motions of the foundations had been noted under the static loading imposed.
In the interim, since the PSAR review, two lateral load tests of piles were made to verify the computed deflections used in the design of the fuel building.
The pile tests showed that the computed values were reasonable; the values of possible lateral motion used in design are presented in Table 2.4-5 for the piling.
These values appear reasonable to us.
The calculated expected displacements for the various structures are presented in Table 2.4-5; the rattle space provided between buildings is pres~nted on page 2.4.8-3, namely a space of 6 in. between the pile-supported fuel building and the auxiliary building and between the fuel building and the containment structures, and a clearance of 3 in. between other structures.
The applicant indicates on page 15.2-17 that the clearance provided between structures is a minimum of 2 in.
In any event, whether the clearance is 2 in. or the larger values as cited in Section 2 on page 2.4.8-3, we feel that ~hese clearances are adequate.
The applicant advises in the answer to Question 2, 14 that a check of the vertical*motions have also been made, and that afurther check has been made to insure that the pipe stresses of all Class I piping between structures is within acceptable limits.
We concur in the approaches followed.
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. 3 A discuss ion of the analyses that have been carried out for*
1,quefaction potential 6f strata in th~ site area is presented in Section
-2.4.5.1; The summary of* the liquefaction studies indiiates that with the depressed piezometric levels under the Surry structures, the potential for liquefaction is quite low.
No indication is given for the region beyond the site structures, whether or not the depressed piezometric levels are applicable.
The shear stress values presented on pages 2.4.5-6 ~nd 2,4,5-7, resulting from computations, are justified on the basis of surface velocities only.
Question 2.16 addressed itself to this point and requested that the surface accel~ratio~s which are consistent with these calculations be provided; the answer to Question 2. 16 did not present such accelerations but instead attempted to justify the choice of surface velocities on the basis of matching surface intensities scaled from.other observations.
On this basis we find it diffi~ult to evaluate the shear stress values presented in the FSAR and must base our evaluation of the probability of liquefaction on the depressed piezometric _level and other studies and judgments which we have made independently.
However, it is not clear to us whether the possible 1 iquefaction of sands A and B surrounding the facility could lead io difficulties with the plant structures in the sense of causing a loss of lateral suppo~t for some of the~e structures.
The applicant should be asked to respond on this point.
Dynamic Analyses a)
Class I Structures The procedures followed in the dynamic analysis of major Class I structures is*described in Section 15.5.1.4 of the FSAR.
It is noted that the structures were analyzed using a Stone and Webster program, 11Container
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Vessel Seismic Analysis.
11 The details of this program are not presented in the FSAR.
It is indicated on page 15.5.1.4-2 and -3 that for the OBE a value of 5 percent of critical damping was used for the.entire reinforced concrete containment whereas for the Design Basis Earthquake a value of 10 percent critical damping was used.
The approach is difficult to follow, and as a matter of fact, Table 15.2.4-1, which lists damping values, suggests that a damping value of no larger than 5 percent was used in any of the analyses, which is contrary to the values cited above.
In any event, Question 5.12 was directed to obtain additional information in this regard since it had been stated in the FSAR by the applicant, on page 15.5. 1.4-3, that the damping factors used for design and for the resulting seismic response characteristics are conservative.
Unfortunately the answer to Question 5.12 is unintelligible without further detai 1 as to the method of analysis employed.
For example, it is indicated in the answer to Question 5. 12 that the damping values in the table therein are for radiational_ damping due to soil structure interaction, which is part of the structural system damping -- on the other hand it is noted that for the DBE the structural damping is 1 isted as 5 percent as contrasted to a value of 10 percent listed in the FSAR on p. 15.5.1.4-3.
Additionally, the summary presented in the last paragraph to Question S5.12 does not serve to ~larify the situation.
Question 5. 11 was directed at attempting to assess the relative magnitudes of the seismic stresses in the structures but these evidently can not be provided at critical locations in the containment structure in view of the method by which the analysis was carried out.
In any event, the applicant should indicate in some manner a) the phys.ical justification for
C
- 5 using the high value of 10 percent of critical damping for the reinforced concrete structure, which generally would imply a large and s ignif leant amount of deformation and stresses in excess of the yield strength, and b) whether or not it would be expected that zones of the containment would be overstressed if a value of 5 percent of critical damping were used in the structure under the DBE conditions.
b)
Piping The seismic de~ign approach for the nuclear steam supply system is outlined in Appendix B of the FSAR.
It is indicated that the Westinghouse portions of the system are analyzed in accordance with Westinghouse Report WCAP-5890, Rev. l, and we are in agreement with the approach adopted.
The answer to Question 4.12 indicates that the calculated seismic stresses in at least one main loop were low in comparison with the allowable seismic stresses.
If in general this is the nature of the results for all of the critical Class I piping, the approach adopted would appear to be *satisfactory.
The appl leant should verify that 0.5 percent damping was used in all the analyses for both the OBE and DBE conditions.
With respect to the dynamic piping analysis, it is not clear from the answer to Question 4. 12 whether or not the differential support motions for piping are included in the analyses.
Similarly, if there is piping running between buildings, the supports of which can move relative to one another, is this taken into account in the analysis.
c)
Vertical Earthquake Excitation, Critical Instrumentation and Controls The answer to Question 4.23 describes the approach employed for the vertical excitation.
For the containment structure and other rigid structures,
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6 the approach of using two-thirds of the horizontal acceleration is probably s,at isfactory.
In those cases in which rigorous dynamic analyses were made, from the descriptio.ns given in the FSAR and supplements thereto, it appears that the approach would be satisfactory.
In the case of equipment, it is indicated that the*acceleratioM magnitudes are included in the equipment specifications.
On the other hand:, the answer to Question 4.10 indicates that the vendor is required to validate component integrity under the specified seismic conditions.
The applicant should indicate how this validation is checked.
Moreover, merely indicating that the component is adequate to meet a specified acceleration level does not insure that, when it is mounted on some other element, frame, rack, etc., the response characteristics will be such as to maintain its adequacy.
The applicant should elaborate on the check-out approach followed for critical items of instrumentation and controls.
Reactor I~ternals The answer to Quest ion 4.13, dealing with reactor internals, makes reference to the H. B. Robinson response to Question 1. lB dated November 5, 1969.
Unfortunately, we have not had occasion to examine this response and therefore can pass no judgment on the adequacy of the reactor internals seismic design.
Cable Trays and Battery Racks There is no indication in the FSAR that the cable trays and battery
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racks have been analyzed for seismic loadings, and the applicant should provide additional information on these points.
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Design Stresses The di~cussion on pag~ 15.2-17 of the FSAR indicates that for those structures subjected to the DBE loading and analyses, allowable stresses do not exceed 90 percent of the yield strength for structural steel or in the case of reinforcing steel and concrete, the capacity reduction factor times the specified yield strength for reinforcing steel or the specified strength for conc~ete.
The answer to Question 5.8 indicates that the stresses cited are of the order of 50 percent of ultimate strength.
On.this basis there is reason to believe that the deformations associated with these design stresses are generally low although the amount of deformation that may be control] ing under the maximum loading conditions between the various materials used may not be precisely similar.
We must assume, for lack of other evidence, that where composite action is called for the compatibility of the deformations in the various elements have been studied carefully and taken into account.
SUMMARY
COMMENTS On the basis of our review of the FSAR and the answers to the questions given in the supplement, it appears to us that the design probably can be considered to be adequate in terms of provisions for safe shutdown for a Design Basis Earthquake of O. 15g maximum horizontal ground acceleration and to withstand otherwise the effects of an earthquake of half this amount.
However, the information made available to us for evaluating the adequacy is not as complete as would be desired; on the basis of the information made available to us it would be our judgment that the margin of safety inherent in the design is probably not as great as that intended.
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In arriving at a better evaluation of the margin of safety inherent in the design, we have pointed out a number of items requiring additional clarification by the applicant, specifically with regard to: (a) the 1 iquefaction analysis; {b) the dynamic analysis of the major Class I structures, including the amount of d_amping emp*Joyed and the details of the type of analysis carried out; (c). piping analysis to account for differential motions of support points; (d) seismic design of mounted equipment; (e) analysis of the adequacy of such equipment items as cable tray supports, reactor internals, and battery racks.
Additional information on these items will provide a better basis for evaluation of the margin of safety inherent in this design.
REFERENCES
- l.
"Final Safety Analysis Report -- Vol. l -
5* (Part B), plus Supplement Vol.
land 2, 11 Surry Power Station Units land 2, Virginia Electric and Power Company, AEC Dockets No. 50-280 and 50-281, 1969.
- 2.
Supplementary material:
(a)
"Surry Nuclear Power Station Preoperational Environmental Radiation Surveillance Program," Report, May 1, 1968 through June 30, 1970, Virginia Electric and Power Company.
(b)
"Lateral load Pile Tests -- Surry Power Station -- Virginia Electric and Power Company," prepared by Stone & Webster Engineering Corporation, Boston, Massachusetts, July 1968.
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11Adequacy of the Structural Criteria for the Surry Power Station Units.1 and 2, Virginia Electric and Power Company, 11 AEC Dockets 50-280 and_ 50:-281, by N.
M. Newmark, W. J. Hall, and A. J. Hendron, Jr., -
March 1968 *
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