ML20127K063
| ML20127K063 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/13/1985 |
| From: | Gucwa L GEORGIA POWER CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| 1700N, NED-85-391, NUDOCS 8505220026 | |
| Download: ML20127K063 (87) | |
Text
- Georgia Powdr Ccmpany -
333 Piedrnant Atenue
- Atlanta, Georgia 30308 c Tdephone 404 526-6526 Mailing Address'
- Post Offee Box 4545 Atlanta, Georgia 30302 Georgia Power L T. Gucwa the southern ehxtnc system
~- Manager Nuclear Engineenng ED-85-391 and Chet Nuclear Engineer 1700N May 13, 1985 Director of Nuclear Reactor Regulation
- Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S.' Nuclear Regulatory Commission Washington, D. C.'-
20555 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 EDWIN I. HATCH NUCLEAR PLANT UNITS 1, 2 SUPPLEMENTAL INFORMATION ON SEISMIC DESIGN ISSUES r
Gentlemen:
A request for supplemental information to allow completion of the EC's review of. submittals regarding the seismic ' design of structures and equipment at Plant Hatch was transmitted by your letter of April. 2, 1985.
The attached information is submitted pursuant to that request.
If you require amplification of any of the attached information, please contact this office.
Yours very truly,
/f 4m L. T. Gucwa WEB /eb Attachment xc:
J. T. Beckham, Jr.
H. C. Nix, Jr.
J. N. Grace (W C-Region II)
Senior Resident Inspector g2hk go 21
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.m RESPONSE 'IO NBC APRIL 2,1985 RBQUESP FOR'
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- ADDITIONAL INEDf?RTION GT SEISMIC DESIGN 1
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'Q.1 In' Georgia Power Company's ' (GPC) letter _: to 'NIC Region' II'l dated January 6,' 1984, you stated that "In: some cases ' discrepancies were found to
. exist between.the : floorc response spectra ' used.and spectra broadening ocamitments.". Please clarify which-spectra broadening comitment you 'are -
. referring toL(PSAR vs. FSAR and respective percentage).,
Response
'Ihe floor response' spectra peak broadening ccsmitments for Plant Hatd. are outlined 1in the "Sumary of Reviews ' and Analyses Performed for. the Part 21 f Evaluation. Concerning Floor Response Spectra Peak Broadening for E. I. Hatch Nuclear Plant Unitsl and 2";'section I.B, included as an attachment to letter NED-84-622, dated December 14, 1984.
The spectra broadening commitment referred to in letter NEMM-000, dated
-January 6~,1984, is that of + 10% peak broadening as specified in the Unit 1 FSAR and the thit 2 PSAR.
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" [GPC : letter 'of Plebruary 10. 1985,,(note should be ' January 6, ^1984),
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' indicated - that' the PSAR' and FSAR ccmunitments were - exceeded in - some ~
'- ~
cases. _~ ~Also', in-the same letter you stated that "the floor response V
spectra. used for.-the analyses 'were.. plotted as smoothed; upper envelopesj of the - calculated; raw curves;" _ Please provide a detailed
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' discussiono explaining this process and emphasize its adequacy and
. acceptability. by. reference ;to and~ comparison - with the. staff
. acceptance criteria in the pertinent sections ~ of the SRPs and.RGs..
~
~
. In ; addition,.. provide J the'. screening criteria ~ L used for establishing
- potentially unsafe conditions as 'related to the floor' response peak broadening and seismic analyses evaluations.
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4 Response :
A.
'1he staff acceptance criteria contained in the pertinent sections of the
"~
SEP and IC's 'is. not a commitment for the E..I. Hatd1 Wclear Plant thit ~
11'or 2.
As discussed in section I.B of the December 14,'1984 report (NED-84-622), GPC's : ccanitment to peak broadening - for both IJhit 1 and
. thit - 2'- is +10%.
Due to 'the fact this. commitment appeared not ' to be
- met, t.
Part 21: safety evaluation was initiated.
The results of this i"
evaluation were stated in GPC letter of September 12,1984 (NED-84-484),
the November 16,.1984 meeting with the NBC, and the report submitted to c
.the. NIC urder GPC's letter of December 14, 1984 No reportable
' condition-in accordance with the criteria of Part 10CFR21' existed for>
n the original floor response spectra.
In GPC letter' of January 6,1984 (NE[>84-008), it stated that "in all cases examinedi the (original) floor ~ response spectra used fac ~ the
- analyses were plotted as smoothed upper envelopes of the calculated raw -
curves."
'Ihe following is a discussion of the processes used in plotting the original floor response spectra for tJnits 1 and 2.
D y 13,L 1985 '
-7_. <
Response to 0.2 (Continued)
Page 2 of 5' We process used for all original Ibit 1 floor response spectra (FRS) was to hand envelope the combined raw N-S and E-W FRS to form a smooth horizontal FRS at a given elevation.
This hand enveloping process removed valleys and rounded the major peaks.
The major peaks of.the thit 1 Reactor Duilding were broadened by hand by normally drawing a horizontal line be ween the two adjacent vertical frequency lines in between which the frequency of the major peak spectra acceleration occurred.
The original Ibit 1 Diesel Generator Building FRS were the envelop of the combined N-S and E-W raw FRS peaks of two separate analyses to form one horizontal FRS at each elevation; each analysis considered an estimate of a limiting value of the soil shear modulus.
The process used for the original Unit 2 floor response spectra (FRS) except for the Diesel Generator Building was to plot the FRS using mmputer programs that performed the smoothing and broadening of the raw FRS data. The thit 2 Reactor Building FRS of record were broadened 110%
at all frequencies.
The thit 2 FPS for the Control Building and the Intake Structure were plotted so peaks were enveloped and all valleys removed and the major peak was broadened. The broadening for these two structures was obtained by raising all spectra acceleration data point values that fell within 110% of the frequency of the maximum spectra acceleration value to that maximum spectra value.
The original thit 2 Diesel Generator Building FRS were the envelope of the canbined N-S and E-41 raw FRS of three seprate horizontal analyses to form one horizontal FRS. The vertical FRS was the envelope of the three separate thy 13,1985 -
.kh -
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_ Response to 0.2 (Continued)
Page - 3 of - 5
. s._
-vertical analyses.- We three separate analyses represent the estimated ~
variation of the ~ soil shear modulus.
The ~ FPS for' the main stack was
[
plotted as an envelope of the peaks and elimination of all valleys;with some broadening of the major peaks.
Examples-of the smoothing and enveloping' procedures used in the development of the original FRS are shown in Figures 1, 2, and 3.
B.
In. regards to ~ providing the screening criteria used to establish.
potentially unsafe conditions, the following discussion is provided.
We purpose of the initial screening criteria was to identify those areas that had the highest potential for an unsafe condition as related to the new FPS.
After a review of equipment and structural subsystems associated with these areas, no unsafe cx>ndition was found relative to
)
the new FRS.
Werefore, based on the safety evaluation, there is no reason to expand the screening process to. include other floors where the FFS exceeded the original FRS by an even lesser amount.
Se general aspects of the screening process have been presented in the November 16, 1984, meeting and are discussed in sections I.E and II.B of the report submitted on December 14, 1984 (NED-84-622) by GPC.
%e detailed aspects of this screening process are presented here.
W 13, 1985. -
1 Response to Q.2 (Contirated)
Page 4 of 5 Initially, the original FRS at 5% damping was conpared to the new (1984)
FRS at 5% damping for the DBE to determine if there were any cases where the new FRS ' exceeded the original FRS.
An increase of less than or
. equal-.. to 25% was considered insignificant due to the inherent conservatism that has been docunented to exist in the seismic analysis process.
This process includes conservatisms in the way FRS are generated, the fact that equipment tends to be overtested and permissible stress levels are well below actual failure levels, and past earthquake experience has shown that many subsystems have significant earthquake resistance.
GPC consultants, Dr. Robert Kennedy and Mr.
Donald Ianders, concur with the use of 25% as used in this Part 21 evaluation.
In the case where the new FRS exceeds the original by more than 25%, an ergineering evaluation was made to determine if the t
increase was significant enough to warrant an evaluaticn of equipment and structural subsystems at the particular location.
This evaluation included a review of the amount the new FRS exceeded the original FRS, the frequency range (s)~ at which these excesses occurred, and the magnitude of the spectral accelerations.
The consultants were involved in the engineerire evaluations and concurred with the judgment that only eight locations existed where the original FRS were exceeded enough to warrant an evaluation of the affeet of these new FPS on equipment and structural subsystems.
Figures 4 through 11 of the report (NFD-84-622) show these eight FPS cceparisons.
The FRS that exceeded the original FRS by more than 25% but were not judged to be significant had the
' May 13, 1985 - -
f Response to Q.2 (Continued)
Page 5"of 5 excesses over a very narrar frequency band and/or the spectral-s
- acceleration values were very small.
Figures 4 and 5 are examples of cases where. the nw FRS exceeded the original EPS but were not
~ considered significant.
Ebr any case that was judged.to warrant an evaluation of equipment and i
structural subsystems, additional comparisons were made of the original DBE FPS at the original DIE damping value used in design to the new FRS at - the PVIC-r+y Taed damping and 7% damping for piping systems and cable tray supports, respectively. The use of higher damping for piping i
systems and cable tray supports has been justified by recent tests.
Additional discussion and justification for higher damping for these two subsystems can be found in sections II.C and II.D of the report sulaitted to the NIC on December 14, 1984 (NED-84-622 ),
and also the enclosures submitted to the NIC on January 16, 1985 (NED-85-031),
addressing the 50.59 evaluation for higher damping. No cases existed
- that warranted an' evaluation of the inpact of the new FRS on piping and cable tray supports.
The evaluation of equipment and structural subsystems at the eight' locations is discussed in section II.B.2 and II.B.3 in the report submitted on December.14,1984 (NED-84-622).
my 13,1985,
7 -
l COMPARISON Or FRS CURVES VALUES SHOWN ARE FOR RAW N-S SPECTRtM (SOLID).
DAMPING 3%
OPER TING BASIS EARTHQUAKE.
RAW E-W SPECTRUM (DASH)
DAMPING 3%
MULTIPLY BY 1.7 FOR l
HORIZONTAL SPECTRtM (CHAIN)
DAMPING 31 DESIGN BASIS EARTHQUAKE.
FIGURE 1 l
Example of. the enveloping process for the original
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Plant Hatch Unit #1 Reactor Building FRS.
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Unit #2 FRS.
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-O.3 We understand from GPC letter of September 12 E 1984, "that a_ reportable-condition per. thei criteria of. Part l 10. CFR 211 does not exist for the
-. discrepancies in the analyses. of the floor response spectra - (FRS) for -
,~
_ Edwin I; Hatds' Nuclear Plant-tktits 1 'and. 2."
Therefore, we request a discussion comparing the results of the reanalyses with the proposed
=<v=*vx:e criteria, nesponse
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1hese reanalyses were performed in compliance with FSAR commitments except for.
s the followings (a)
'Ihe tktit 2 PSAR conmitment of i 10% peak broadening was used in lieu of the unintendel erroneous value of i 15% given in the tktit 2 FSAR.
(b)
Higher damping than _specified in the FSARs was used to evaluate the c
-impact of the new FPS on piping systems and cable tray supports.
All. modeling changes and the use of new - synthetic time histories are ' in conformance with the FSAR casatitment. The FSAR commitments as amended above are considered to be the acceptance criteria for Plant Hatch.
A presentation of f.he sumanary of the Part 21 evaluation was made to the NRC on Nbvember 16, 1984 A report entitled "Supunary of Reviews and Analyses Performed for the Part 21 Evaluation Concerning Floor Response Spectra Peak Broadening for E.
I.
Hatdt Nuclear Plant-thits 1 and 2 Decenter 1984" was submitted to the NRC under letter of December 14, 1984 (NED-84-622).
Both presented a discussion of the engineering methodology, engineering criteria, and the results of ~ the remnalyses and how they relate to the FSAR counitments.
'the use ' of i 10% peak broadening for Hatch thit 2 ns being the correct licensing comunitment has been doctanented in Section I.B of the referenced December report and the response to Question 5 given in this submittal.
GPC
_ ? v 2', 2'85-L
Response to O.3 Page 2 of 2 -
'is revising the thit 2 FSAR to the correct licensing connitment of +10% peak broadening..
I'
' GPC has completed a 50.59 evaluaticn based on the attachment to GPC letter to the NBC _ of January 16, 1985 (NED-85-031), concerning higher damping than pre viously allowed in the IIatch FSARs for piping systems and cable tray supports.
'Ihe result of this evaluation is that the use of these higher darpirg values does not constitute an unreviewed safety question per the criteria of 10 CFR 50.59.
GPC is revising the Unit 1 and Unit 2 FSARs to permit the _ use of the damping values as described in the attachments of January 16,1985 (NED-85-031), letter and the response to Enclosure 3 given in this submittal.
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-In GPC: ' letter 'of February _loi 1984, -you-stated- 'that~ "the.
W' architect / engineers have : found some less significant l discrepancies. -
'Ihese - discrepancies are. minor. in-nature and. will. require - that. some -
1 revisions be, made to' the-FSARs of both units'"
Although we) understand final' resolution of these'-discrepancies 'can be documented as FSAR updates, Lwe currently need identificatiori and' discussion of any discrepancies that: effect-the modeling,, analyses and Jdesign of category I structures. Please provide this information.'
D eponsef The 'following statement was made in GPC's ' letter of -February 10i 1984, ' to the 10 0.:
ll
";..the architect enjineers have found some less sig-
.nificant discrepancies. These discrepancies are minor in nature and will require that some revisions be made to the FSARs of both units. - Georgia Power company believes that these concerns can be resolved in FSAR updates."
A draft of revisions to both FSARs concerning seismic, analysis and design will be sutaitted for your information on or before my 31, 1985. 'these revisions
. correct the "less significant discrepancies" noted in GPC's letter of February 10,1984 (NID-84-066), as well as reflect the inclusion of the use 'of higher-damping for piping systems and a ble tray supports, the development and use of
. the now (1984) ; FRS, and correcting thit 2 peak broadening commitment to the correct percentage of + 10 percent.
my 13,1985 -
1
~
. Response to O.4
' Page 2 of'3-The-' following :is an identification and discussion of these "less significant-discrepancies" that are felt to effeet the modeling, analysis and design of Catiegory I structures.
(1) thit 2 FSAR 3.7A.l.3 Damping Value, p. 3.7A-3 This section was revised to indicate that the soil damping values used 1
in the analysis of seismic category I structures may be taken from Table 3.7A-1 or may be calculated using equations listed in Table 3.7A-2.
(2)
' Unit 2 FSAR 3.7A.1.6 Soil-Structure Interaction, p. 3.7A-3 This section was revised to refer specific discussion of the soil damping to section 3.7A.l.3 where it clearly indicates what soil damping values are allowed.
(3)
Unit 2 FSAP 3.7A.2.6 Development of Floor Pesponso Spectra, p. 3.7A-7.
This section was revised to show that the frequency steps listed in the section may not have been used for all time history analyses for development of FRS. The frequencies listed are only an example of one set of frequency steps. However, the frequency steps were selected my 13,1965 -
Response to 0.4-Page 3 of'3
. such : that the" frequency -interval between consecutive. frequencies typically did not exceed ten percent and in no case exceeded-thirteen
_ percent of the lower frequency for the frequency range from 1 to 22 Hz. The structural frequencies were included in all cases.
(4) thit 2 FEAR 3.7A.2.14 Analysis Procedure for Damping, p. 3.7A-9 This section was revised to indicate that the mass proportional method described in the FSAR was an example of one of the cxxnposite modal damping methods that was acceptable for use in analyses of seismic Category I structures.
The primary methods used were stiffness proportional, modal weighing, or the Tsai method.
m y 13, 1985 y -
,g Q.5' In GPC letter of my 24, 1984, you stated that the new seismic design evaluation would be performed for 110 percent peak broadening.. You assumed that the +15 percent stated in the FSAR was an erroneous value.
Nevertheles's, we believe that the use of 10% peak broadening requires justification-as indicated in the current SPP Section 3.7.2.II.9.
mreover,- the staff acceptance criteria require 115 percent peak broadening if other requirements described in the SRP are not met. We believe that this criteria was in effect during the FSAR review of Hatch-Unit 2.
Therefore, please demonstrate the adequacy of the +10 percent peak broadening criterion considering_ structural properties, soil properties, and soil structure interaction.
^ Additional references can -be found in Pegulatory Guide 1.122 "Developnent of Floor I:esign Besponse Spectra for Seismic Design of Floor-Supported Bluipment er Components".
Response
It is our position that the use of 110 percent peak broadening on Hatch-Unit 2 does not require additional justification. The " Summary of Peviews and Analyses Performed for the Part 21 Evaluation Concerning Floor Response Spectra Peak Broadening for E. I. Hatch Nuclear Plant Units 1 and 2", section I.B.,
included as an attachment to letter NED-84-622, dated December 14, 1984, outlines our basis for this position.
f A discrepancy in the Final Safety Analysis Report (FSAR) commitments.for peak broadening of the seismic floor response spectra curves was brought to the attention of Georgia Power Company (GPC) on December 20, 1983, in the course of an analysis performed for the recirculation pipe replacement on Unit 2.
Section 3.7A.2.8 of the Unit 2 FSAR states that the computed FPS were smoothed, and peaks associated with the structural frequencies were widened by "i 15 percent."
It was discovered that the floor response curves generated.
for the Unit 2 Beactor Building were widened by i 10 percent. Based on my 13,1985 -.
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.seiWc. analysis,- GPC has concluded that-the. intent and appropriate commitment N
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..was' to broaden the cFPS by 1110 percent.
The commitment to a 1.10 percentj,.
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J-w broadening, was3, accepted bydtM, Nuclear Regulatory Commission's (NPC's) at g.
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- Yeonsultants in theirf review ofjthe PSAR for Unit 2 and by-the NBC
.y in the 4 Safety Evaluation (Reports for the Unit 2 PSAR and the Unit 1 FSAR.
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,4%W c,cIts'_ question A-5 of the Unit. 2 PSAR the NBC requested the followings - List all ic r
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M JCategory I((specified as Class 1 in PSAR) structures, systeus;. and 'bongonents, g
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... including reactor' internals; piping, le hanger trays and ' mounted { equipment; and the method of seismic analysis '.(modal analysis response spectra, modal
_ tests)
(
" analysis -time history, equivalent. static loads,' etc.) or empirical:
analyses which will-be employed in the design, including applicable stress or t
deformation criteria. ' Provide a brief description of. all methods that are w
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used'for seismic analysis.
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In Amendment 7, page A.S.2 under Equipment Georgia Power f h ny responded:
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- The. floor. response ' spectra will~ be anoothed 'such that,ithe response curve will i
']
E (be' an ; upper bound envelope of 'all the acceleration' points.
Whenever the u t,e response curve comes to a peak, the curve will be made flat in a region i 10%-
t
- of that peak.l frequency.
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Response to O.'s..
= - Page 3 of 4 4
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In the Safety Evaluation ~ Report '(SER) for Hatch-Unit 2 issued July 21, 1972, page 172;ithe ABC noted, "The answer to Question A.5 in Amendment 7 elaborates-
. on! the dynamic ' analysis. procedures 'to be -' employed for structures;. reactor -
pressure vessel and ~ internals; equipment, - piping,._ and cable trays. - The L'
criteria described there, and in the answer' to Question A.7 (Amendment 8), are
. acceptable to us.
In theiriletter! to the' AFC dated Ibvember 2, 1971, Mr. ' N. M. Newmark, W. J.
. Hall and A. J. Hendron stated:
"We believe thelprocedures used in the design and analysis are in accord with
~
the~ state-of-the-art.
It is our conclusion that the design ircuryczates an acceptable range of margins ~of safety for the hazards consideration."
Signed:.W. J.' Hall f
- The ABC staff and consultant's concurrence with the. Unit 2 seismic design
~
basis,followed detailed discussions prior to Advisory. Ccamittee on Reactor Safeguards - (ACRS)y meetings.
GPC agreed ' to revise the design basis seismic
~
- response spectra, thereby using a' more recently accepted licensing spectra, however, no discussion regarding response' spectra broadening occurred.
4
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. m y 13,'1985:
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Section 1.3.2 of ' the Unit.2 FSAR lists all-the significant changes that have been made.to the plant since submittal of.the PSAR. There is no reference to~
a' change.to increase the peak broadening to i 15%.- ' This change would have been extremely significant since it 'would necessitate changes in the purchase specification of all equipment for Unit 2 because the plant was substantially designed at^ this time.
It is based upon this review that GPC contends the appropriate peak broadening conmitment for Unit 2 was i10%.
In additions we are aware of at least five (5) other units - that received operating licenses during' the same time frame as Hatch-Unit 2 (July 1975-June
'1978). that used 10 percent peak broadening to account for expected variations
.in. structural properties, damping. soil properties, and soil-structure interactions. We have also-identified a number of units licensed -after this time' period that used 110 percent peak broadening as well, therefore we feel that the requirements of the current SRP regarding peak broadening should not
.5 be. applied to Hat &-Unit 2.
1
?.
- ley 13,1985 e:
=
0.6
'In' GPC letter of my 24, 1984, you stated that new analyses would address ' the impact of new floor response spectra' on cable tray supports and that these analyses would be based on higher damping ratios than specified in the FSAR.
In support of these new higher damping values you stated, '*Ihe new damping ratios are consistent with those that have been reviewed and accepted by NIC for seismic analysis of some recent projects."
The staff has based approval of higher damping values than those identified in R.G.
1.61 on test results of the MCO/Dechtel tests.
The staff has requested for each case, where the applicant requested the approval of higher damping values,.that - the applicant establish the applicability of test results to the system used in their plant.
The applicability could be demonstrated by establishing positive correlation between the cable tray systems for the case in question and~ those used in the NCO/Bechtel tests.
Provide a detail discussion substantiated by detailed test results supporting your claim and/or show comparison between the Hatch cable tray systems and applicable ones utilized in the MCO/Bechtel tests.
' Response
-As stated in Section 3.0 of our previous subnittal, attached 'to letter NED-85-031, dated January 16, 1985, the specific tray and trey supports used at MP have not been specifically tested.
However, the diversity of tests, and the conclusions readhed regarding the mechanisms which make available increased damping, can be confidently extended to the IRP support system.
Details of the tray and supports used at MP are discussed in the response to Question-15.
In
- general, the supports consist of structural tubes cantilevered from floor slabs and concrete walls. Strut members are welded to the sides of the strong structural tube and serve as a bolted attachment point for strut-type support arms which, in turn, accept the cable tray.
As indicated in the previous submittal, the IRP cable tray is primarily P-W Industries ladder and solid bottom trough. The Bechtel/MCO test program my 13,1985 ' w.
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' Response to Q.6'(Otmtinued)~
Page-2'of 3
? adder trays; a punched
~
. utilized 1various s trays, - including ^ three - types of l
- bottout type tray, - and l trough type tray, provided by 'various. manufacturers..
^
.Ei anpports were primarily ~ flexible strut trapeze type supports, - both braced and 3
unbraced,'.as shown in Figure 3 of the January.16,1985 (NED-65-031). sutaittal, 4..
and accepted from one to five: tiers of tray. Additionally, cantilevered strut,
~
rod-type hanger and' rigid attachment of the tray to Ltest: apparatus were considered. Cable loading was varied from 0 to 100% -(50%.to -100% loading was -
- a..
considered for the important earthquake input' testing)... Cable diameter tie
.down methods were also evaluated..
~
!The. general conclusion reached from evaluation of' the results ~ of.over 2000 dynamic tes's in this program is that 'neither cable diameter or traf type is t
-an isportant factor in determining the effective damping available to the
. total system.'. LiKewise, support configuration-and attad. ment details also do
- c
. not directly ' affact the L availability of damping.
'Ihe support configuration and attadument details..are important only in that they (1) have sufficient _
- strength ' to. accept necessary seismic loading, and (2) ' acting as essentially massless. springs provide adequate stiffness such that dynamic characteristics of the system provide the -mechanism to create dasping.
As indicated in our previous.sutmiittal, the significant portions of system damping were the result of the amount of energy absorbed between the ' adjacent cables and through friction between cables and the cable tray, n
i, l
~ '
' lety 13, ' 1985
.t h
r,--s to 0;6 (Continued)
- Page 3 ~ of 3 '
. r.
Fundamental natural frequency - of the, system ' tested ' with flexible supports ranged primarily. from 2-6 Hz although in sore cases ' natural frequencies lower than'2 Hz were encountered with damping in excess of 20% critical. The. rigid supports had natural frequencies between 9 and 25 Hz. The test data suggests that; the natural frequency range - between 6 and 9 Hz also 'will result in similar dampireJ characteristics.
The system natural frequency of the ISP system is within the range of.2 to 25 Hz.
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May 13, 1985 -
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.Q.7 In ' GPC letter 'of December 14, 1984, you stated that you, "have
. reviewed the te&nical specifications and have determined that no te&nical specification - violation exists."
Please explain the relationship of the technical specifications and the reported concern
- with respect 'to reduced. percentage value of floor response spectra peak broadening, increased damping values for the cable tray supports
-evaluation, and 'other : structural ~ engineering items reported 'in your previous letters (NED-84-008, NEIb84-0066, NED-84-274, and NID44-484).
~
Asponses?
Based on a thorough review of the tJnit 1 and 2 technical specifications, there
-is no explicit requirement in _regards to -seismic design and analysis of structures. 'The IJhit 2 technical specifications make no reference to seismic 3
l-requirementst while the Unit 1 technical specifications make one reference to lseismic requirements in - Section 5.0.F which deals with horizontal ground accelerations for OBE and DBE.
In any event, it is GPCo's position that the technical specifications would 6*
not be impacted because, to date, no operability concerns have been identified,-- as stated-in our previous letters (NED-85-032, NED-84-641, NED-84-637, NED-84-622, NEDM S4,- NEIF84-274).
1 I
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4 d
m y 13, 1985.
m 0.8 In your letter of December 14; 1984, you stated that the results of an independent audit of design commitments in the FSARs for Units 1
.. and 2 indicated that,
" Preliminary,information from this audit indicates that, 'other commitments in the FSARs were met;'"
Please explain what is meant by preliminary.information and state if the final results confirm the same conclusions.
Also, provide' the findings of the audit. and a discussion for each of the concerns
-resolved by the audit.
Responses At the - time of submittal of our letter bTI)-84-622; December 14, 1984, the contract auditor; Impell' Corporation, had not issued a final report of their audit findings.
Our' statement that "tther commitments in the FFAR were met..." was based on verbal preliminary reports made by the audit agency.
A week later, on: December 21, 1984, Impell Corporation reported their findings formally to GPC. The overall conclusion of the audit was "As a general rule, the review did not find any. generic problems in the civil / structural area, except for the generation of the design spectra, which had been identified prior to the start of the Impell Review." Thus, the final report confirmed the preliminary results cited in our letter.
A few discrepancies of a minor nature were noted in the audit. These items will be addressed through the GPC Quality Assurance Department and the apr quiate engineering agency.
D m y 13, 1985 -
?
Q.9 In the enclosure to GPC December 14, 1984, letter, you restated your cmsiderations -for the' values of peak broadening that did not meet current NPC criteria.
We staff evaluation of-the FSAR was performed based on information presented in ESAR Sections 3.7 and 3.8.
It is the staff position that if criteria differing from those criteria presented in these FSAR sections were used, justification and an amendment to the FSAR are' required. In this regard, provide in-depth discussion on engineering methodology, assumptions, and.other~ evaluation data esployed during your reevaluation of the Hatch
~1 and 2
seismic / structural design.
In as much as construction is completed, the information should be much more detailed than that presented for a typical FSAR, 'and should duplicate information, as required, when comparisons need to be made to the original FSAR.
. Response:
'The report enclosed in the GPC letter of December 14, 1984 (NFD-84-622),
provides a summary of the safety evaluation performed in regards to the peak broadening of the FRS. As stated 'in that report and as noted in response to Question 3 herein, the generation of the new (1984) FRS met or was more cmservative than the licensing cxmmtitments.
Details of the process of generating the new FRS are given in section II.A of the report. The FSARs for both Units 1 and 2 are being revised to include a detailed discussion of the new (1984) FRS.
These additions to the FSAR will also contain data on the
. synthetic time histories used to develop the new FPS. The Unit 2 FSAR is also being revised to state the intended percent broadening of +10% as discussed.in 3
section I.B of the report.-
As' discussed in sections II.C and II.D of the December report, the evaluation of the new FPS for piping systems and cable tray supports used higher danping
~ han specified in the FEAR. We attachments to the GPC letter of January 16, t
1985, (NED-85-031) provide a description of the higher damping values for Ptiy 13,1985,
,+
- }.
Desponse to Q.9 (Continued)
. Page 2 *f 2 these particular subsystems, as well.as the preliminary 50.59 evaluation. The FSARs are now being revised to incitxle these higher damping values for new cr replacement systems and load reconciliation work.
A draft of the 1985 revisions to both Unit 1 and 2 FSAPs concerning seismic analysis and design to be sutmitted for your information on or before my 31, 1985, includes the FSAR changes just discussed.
~ It is believed that the information provided in the report attached to the GPC letter of De._ M r 14, 1984 (NED-84-622), and the attachment to the GPC letter of January 16, 1985 (NE!> 85-031), plus the information of the my 31, 1985 submittal provides the necessary in-depth discussion on engineerirrt methodology, assunptions, and other evaluation data employed during the
- reevaluation of the Hatch 1 and 2 seismic / structural design.
I Msy 13, 1985 -
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O.10 Provide a comparison of the ground response spectra. developed by the '
synthetic time history method for damping values of 5 percent and 2 percent.-
V hsponset Sununarized below are the ground response spectra versus design spectra comparative plots that are being supplied.
(1) Unit 1 Plots a.
(bmparison at 5 percent damping - Figure 1A.
b.
Camparison at 2 percent damping - Figure 1B.
(2) Unit 2 Plots a.
Comparison at 5 percent damping - Figure IC.
b.
Comparison at 2 percent damping - Figure 1D.
The Unit 1 and Unit 2 time histories used in the evaluation described in the A.-Nr,14,1984 (NED-84-622) report were generated such that the 5 percent a
damped ground response spectrum for each time history envelops the corresponding design spectrum at all frequencies from 1.0 to 33.0 Hz.
The decision to envelop the 5 percent damped design spectra was based on the fact that 5 percent is the nost consnonly used DnE danping value (see table 12.3-2 of the Unit 1 ESAR and table 3.7A-1 of the Unit 2 FSAR).
In addition, the
- significant structural modes of vibration for the various Hatch category I structures are characterized by DBE damping values equal to or greater than m y 13, 1985.-
posponse to 0.10 Page 2 of 2 5 percent.. As can be seen from figures lA and IC, the 5 percent damped ground
. response spectra for both units do, in fact, envelop the corresponding design spectra at all frequencies. The fact that the 2 percent damped ground spectra for the two units do not envelop the. corresponding design spectrta at all frequencies is not significant. This is because the FPS used to perform the uncoupled analysis of subsystems are generated frca floor time histories of structures whose responses coincide with damping values of 5% or higher.
In addition, the use of 2% daw FRS is limited to defining input for piping systems' frequencies of 20 Hz or higher (using the PVPC damping criteria for piping) where the spectral accelerations are low and typically there is little or no difference between the accelerations for different damping values.
2 May 13, 1985 -
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Q.11' Explain how the. equivalent radiation damping was used in the development of the floor response spectra, as stated on Item 3 of page 11 of GPC December.14, 1984,. outatittal.
Also, discuss Why the
- use of equivalent radiation desping per - table 3.7A-2 results,in hi@er desping than ordinarily used.
nesponses originally, the thit 2. analyses of the Control Building,. Intake structure, Diesel. Generator Building, and the min Stack used 5.0% of critical desping for' DRC as given in PBAR Table 3.7A-1.
We Unit 2 Reactor Building analysis used the radiation-desping por the equivalent foundation desping ocefficients of the FSAR Table 3.7A-2.
We Unit 2 FSMt allows the use of either Table 3.7A-1 or Table 3.7A-2 for soil damping.
As part of the Part 21 evaluation for peak broadening of FRS, new FPS were generated. This required a time history analysis of each 8sissiic Category I structure in Which soil damping, among other things, is required.
Based on the recommaandation of GPC consultants, the soil damping was recalculated as discussed in section II.A.3 of the December 14,1984 (NED-84-622) report.
In no case did the calculated soil desping values exceed those allowed by Table 3.7A-2 and, therefore, never exceeded the PBAR commitments. %e recalculation of. soil damping for thit 2 was considered an enhancement of the building models and would, therefore, produce more realistic floor response spectra.
Se use of equivalent radiatica desping per Table 3.7A-2 does not result in hi@er desping than ordinarily used, but is higher than used in the Part 21 evaluation.
Soil dasping calculated using Table 3.7A-2 did produce higher i.
l l
.my 13,.1985.
4 p.
t nesponse to 0.11 E'
l-Page 2 of 2 -
p-r.
damping values than 5% dtich was originally used for most Unit 2 structures.
Response to Phclosure 2 of the NIC's letter of April 2,1985, provides soil damping data and other information to illustrate how soil damping was
. originally used and how it was used for the Part 21 evaluation.
l 1
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l m y 13, 1985,
i
-Q.12 2:xplain why you' consider the current cable trays' evaluation. " overly conservative" with the assumption that the trays are considered to be continuous for the entire run.
- Also, state your criteria for identifying the cable tray supports whidt have the greatest chance of being overstressed for all load conditions; and identify with some specificity the proposed action.
Response
A.
In letter NED-84-437, dated December 21, 1984, the following information was included.
the current tracking system is
' overly conservative' since cables are asstaned to run the entire length of the tray rather than the distance between the entrance and exit points of the cables in the tray."
This means that the tracking system assumes i.
all cables located in a tray enter at the beginning and exit at the end. In actuality, cables enter and exit along the entire length of the tray. We consider this system to be conservative because in almost all L
cases the load on a cable tray section is calculated to be greater than it actually is, i
R.
In letter NED-85-032, dated January 16, 1985,. Georgia Power ' Company 1
outlined a program which will ensure that all cable tray supports will 1
be upgraded, if required, to meet PEAR comitments.
%e program consists of the following steps.
'4.
m y.13,.1985
+
7 L
/
i Response to 0.12 (continued)
Page 2 of 5 t
(1)
Establish a program to track additional loads on ' cable tray i
supports:
I a.-
Beginning in May 1983, donever a new conduit or other r
attadmont is N to an existing cable tray support, the entire support is as-built and analyzed for all loads, new and existing.
b.
On November 14, 1984, a program was established to track i
and monitor all increases in loads on cable tray supports due to the additions of cable to cable trays.
Prior to November.14, 1984, weights of cables in trays were tracked but were not monitored.
(2)
Walkdowns conduct a walkdown of all supports for cable trays containing essential cables located at Plant Hat & Units 1 and 2.
%e r
following criteria were utilized in the walkdown.
t a.
Rapports were to be dosen by a tm team composed of f
engineers familiar with seismic analysis and cable tray support design.
b.
%e' most critical. supports were to be chosen based on the following criteria.
7
-l 1.
A large percentage of the total weight on the support is due to atta&ments other than cable trays.
j l
L May 13,1985 l
f Mesponse to 0.12 (continued)
Page 3 of 5 2.
We support configuration is different than typical.
details, indicating modifications at some point.
3.
Supports Which envelop difforent types of supports were to be chosen.
4.
Supports were to be chosen from different locations in the plant.
5.
%e total load on. the support is the greatest for that type of support.
6.
Other conditions existed Which appeared likely to mke the support more susceptible to a seismic event.
These conditions could incitrie missing bolts, gaps urvier base plates, attachments Which could cause torsion, etc.
7.
In addition, on tklit 2, no supports were to be chosen Whidt had previously been evaluated for the as-built condition.
' "I.
~
r Response to 0.12 (continued)
Page 4 of 5 3.
Analysis:
The supports chosen during the walkdown will be analyzed for all loads based on actual as-built information.
All FSAR requirements as notel in respcmso to question 3 heroin will he used for this analysis.
If any supports do not meet FrRt requirements, an ovaluation of that support against the operability criteria attached to lotter NFJ>04-641, datal December 28, 1904, will bo performed.
If the support cannot be shown to moot this critoria, the fact will be reported using normal reporting requironents.
In addition, any support which cannot be shown to meet FrW1 critoria will be upgraded to moet those critoria whether or not tho support mcots the operability critoria.
4.
INaluation:
An ovaluation will be conductot of the results of the analysis of all supports omlyzed for the as-built condition.
The results of this evalmtion will be unal to datormine the neal and critoria for additiomi walkdowns. The ovaluation will incitvlo nuch factors as the percent panstry, the percent not panning, tho exact reason cach support did not pann, tal the weak point in onch support which did not pass.
Mmy 13, 1905 -
7teepense to 0.12 (Continued)
Page 5 of 5 5.
Pbilem Walkdowns:
Pbilow-on Walkdowns will be conducted as required to choose additional supports for analysis.
The criteria for choosing additional supports will be determined based on the results of evaluating the previous analysis.
The process will be repeated until FTER coup 11ance can be demonstrated.
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% 13, 1985.- - -
0.13 Expand on the "more refined analytical techniques" referred to in GPC letter of Decc.i=r 28, 1964, for the evaluation of the cable tray supports.
Present the staff with information which supports your statement related to the adequacy of the analytical techniques used in your analysis.
Naponse Cable tray supports are analyzed using established design criteria arvi ITM cxymitments.
In some cases more refined analytical techniques could be used to demonstrate compliance with the ITM.
These more refinal techniques all eensist of standard analytical methods such as performing a time-history analysis, tnieling trays arx1 cupports toyether, arsi accountirvj for stiffnesa due to attached structural members.
m y 13, 1905 -
')
0.14 5'c January 16, 1985 letter on cable trays does not address' the connections utilised to secure.these cable tray systems to the structure. Please provide detailed information on the reevaluation of these andoring systems, and include considerations for the changes of loads and damping values referred in your submittal.
Response
Connections utilised to secure cable tray systems to the structure are
~ evaluated for all omble tray supports Whis are analysed.
As noted in the response to question 12 herein, the condition of the anchoring system is specifionlly included in the parameters for choosing the sapple of emble tray styports for the evaluation program.
e Ficor response spectra associated with the appropriate dasping levels (per the r
requirements of the attadment to letter NPD 85-031 dated January 16, 1985) will be used as irput to the analysis of emble tray support systems.
I Connections between the tray support and the structure will' be based on as-built condition and will he. analysed for all resulting inada.
Chiculated stresses are kept within raut requirements or the support is modified..m determine impact on plant safety, tray connections are analyzed per the requirements of the operability criteria attached to letter Nfp e5-641.
l P
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- . W 13, 1985.
y 9
0.15 Figure ~1-Plant Hatch Design Damping Curve, susuarises your position
- with regard to the allowable damiping values of the E. I. Hat & Nuclear Plant cable tray systems.
State the types and percentage of cable trays and supports that fall in the various segments of the curve.
Also, expand on the - justification for the criteria established in the same figure with. respect to fully IW,. partially loaded, and unicoded cable tray systems.
Response
Attached is a table whis provides a list of the number and percent of each type of cable tray support which falls in various segments of the allowable desping value curve.
For purposes of preparing this table, all. trays were ma= =ad to be at least 50% loaded.
For cases where trays are less than 50% -
loaded, the corresponding damping for the associated cable tray supports would be lower. Also atta&ed is a drawing of ea& of the eight typioni types of cable tray supports used at Plant Hatch.
Although 0 to 100% cable tray loading was considered in the test program, the earthquake simulation testing, which is most crucial in developing the response necessary to create the increased damping, was conducted. for 50 to-1004 cable tray loading. This loading range represents the more severe design requirement and would be expected to more accurately reflect most as-built conditions.
No detectable variation in damping for - the system was evident '
across this range (i.e. damping at 50% loading was 42% critical).
The possibility exists that these high damping ratios extend downward for even less loaded tray.
However, realizing that empty cable tray must meet FSAR commitments (for steel bolted structures) a linear interpolation was conservatively adopted between 0 and 50% losdh g.
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0.10-0.15 0.05-3.075 Reactor'13) 158; Dryweils Co2 trol 130, 147
,s 2,
0.15-0 20 0.07S}.10.
. Reactor 185; Diesel 146; tantrol 164 %
s
\\
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0.20-0.25 O.10-0.125 Reactor 203; Contml 180
-4
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4 0.25-0.30 0.125-0.15 Reactor, 228; in';ake 128 5
0.10 -*
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PLANT HATCH DESIGN DAMPING CURVE Zones Type Of 1
2 3
4 5
Support 240 15.5 284 18.3 25 1.6 24 1.t A
13 0.8 302 19.4 24 1.6 102 6.t B
16 1.0 C
50 3.2 383 24.6 52 3'. 3 11 0.1 D
1 0.1 23 1.5 1
0.1 4
0.;
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G H
319 20.6 992 63.8 102 6.6 141 9.1 1
Toaes ZPA (g)
Damping Ratio Locations 1
0.10-0.15 0.05-0.075 None 2
0.15-0.20 0.075-0.10 Reactor Bldg. 130; control Bldg. 130 3
0.20-0.25 0.10-0.125 Reactor Bids. 158; Drywell 165; Control Bldg. 147; Diesel 146 4
0.25-0.30 0.125-0.15 Reactor Bldg. 185; control Bldg. 164 5
- 0. 30 ---
0.15 Reactor Bldg. 203, 228; Control Bldg.180; Intake Struct 128
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PLANT HATCH DESIGN DAMPING CURVE l'
l l
Zones s
Type Of 1
2 3
4 5
Support
('
i 5
A
-138 1211(
138 12.1 19 1.7 35 3.1 B
150 13.2 15 1.3 C
50 4.4 D
377 33.2 21 1.8 70 6.2 32 2.8 E
42 3.7 17 1.5 8
0.7 i'
F' 19 1.7 t
G 1
0.1 H
4 0.4 781 68.8 191 16.7 97 8.6 67 5.9 Zonts ZPA (g)
Damping Ratio Locations I._ 3 0.10-0.15' O.05-0.075 Reactor 130, 158; Drywell 165; Control Bldg. 130, 147, i
168
\\.
2 0.15-0.20 0.075-0.10 Reactor Blds. 185, 203; Control. Bldg. 180
'3 O.20-0.25 0.10-0.125 Reactor Bldg. 228 4
' 0.25-0.30 0.125-0.15 Reactor Bldg. 256 5
0.30 +
'O.15 None k
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PLANT HATCH DESIGN DAMPING CURVE Zones Type of 1
2 3
4 5
Support
.I 8
0.7 19 1.7 111 9.8 192 16.1 A.
~
44 3.9 106 9.3 15 1.:
B 50 4.4 C
D 286 25.2 15 1.3 76 6.7 123 10.1 42 3.7 25 2.:
E 19 1.7 F-1 0.1 G
4 0.4 H
294 25.9 78 6.9 409 36.1 355
- 31. ;
i Zonas-ZPA (g)
Damping Ratio Locations
~1 0.10-0.15 0.05-0.075 None
'2 0.'15-0.20 0.075-0.10 Reactor Bldg. 130
-3 0.20-0.25 0.10-0.125 Control Bldg. 130 4
0.25-0.30 0.125-0.15 Reactor Bldg.158; Drywell'165; Control 147 5
0.30 -*
0.15 Reactor Bldg. 185, 203, 228, 256; control Bldg.
-164, 180 4
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y -~ q ~ w s- [ ] EDWIN I. -HNIUI NUCLEAR PLANP - UNITS 1 AND 2-REQUESP MR ADDITIONAL INMRMATION - SOIIS 1 - l'. - Prepare tables to provide numerical values actually used in ' seismic studies for parameters relating to~ soil foundation materials. A. Original FSAR Seismic' Analysis-Unit 1 <a.- Dynamic soil shear modulus, G b; Shear strain (%) c. Pange of_G varied. d.
- Poisson's ratio e.
soil damping (% of. critical) - geometric, material, and~ total design damping for the appropriate modes of motion.
- (
~ Provide the above information for all Unit 1 seismic Category-I
- structures.-
B. Provide a second table for. the recently completed seismic "~ reanalysis study for. Unit 1 structures. C. Provide similar tables (as A. _ and B. above) for. all ' Unit :. 2 ' structures _ for the original FSAR seismic analysis 'and for the seismic reanalysis. D. _ Identify FSAR or other. document references which provide the' basis . for the listed soil shear moduli and soil damping values given in - the four tables. 2. Compare -and discuss. foundation material inpu't listed in the tables and actually used in the original FSAR seismic: analyses with. specific design values used. in the seismic reanalysis ^ studies. Provide justification ~ for :less conservative values.~wherever used in reanalysis / redesign studies. Provide available references-to' support justification-1 _ discussions. e a J ^ _ ; g__. - my 13,: 1985-s
~ _~. -. - - - _. =. - _ - ,, ~ Response to 0;1-niclosure 2:- , Thel following - tables listed below ' identify the soil' parameters used in the-various DBE analyses performed for Plant Hatch: u
- A. ;
-Iktit'l'Beactor Building and Internals' Table 2A B. Unit 2 Beactor Building and Internals Table 2B C.. Cbntrol Building Table 2C
- l. ~ Unit 1 Analysis
- 2. -Unit 2 Analysis Table 2D.
D.' . Diesel' Generator Building 'l. Unit 1 Analysis Table 2E 2. Unit 2 Analysis . Table 2F l E;. Intake Structure 1. Unit 1 Analysis Table 2G
- 2. - Unit 2 Analysis Table 2H
'F. Main Stack Table 2I -Identified in each table are the following soil parameters: ' A. - Dynamic Shear M3dulus (including the range of values where applicable)< B. Shear' Strain C. Poisson's Batio l D. .Danping ' (see the response to 0.11 of ' Enclosure 1 for a discussion of -the. soil danping used in the Unit 2 seismic analyses) The parameters identified in the tables coincide ~with one of three types of ~ foundation and backfill materials. The first material type is e the natural ~ i soil deposit that supports both the Unit 1 and Unit 2 Peactor Building and Internals, - Control Building, Intake - Structure (foundation material only), and o- 'b s [.. L ayL13,?1985- ^~ M ~,,. _.. _. _,. - -.
l lbsponse to 0.1-Enclosure 2 (continued):- - Main ~ Stack - (pile : supported). The second material type is the backfill that-j supports the Diesel Generator Building.1 Finally, the third material type is a composite of the backfill and.K-Krete that surrounds a portion of the Intake Structure. r . thy 13, 1985 u., Y' 8 Y n 4 E-p TAEE 2A S w ry of Soil Parameters for.DEE Analyses of Unit 1 Recaor B2ilding and Internals Original-Devised Parameter Analysis Analysis Dynamic S$nar tedulus (ksf) 23,300(1) 23,300(1) Shear Strain'(%). '(2) '(2) Poisson's Hatio-0.42 0.42 Soil Dunping (% of critical)(3)- N-S and E-W Analyses (4). Translational and Botational 5.5 5.5 l I ["I e re ~, t t 4' u I 4 L - l - t
- h y 13,'1985-.
r-. e 4 .- Y {, + u. - i. ^ 5 L: ' TAILE 2B samunary 'of ' Soil Parameters for DBE Analyses of Unit 2 Reactor Building and Internals _. + _ ~ ' Original-Revised Parameter Analysis Analysis Dynamic shear m oulus'(ksf) 23,306(5) 23,300(5) ~ Shear Strain (%) -(2) (2) ~ Poisson's Ratio 0.42 0.42 l-l. Soil Damping (% of critical)(6)(8)(9) L, -N-S Analysis ~ Translational Radiation (Geometric) 36.5 34.2 t mterial 0.0 6.0 l Total 36.5 31.7(10) L - Ibtational- + Radiation (Geometric) 10.4 9.2 - mterial O.0 0.0 'Ibtal 10.4 9.2 / E-W Analysis Translational ~ Radiation (Geometric) 37.2 34.9 Material O.0 6.0 'Ibtal. 37.2 32.2(10) Ibtational Radiation-(Geometric)- 10.2 9.0 mterial 0.0 0.0 Total-10.2 9.0 ~ ' __ Vertical Analysis Translational j.: Radiation (Geometric) 63.3 59.3 i mterial-0.0 6.0 Total 63.3 50.5(10) a i. _Q a my 13,1985., _
p. TABLE 2C i' Stumnary of Soil Parameters for Unit 1 DBE Analyses of Control Building _ ' Original Revised Parameter Analysis Analysis Dynamic' Shear !bdulus (ksf) 23,300(5)- 23,300(5)- Shear Strain (%) (2) (2). [ Poisson's Ratio 0.42 0.42 Soil Damping (% of critical)(3) N-S and E-W Analyses (4) Translational and Ibtational 5.5 5.5 j t 4- ~ my 13, 1985 (b ' '. ' t .,( } s } E.' l- ' TAEE 2D - I (; ' Simenary of Soil Parameters for Unit 2 DBE' Analyses .of Control Building: Original. . Revised Parameter Analysis-Analysis-23,300(5) 23,300(5); 2 Dynamic Shear m dulusL(ksf)- shear Strain (%) (2) -(2)- l' Poisson's Batio 0.42 0.42-Soil Damping'(% of critiical)(7)(8) N-S Analysis-Translational-50.2 . Radiation (Geometric) -Material 6.0 c Total 5.0 43.7(10) i Rotational Radiation (C e tric) 30.0 Material 0.0 Total 5.0 30.0 E-W Analysis .Translational I-Radiation (Geometric) 49.0, 6.0 Sterial Total - 5.0 42.8(10)- Ibtational: Radiation (Geometric) 28.3 0.0 Material - Total' 5.0 28.3 1 Vertical Analysis 'Translational I. Radiation (Geometric) 84.5 6.0 L mterial-L Total 5.0 69.4(10) V,- t = L w - i-l 3; m y 13; 1985, (
t -- - y. , pw e,3.. .y-g _ 4 .W. .- a w -, -) ' , 3 1 e sq., w, mp A" N.w.. s a b,- L .;y 'h. \\ % g -- h. . TALE 2E x o, Stannary of. Soil Parameters for. Unit 1 DEE Analyses ; lof Diesel Generator Buildirrf J 6 c.. g s ."e. .l
- * + '
p. n $lgIhl( Nwised Parameter - Analysis ' .iAnalysis - Dynamic Shear Modulus l(ksf)(1) t 1 's + '1,522'1.s A h630 Upper Bound Value 'N 1,750 Average.Value Lower Bound Value' 144 700 7 Shear Strain (%) -(2). (2) Poisson's Batio 0.36 'O.36 x Soil Damping'(% of critical)(3) 'N-S and E-W Analyses (4) -e Translation and Ibtational 5.5 5.5 gr w s af w +
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s ~. t 4-2 TABLE'2F Sununary of Soil Parameters for Unit 2 DBE Analyses of Diesel Generator. Building. Original Revised ~ -Parameter ~ Analysis Analysis Dynamic Shear Ptdulus (ksf)(11) 2,520 Upper Bound Value 2,520 ~ Average Value 1,750 1,750 Lower Bound Value 1,120 1,120 'tf Shear Strain (%)' (2) (2) ~
- Poisson's Ratio 0.36 0.36 Soil. Damping (% of critical)(7)(12).
_,l .'N-S Analysis- "M, Translational Radiation (Geometric) 82.6 . Material 6.0 2 'Ibtal 5.0 20.0. NY: Ibtational Radiation (Geometric) 64.7 ~ Material 6.0 Total 5.0 20.0 e E-W Analysis Translational Padiation (Geometric). 82.3 6.0 mterial -Total' 5.0 20.0 ~.% t Ibtational' Padiation (Geometric) 74.3 6.0 Material. Total-5.0 '20.0
- Vertical Analysis dy ;, '
Translational 133.7 Radiation (Geometric) e f; ~,
- f 6.0
'y-Material Total. 5.0 20.0 50 p. r si .m l {.N *. ~ 79 2 .x. o>- Q my;13,- 1985 -- ', ' .. - _ ~. =
.y' 7.. yg; k N.~ n-3. y-P as - -sa.- y: .- (- % 3 i y Tams 20 - i}:(,.;g ' (W Summary of'. Soil Parameters for-Unit 1 DBE Analyses p;e Q
- f., l, of-Intake Structure
-g Original . Revised Parameter' Analysis -Analysis ~' Dynamic Shear Modulus (ksf) Ebundation Soil 23,300(5) 23;300(5) . Side' Soil (13) 2,700(14) Shear-Strain (%) Foundation Soil (2) (2)- { Side Soil (13)- (2) Poisson's Patio. Foundation Soil 0.42 0.42 = Side Soil. (13) 0.34 . Soil Damping (% of critical)(3) Foundation and Side Soil N-S and E-W Analyses (4) 5.5 Translational and Potational 5.5 {im c - t f5 W r t i i -~ s, t .l ' May 13, 1985-r~
7-s. r.1 a , + c ' TAILE 2H. Sununary of Soil Parameters for Unit 2 DBE Analyses of Intake Structure Parameter Analysis Analysis Original Revised. ~ Dynamic Shear Modulus (ksf) Foundation: Soil 23,300(5) 23,300(5)- Side Soil (13) 2,700(14) Shear Strain (%)- Foundation Soil' (2) (2) ' Side solli (13) (2) 7 -Poisson's Batio Foundation' Soil 0.42 -0.42 Side Soil' (13) 0.34-Soil' Damping (% of critical)(7)(8)(13) Foundation and Side Soil' N-S Analysis Translational 39.0 Badiation (Geometric) 6.0~ _ Material 2 Total 5.0 .35;3(10). 4 ~- Ibtational 15.9 Radiation (Geometric) 0.0 ~ . Material Total' 5.0 15.9
- E-W Analysis.'
Translational 35.4 Radiation (Geometric) mterial-6.0 Total 5.0 32.5(10) lbtational - Radiation _(Geometric) 7.8 Material 0.0 -Total-5.0 7.8 -Vertical Analysis Translational 60.0 -Padiation (Gecunetric) 6.0 m terial Total 5.0 51.0(10). my 13,1985._ mm = r .' TABLE 2I Sununary of Soil Parameters for DEE Analyses of~ min Stack (15) u. Original Revised Parameter Analysis Analysis Dynamic Shear Modulus (ksf) .29;100(16) (17) Upper Bound Value Average value 23,300(5) (17) Iower Bound Value '17;500(16) (17) Shear Strain'(%) (2) (2) Poisson's Ratio 0.42 0.42 Soil Damping (% of critical)(7)(8) N-S and FAT Analyses -Translational 38.5-Radiation (Geometric) 6.0 Nterial- ' Total 5.0 ~. 34.9(10) Ibtational Radiation (Geometric) 0.8 mterial ~ 0.0 -Total 5.0 0.8
- Vertical: Analysis
.Translational ~ 66.1 Radiation (Gecznetric) mterial 6.0 Total. 5.0 55.6(10) 7 'c A 4 N . m y 13; 1985; ;
- M, a; f
E r u - g, l a ,qey NOTES TO TAEES 2A 'IEROUGH 2I. Notes ~ (1)' From Table-12.6-3 of the mit 1 FSAR. ~ ~ _ (2). %e material' ' value used corresponds to an estimated. shear strain of 10-2 10 percent. -(3 ) ' ; From Table 12.3-2 of the Unit 1 FSAR.
- (4),
No Unit 1 vertical analyses were performed (see Section 12.3.3.2 of the ~ Unit 1 FSAR). s [(5') ' Same 'as used. in the analysis of the Unit l' Beactor Building.and Internals. See Table 12.6-3 of the Unit 1 FSAR.
- (6)
In'the original-analysis, only radiation damping was considered. Table 3.7A-2 of the Unit 2 FSAR was used to calculate the appropriate damping . values. We.5.0 percent damping value' specified for the original analysis was --(7). obtained from Table 3.7A-1 of the Unit 2 FSAR. ^ L(8) We translational damping used in the revised analysis was. to be equal to the -lesser of 100 percent of the radiation. damping calculated using Table 3.7A-2 of. the Unit 2 FSAR or 75 percent of the radiation damping plus an allowance for material danping_(6 percent). For this analysis,. ~ .75 percent of the radiation damping plus material damping ' governed _ and .-was therefore used.- The calculation of the _ rotational damping used in - the_ revised analysis considered ' only radiation : damping. Table 3.7A-2 of : the Uhit-'2 FSAR : was :used to calculate the appropriate damping ' values. %e methodology defined ' above : was established _ by the "SSRP-Guidelines for SEP Soil-Structure Interaction Review." /
- (9)-
The radiation darping values for the original and revised analyses differ due to' the fact that while an assumed soil density was used in the original calculations, a different, soil density, based on field data, was used in the revised calculations. ~ -(10) Blual to 75 percent of the radiation damping plus 6.0 percent (material ~ damping). z 4 May 13,1985 u
. o 7 4 a i. r s t d Notes to Tables 2A % rough 2I. Page - 2 of 2 2--- -(ll) ' We 1 Diesel l Generator Building. is founded ' at - grade, and. there ' is a degree of: uncertainty to its soil shear modulus that does not exist for-4: - the -. other Seismic Category 1 J structures 1which are. founded in-deeper - ? soil. Accordingly, the; soil -shear modulus ' values were_ varied to' account. for i this uncertainty. %e ;originalz Unit' 2 analysis of the Diesel' Generator Building reflected a sore recentEestimate of the mean ~ ,Gg and 'its. variation, 'and this mean value was used for ' both Unit,1 ?and Unit:2 revised analyses. All available soil data was reviewed-to - < ensure the mean value used in the revised analysis was indeed agr up iate. ;%e upper.and ' lower bound values identified for the-p revised Unit lj analysis / meet or. exceed the: ccustitment made in Section 12.6.2.1 of thel m it.1 FSAR. %e upper' and lower bound -values -identified for the original and revised Unit 2 analyses correspond to the best estimate of the variation in' the. shear modulus. } -(12): Toi account for possible impedance mismatch, the criteria used to
- calculate soil-damping in the revised Unit 2 Diesel Generator Building -
analysis was: more conservative than the - criteria used -in-the other i 1 revised Unit 2 analyses. We translational. and rotational damping used . as' to be 1 equal to the lesser of 20 percent of critical or 30 percent w of the radiation' damping calculated using Table 3.7A-2 'of ' the Unit 2 FSAR plus an-. allowance for material danping (6 percent); - As can be - a' -.seen from the data in ' the table, ' 20 percent damping -governed and was T ll therefore used. -(13) We effect of the side soil was not accounted for in the original -analysis. . (14) ' side soil. shear modulus is a composite value which accounts for the i presence of sand backfill and K-Krete that surround a portion.'of the structure. (15) Although - the - min Stack is shared by.both units, -only the Unit 2 L seismic criteria was employed in the analysis of this structure. W e upper and. lower bound values correspond to a + 25 percent variation J(16) . in'the average value.- ~ We min Stack is supported on driven steel H piles. Since driven ~(17)' piles increase the stiffness of the soil, two bounding analyses were performed. While one analysis used the average soil shear ~ modulus. value of 23,300 ksf, the other analysis assumed the soil to be infinitely rigid ^ (i.e., a fixed based analysis was performed). t x J m tp j my 13,1985,
Response ~to Question 2: There is only one case where the treatment of soil parameters in the' revised analyses can be considered to ' be' less conservative than in the original analyses. It has.to do with the fact that the -effective soil damping _ used in the following revised = analyses is higher than the damping used in the corresponding c,riginal analyses: A. Unit 2 analysis of Control Building. B. Unit 2 analysis of Diesel Generator Building. C. Unit 2 analysis of Intake Structure. ~ D. Analysis of Main Stack. The increased damping resulted from the use of a procedure that is at least as conservative as the procedure defined in Section 3.7A.1.6 and Table 3.7A-2 of the Unit 2 ' FSAR. As such, its use does not constitute a departure from previous licensing cxmunitments. May 13, 1985 a: 4.. ry- } ;? n. -Enclosure 3 EIMIN I; 192CH NUCLEAR PIANP - INITS 1 AND 2 - 4 IEOtHSP FOR ADDITIONAL INF0FMPTION - PIPING 4: . Thel licensee's use. of. Code, Case N-411' damping values in. pipingLseismic ~ analysis as an alternative to Img. Guide 1.61.. damping values is acceptable to - 1 the staff....The licensee also stated that it. intends to use the Code Case f N-411 damping. values for 'new.or_ replacement piping systems and load 3
- reconciliation work.at 'the Hatch Plant; The licensee should. provideL a commitment to do the following:
- (1). If; as a result lof using the A9E Code Case N-411 damping values, ' ~ zpiping. supports. are moved; modifiedL or-eliminated, 'the. expected ' increased piping ~ displacements -due to greater piping flexibility will be -checked to assure that - they can be acocamodated and that there will be no adverse-interaction 'with adjacent structures,- u.mp,Ents. and . equipment. c:(2)? .The licensee will :not use Code Case N-411 damping values for time
- history analysis' and. will use them only for seismic response sgb.um
.g analysis-. ' Response:- 3 Georgia Power Company agrees that the two commitments listed in-Enclosure 3 of the letter from Mr. John F. Stolz to Mr. J. T. Beckham, Jr., dated April 2;
- 1985, are : appropriate. Georgia Power intends to take all necessary steps to ensure both ocamitments are met'and reiterates our-intention'to use ASME Code 7
- Case N-411 only on new or replacement piping systems and' load reconciliation work.
a. k v A. ?May13,.Li985 -
Wg' g 3.T ENCWSURE 4 EDNIN I.1930I NUCLEAR PIANP-UNITS 1 AND 2 - ~ RBOUESP FOR ADDITIONAL INFORETION - EQUIPMENP OUALIFICATION 'The Equipnent; Qualification Branch ;(EDB)' has reviewed the Licensee's Part - 21 a evaluation 1 report, regarding. the - subject discrepancies ' in seismic analysis, ' submitted with ' its '. letter of. Der 14, 1984. Based - on the information provided, z the following staff ' comments.will need to be. responded :to by the
- licensee before a final staff evaluation can be. performed.
(1)5
- In Figure. 4 of the ~ above. report, the peak. - acceleration - of the new
~ ~ floor ' response spectrum '(FPS) exceeds the original one by about ' 50%.
- The iapplicant should provide ' more detailed information. of ' seismic qualification for the six equipment types (see; Table 6 of the report) which are located -in Control Building at floor elevation '112. ft. and
-are governed by Figurei 4 in their. qualification. 'Ihe. specific
- information which may be of interest includes, but is not limited to,
..the original methods of : qualification and the basis of concluding seismic qualification against the new FRS. In ' addition, - the staff may, -in' the future, elect to audit the corresponding' equipment qualification documentation. f (2)' on page ~16 of the above.: report,. it is stated that insufficient
- information pw.luded a decision regarding the impact of - the new: FPS on ' seismic qualification of L10 equipnent items.
In order -'for the staff to concur that. the equipnent seismic qualification was indeed not adversely affected by the new FRS, however, more supporting z. information.will need to be established to confirm the qaalification. Otherwise,~ 'a requalification will have to be performed' and approved by the staff., -Response to Question-(1): JAs noted by the Equipment Qualified Branch (EQB), the peak of the new Floor
- Response Spectra (FRS) at elevation 112 in the control building exceed the original"FRS by about 50%. However, it must be,. emphasized that the seismic environment at this building level is quite low, with - a peak acceleration of m
~ -1ess than 0.75g:and a ZPA of less than 0.2g. s w s w o m y 13, 1985-w 4 EDWIN I.19ftCH NUCLEAR PLANP-INITS 1 AND 2 REQUESP RR' ADDITIONAL INNTION - BOUIP1ENT OUALIFICATION a Page _2 of _4 ^: u %e qualifi' cation details of interest for the'six equipment types located on N this floor (Ref ~ ~ Table 6 of the Part 21 evaluation report - NED-84-622) are as follows. s The ! station. battery room exhaust fans (MPL -. #1Z41-C014,15) were originally analyzed to seismic levels which exceed the new FRS at elevation 112. This is -due;;in_ part;. to the fact that fans identical to these are located elsewhere-in ' the plant and subject-to higher postulated seismic levels. The station ~ batiteries (NL #1R42-5001A, B) were originally tested to levels in excess of - the new FRS.- ' Two of the valve types (MPL # IP52-F102A, B & IP52-F2Ol) were assessed 'in a manner consistent with the findings of the FPS comparison. for - piping systems (Ref: - Section II.C of the report). Specifically, the original-one percent damped FRS at this ' elevation was found to essentially envelope the new FRS at the PVEC damping (see the attached comparisons).. It was concluded,. u therefore, ' that the seismic levels at the valves' location would be less severe. The remaining valve (MPL #1C71-N003A, B) is a simple root valve with no extended parts. It is a - manual, passive valve and standard. ASME proportioning was used to-insure pressure, boundary integrity.
- This qualification then, is insensitive to the changes in the FRS.
Finally, lthe exhaust. flow switch (WL # 1Z41-N019) is under further review and is addressed ?in our response-to'Ouestion 2. 1" May 13; 1985 '
V EDWIN I. HNICH NUCLEAR PLANT-INITS 1 AND 2 PEQUESP FDR ADDITIOlAL INFOBl%TIG7 - FOUIPMENT OUALIFICATIO7 Page 3 of 4 Response t'o Question (2): The. ten equipment items identified on Page 16 of the Part 21 Evaluation Peport are the exhaust battery room fan flow switch (MPL #1Z41-N019), the Control Bililding supply fan discharge flow switches (MPL # lZ41-N016A, B, C), the Control Building exhaust fan discharge flow switches (MPL #1A41-N017A, B, C) and the neutral. grounding resistors (MPL # 1R34-S004A, B, C). These equipment represent the few cases in which insufficient information at the time of our evaluation submittal precluded a decision regarding the impact of the new FPS on seismic qualification. As ~ stated in the report, however, it was felt that the equipment sample remained sufficiently large to evaluate safely. Although it was determined that. evaluation of these two equipment types was not. necessary to evaluate overall plant safety, it has always_ been our intention to pursue the information necessary to assess these equipment. Earlier. this year, an effort was initiated to address these items. To date, nine of the ten items have been resolved. The assessment of the three neutral grounding resistors was resolved by the use of the Unit 2 qualification report for the identical component. The Unit 2 qualification test levels exceed the new thit 1 required seismic levels. The increase in the Unit 1 FPS, therefore, do not affect the seismic qualification. The six control building supply and exhaust fans discharge flow switches have been determined to have
- May 13, 1985 _
7.-: EDWIN I. EWICH NtDEAR PLANP-IDTITS 1 AND 2 moutsP FOR~ ADDITIONAL IlmTION - EQUIPtB7P QUALIFICATION Page 4 of 4 no safety = related function, therefore further assessment' of the seismic qualification was deemed to be unnecessary. The evaluation' efforts for the battery room exhaust Ifan flow switch are still in progress. These efforts include vendor contact; pursuit of qualification levels of similar components, and researching qualification programs of other Nuclear Plants. This is a-time consuming process, and a definitive conclusion has ' not been ' reached as of this time. Georgia Power Company, however, is ~ committed to the eventual resolution of this qualification item. ~ m y.13,'1985 ..s .. o..m...... C 4 C vCL 6 8. n s u Div.tsout.Is PED speceo Calc. tb stic -uq -on 3 Pne pm d. by e M.cd d -AN $ kiwed by S' % d'G, Tb4l90 y sy p w 3. ac e 4 0 0* m e i. 6 A 4 O O. _... m. - ..i.. s. .a 2 .a a o
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