ML20006F309
| ML20006F309 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 02/15/1990 |
| From: | Hayes J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| IEB-79-02, IEB-79-14, IEB-79-2, NUDOCS 9002270374 | |
| Download: ML20006F309 (35) | |
Text
[oacy*)g UNITE D STATES g;M; 3m g
NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D, C. 2055$
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February 15, 1990
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Cocket 50-395 LICENSEE: South Carolina Electric & Gas Company FACILITY: V. C. Succer Nuclear Station, l' nit No.1
SUBJECT:
SUMMARY
OF FEBRUARY 8,1990 NEETING WITH SOUTH CAROLINA ELECTRIC
& GAS COMPANY ON INSPECTIOkS COVERING BULLLTINS 79-C2 AND 79-14 GENERAL On February 6,1990, representatives of the Office of Nuclear Reactor Regulation and Region Il met with representatives of South Caroline Electric & Gas Company (SCE8G) and their consultants to discuss certairi issues dssociated with inspection conducted November 27 thrcugh December 1,1989 arid Decenber 11 through 15, 1989 at the V. C. Sumer Nuclear Station, Unit No. I with respect to the implementation of Inspection and Enforcement Bulletins 79-02, " Pipe Support Baseplate Designs Using Expansion Anchor Colts," and 79-14. " Seismic Analysis for As-Built l
Safety-Related Piping Systems". The nweting was held at the NRR offices l
in Rockville, Maryland. A list of those perscns who attended the meeting is included as Enclosure 1.
DISCUSSION 4
In Inspection Report 50-395/89-200 four issues were identified which i
required additional review by the NRC.
These issres were:
1.
non-uniform consideration of zero period acceleration (ZPA) at the Sunmer Station; 2.
exclusion of seismic anchor rovements (SAM) less than 1/8 inch without any quantitative technical basis; 3.
exclusion of containment penetration novements in the piping afialysis for the effects of post-accident pressurization or stea@ state temperature growth; and 4.
utilization of a potentially nonconservative piping decoupling criterion.
i A handout was provided at the meeting which presented SCE8G's position with respect to the above issues.
This handout is included as Enclosure 2.
As a result of the meeting the staff concluded that SCE&G needed to do the following to assist the staff in evaluating the four issues.
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l 9002270374 15 l
PDR ADOCK 0 395 G
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1.
With respect to ZPA and the decoupling criterion, SCE8G should
{
formalize the work which was presented at the meeting.
The i
bases for the test case should be enumerated along with the factors in selecting the test case.
The details of the analyses should be presented.
2.
With respect to SAM, SCE&G should evaluate the effects of SAM when piping runs inside a building.
The design guidance docunents should contain criteria for SAM inside of a building.
3.
For containment movenent, tie Sumer criteria of nonadditive displacenent as a result of pressure and thernel effects should be justified.
4.
Sumer's FSAR consnitment with respect to the utilization of the square root of the sum of the squares for cont >1ning displacenent i
of two different buildings should be provided.
Original Signed By:
John J. Hayes, Jr., Project Manager Project Directorate 11-1 Division of Reactor Projects I/II Office of Mclear Reactor Regulation
Enclosures:
As stated cc w/encls:
See next page DISTRIBUTION see attached page l
l OFC :PR
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..... :.k.,.....:.sw HAME IJ DATE :2//j/90
- 2//f790 OFFICIAL RECORD COPY Docunent. Name: 2/8/90 neeting summary
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h DISTRIBUTION FOR MEETING 'SUMERY DATED:
February 15, 1990 Facility:
Summer l
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NRC PDR Local PDR T. Murley 12-G-18 J. Sniezek 12-G-18 E. Adensam 14-B-20 P. Anderson 14-B-20 j-J. Hayes 14-B-20 f-000 -
15-B-18 E. Jorden NNBB-3302 F. Centre 11 RII O. Hehl C. Julian RII P. Kuo 9-H-3 W. Lanning -
9-A-1 4
A. Lee 9-H-3 L. Modenos RII R. Parkhill 9-A-1 E.'Tourigny ACRS (10)
P-315 i
B.'. Borchardt 17 ~G-21 Susurer File l
- Copies sent persons on facility service list c
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1 LIST 0F ATTENDEES
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NRC SCEM SCE&G C0ftSULTANTS i
F. Cantrell A. Barth C. Chen J. Hayes
- 0. Bradham K. Chu C. Hehl A. Koon A. Hoffert
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C. Julian D. Moore D. Landers P. Kuo K. Nettles f
W. Lanning A. Lee L. Modenos R. Parkhill l
E. Tourigny j
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i SCE&G PRESENTATION on i
f INSPECTION REPORT 89-200 l
"lEB 79-02/14" 1
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FEBRUARY 8,1990 i
l ROCKVILLE, MARYLAND 1
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SCE&G ATTENDEES Ollie Bradham Vice President Nuclear Plant Operations Ken Nettles General Manager Nuclear Safety Dave Moore General Manager Engineering Services AlKoon Manager Nuclear Licensing Andy Barth SCE&G Design Enginevrbg i
Fred Hoffert Consultant Gilbert / Commonwealth l
Chang Chen l
Consultant Gilbert / Commonwealth i
Don Landers
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Consultant Teledyne K.Y.Chu Consultant Stone & Webster l-
. _. _ _ _... ~
j o-NRC MEETING t
IEB 79-02 and 79-14
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FEBRUARY 8,1990 AGENDA i
i 1.
Introduction O. S. Bradham
!.icensing issues K. W. Nettles s
. Ill. Technicalissues De R. Moore 9
IV. Summary O. S. Bradham d
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NRC 79 79-02 MEETING February 8,1990 1
PRESENTATION EMPHASIS INSPECTION TEAM CONCLUDED:
e SCE&G met the intent of 79-14 and 79 02.
Identified deficiencies raise no i
significant safety concern.
1 HISTORICAL COMPUANCE
SUMMARY
L.
REGARDING ISSUES RAISED DURING I
INSPECTION.
i DISCUSSION OF FOUR GENERIC ISSUES IDENTIFIED DURING INSPECTION.
s
SUMMARY
OF CONCERNS AND PLANNED ACTIONS.
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I LICENSINGISSUES
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I IEB 79-14 History e
VCSNS Compliance with IEB 79-14 e
i Licensing Basis for Piping Analyses e
l Current Licensing issues o
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O IEB 79-14 HISTORY
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lEB 79-14 issued July 2,1979 to address the seismic analysis of as-built safety-related piping systems Revision 1 issued July 18,1979 to clarify the scope-o of piping systems affected (i.e., NSR piping = 2t" and seismic Cat i of all sizes L
if computer analyzed)
Supplement 1 issued August 15,1979 providing additional guidance and definition for l
licensee action en inspection; on L
nonconformances; and on QA requirements Supplement 2 issued September 7,1979 providing additional gu! dance on inspection; on nonconformances; and schedule.
3 Additionally listed specific differences between design and as-built conditions I
at specific nuclear power plants (Ref:
Appendix "A" to Supplement 2)
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VCSNS COMPLIANCE Warn IEB 79-14 REQUIREMENTS L
l t
e 1980 - 1982 Inspections and Re-Analysis of All Piping Systems Used EDS, TES, gal and W Technical Management by gal Used Different Models and Modeling Techniques Effective Design Control Estimated $20 M Cost e
IEB 79-14 and IEB 79-02 Closed e
August 1983 During)the Period 2/80 - 8/83Nine(
e
)
e Resolved Overlap issue
i PROCESS INITIAL PIPING ANALYSIS SUPPORTS DESIGNED & INSTALLED t
1r 1r ALL INPUTS VERIFIED u
100% AS BUILTS Design specs o
OBTAINED Valve weights e
e SAMs Proper spectra o
e Jet loadings I
3r AS.RUILTPlPING ANALYSIS P
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SUPPORTS EVALUATED FOR LATEST 8
LOADS AND 79 02 EFFECTS 1
l-i r MODS ISSUED TO MEET DESIGN REQUIREMENTS 7
DOCUMENTATION RECONCILED - ALL LICENSING COMMITMENTS MET
l 4
PIPING ANALYSIS LICENSING BASIS FSAR e
ASME Code Compliance e
NRC Reg Guides 1.29,1.48,1.61,1.84 & 1.85 e
SER Section 3.9.1 i
Independent confirmatory analysis by Battelle Northwest Paci Lab Demonstrated compliance with ASME code allowables Confirmation of ability to use computer models Supplement 4 Section 3.7.4
)
system including computer analysis verificationLic Supplement 5 Section 3.7.4 i
Final report submitted by SWEC Report addressed:
Field walkdown for as built verification e
Independent stress analysis and evaluation i
Design Control Audit e
Subsystems analyzed were originally analyzed by TES tiesults:
Minor differences in analytical results due to modeling e
No generic ramifications e
No hardware changes e
Used commonly accepted industry practices e
)
Seismic requirements as stated in design criteria met e
Supplement 5 Section 17.5 SWEC Audit of gal design control and interface control with TES Results:
Overall program adequate
CURRENT LICENSINGISSUES Lack of prescribed criteria Use of industry practices for license timeframe versus current industry l
j practices
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No Significant-Safety Concerns q
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INDUSTRYSURVEY 1
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e ZPA VCSNS PIPING ANALYSIS 8 PLANTS DID NOT CO 5 PLANTS DID CONSIDER ZPA IN THEIR O i
ANALYSIS e
DECOUPLING 1 PLANTUSED 40% MOMENTOFINERTIA i
3 PLANTS USED 25% MOMENTOFINERTIA VCSNS 2 PLANTS USED 15% MOMENTOFINERTIA 3 PLANTS USED 10% MOMENT OFINERTIA 1 PLANTUSED 7% MOMENTOFINERTIA 1 PLANT USED 10% SECTION MODULUS 2 PLANTS CRITERIA UNIDENTIFIED e CONTAINMENTTHERMAL GROWTH VCSNS 7 Of 7 PLANTS OF OUR CONTAINMENTTYPE CONSIDER STEADY STATE THERMAL GROWTH OF CONTAINMENT e CONTAINMENTLOCA PRESSURE 6 OF 7 PLANTS OF OUR CONTAINMENT TYPE CONSIDER LOCA PRESSURE GROWTH OF CO i
SAMs THRESHOLD VCSNS 2 PLANTS ANALYZED SAMs > 1/8"IN THEIR PIPIN ANALYSIS f
S PLANTS ANALYZED SAMs >1/16" IN THEIR PIP ANALYSIS 4 PLANTS IDENTIFIED SAMs AS BEING INSIG DO NOT CONSIDER THEM IN THEIR PIPING AN 1 PLANT ANALYZED ALL SAMs IN THElR PIPIN PIPING ANALYSIS 1 PLANT COULD NOTIDE e
SAMS BETWEEN BUILDINGS VCSNS j
- 7 PLANTS USED SRSS BETWEEN BUILDINGS 1 PLANT USED ASUM BETWEEN BUILDINGS 3 PLANTS USED A COMBINATION OF SRSS AND A BETWEEN BUILDINGS 1 PLANTIDENTIFIED SAMs BETWEEN BUILDINGS APPLICABLE 1 PLANTS COULD NOTIDENTIFY THEIR SAMs BETW BUILDINGS CRITERIA
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i TECHNICAL ISSUES l
POSITIVE ATTRIBUTES OF PIPING e
DESIGN e
ZPA 1
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e DECOUPLING h
SEISMIC ANCHOR MOVEMENTS e
i s
e CONTAINMENT MOVEMENT INDEPENDENT SEISMIC DESIGN e
VERIFICATION SCE&G PRACTICES e
POSITIVE ATTRIBUTES OFPIPING DESIGN PIPING ASeBUILTS AND ANALYSES HAVE CLO e
CORRELATION DOCUMENTATION 15 VERY THOROUGH AND E e
RETRIEVABLE DRILLED ANCHORS FOR PIPE SUPPORTS WE e
INSPECTED ECCENTRIC MASS EFFECTS OF VALVE ACTUA e
CONSIDERED SPRING CAN AND SNUBBER SETTINGS WERE e
VERIFIED INHERENT CONSERVATISMS e
- REG. GUIDE 1.61 SPECTRA DAMPING VALUES USED
- SUPPORT GAPS NOT CONSIDERED - PROVIDES RELIEF FOR SAMs PROCEDURES IN PLACE TO ASS!!RE CONTINUED e
COMPLIANCE
' STRONG SCE&G MANAGEMENT COMMITMENT e-MAINTAIN A WELL DOCUMENTED PIPING PROG I
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gpp TEST CASE EF 02 L
STRESS INCREASES OCCURRED IN LOW STRESS e
SYSTEM i
1507 PSI HIGHEST SEISMIC STRESS IN AREAS GOVERNED L
SUPPORT LOADS INCREASED SIGNIFICANTLY IN ZPA AN e
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HOWEVER, THE LARGE CHANGES WERE FOUND IN THE LIG LOADED SUPPORTS.
CHECK OF SUPPORT DESIGN CALCULATIONS ~ F e
LOADS SHOWED THAT ALL LOADS WOULD BE ACC WITHOUT ANY MODIFICATIONS - 7 (OF 7) $NUB
- 20) RESTRAINTS HAD LOAD INCREASES.
IMPLICATION OF RESULTS STRESS INCREASES APPEAR TO BE IN THE LOW REGIONS OF THE PIPING. HIGHEST SUBSYSTEM S NOTCHANGE.
SUPPORT LOADS APPEAR TO BE MOST SIGNIFICA LIGHTLY LOADED SUPPORTS AND 010 NOT OVER 1
SUPPORTS ON THE TEST CASE DID NOT INVALIDATE THE TEST CASE PERFORMED 1982
):
-- e ZPA WAS NOT A CRITERIA AT SUM
- STANDARD IN 1982.
ANALYZED HAVE ADDITIONAL CON CONSIDERED ZPA DISCUSSION ON ZPA INTRODUCED BY e
FREQUENCY RESPONSE.1984. DISCUS NUCLEAR PIPING OVERESTIMATE e
CURRENT RESPONSE SPECTRA ARE CONSERVATIV e
REG. GUIDE 1.61 DAMPING VALUES USED IN WHICH ARE HIGHERLIEU OF CURRENTLY ACC l
WHEN COMPARED TO TIME HISTORY e
CONSIDERED FOR PIPING ANALYSI e-s VCSNS 8-PLANTS DID NOT CONSIDER ZPA IN T L
ANALYSIS S PLANTS DID CONSIDER ZPA IN THEIR O CONCLUSION L
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AS BUILT PROGRAM AT V.C. SUM L
SYSTEMS.
METHODS WITH REDUCED CONSERVAT
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CHANGE WOULD BE EXPECTED.
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INDUSTRY STUDIES SHOWING THAT CRITERIA USED TO DESIGN 4
NUCLEAR PIPING OVER ESTIMATE SEISMIC RESPONSE e
The EPRl/USNRC Piping and Fitting Dynamic Reliability Program concludes that current C rules based on static collapse for dynamic load considerations are overly conservative (
Taggart,5.W., et. al.," Seismic Analysis and Testing of Piping Systems and Components" PVP Vol.144 Seismic Enaineerina 1988, ASME PVP Conference, Pittsburgh, June,1988 p. 229 e
Tests at the Heissdampfreaktor test facility in W. Germany suggest that the highly restraine piping design typical of a U.S. plant is excessively conservative. (ref. Matcher, L., Sc
- D., and Steinhiller. H., "High Level Seismic Tests of a Piping System at the HDR Facilit Vol.182, Seismic Enaineerina 1989, Desian, Analysis, Testina, and Qualification Methods ASME/J5ME PVP Conference, Honolulu July 1989, p. 231237),
NUREG 1061, Vol. 2 states that piping in power plants subject to severe earthquakes has no e
failed under inertia loading. It further states that the methods, procedures, and acceptan criteria currently used to design nuclear power plant piping greatly overestimate the seismi response of piping, e
EPRI Report NP 5617, Vol.1,( by EQE,Inc.) states the following:
"The primary conclusion reached during the course of this study is that failures of welded steel piping have not been observed as a result of piping inertialloads. All piping has in fact exhibited a very high degree of resistance to failure during earthquakes up to 0.9.g peak ground acceleration."
DECOUPLING e
SMALLER PERCENTAGE COULD HAVE PROBLEMS IN THE AREA OF TECHNOLOGY OF THE TIME COMPUTER e
INDUSTRY SURVEY SHOWS THAT TH FOR DECOUPLING FOR PIPING ANA 1982 VARIES BUT SUMMER STATION IS CONSERVATIVE 1 PLANT USED 40% MOMENT OFINERTIA 3 PLANTS USED 25% MOMENT OFINERT VCSNS 2 PLANTS USED 15% MOMENT OFINERT 3 PLANTS USED 10% MOMENT OFINERT 1 PLANTUSED 7% MOMENTOFINERTIA 1 PLANT USED 10% SECTION MODULUS 2 PLANTS CRITERIA UNIDENTIFIED
DECOUPLING
^
V.C. SUMMER CRITERIAiMPOSED ADDITIONA CONSERVATISM TO DECOUPLING CRITERI OTHER REQUIREMENTS-RUN LINE ACCELERATIONS MUST BE LIMIT 3g HORIZONTAL RESULTANT ACCELERATION 2g VERTICAL ACCELERATION LUMP MASS MUST BE ADDED TO THE R e
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BRANCH LOCATION-l j
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where:
ir ight of Branch Pipe + Contents
=
(
L = 10 x 0.D. of Branch Pipe THERMAL MOVEMENTS OFTHE RUN PIPE e
IN ANALYSIS OF THE BRANCH PlPE e EFFECTS OF CRITERIA ON THE RUN PIPE (LUMP MASS)
- ADDS STRESS IN DEAD WElGHT AND SEISMIC ANALYS e EFFECTS OF CRITERIA ON BRANCH LINE
-DYNAMIC INPUT AT RUN PIPE CONNECTION CONSIDER
- ACCELERATION LIMITED AT BRANCH RUN INTERFACE
-SUMMER STATION PIPING IS GENERAL SMALL 4,
l
.4 DECOUPLING TEST CASE r
REVIEWED P&lDs TO IDENTIFY CONTROLLING e
DECOUPLED RATIO i
L e
ANALYZED REAL LOCATION WITH 3 INCH D BRANCH LINE DECOUPLED FROM A 6 INCH i
DIAMETER RUN LINE IBRANCH
,ggo-IRUN NO OTHER DECOUPLED BRANCH LINES WERE-e FOUND TO BE NEAR THE 15% THRESHOLD.
RESULTS - DECOUPLED VS. SINGLE MODEL '
LESS THAN 10% STRESS INCREASE 3.8% INCREASE IN DEAD WEIGHT STRESS 7.5% INCREASE IN DYNAMIC STRESS LESS THAN 10% SUPPORT LOAD INCREASE L
THE BRANCH AND RUN LINE HAD SIMILAR FREQUENCIES; THEREFORE, THIS MODEL WILL REPRESENT A WORST CASE CONDITION FOR T SIZES INVOLVED.
CONCLUSION:
L BASED ON INDUSTRY PRACTICE AT THE TIME OF ANALYSIS, AVAILABLE COMPUTER TECHNOLOGY, CRITERIA IMPOSED, AND THE RESULTS OF THE TEST CASE THE DECOUPLING METHODOLOGY EMPLOYE FOR V.C. SUMMER 15 BOTH ADEQUATE AND t
PRUDENT.
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SeisMicANCHOR MOVEMENTS i
AND CONTAINMENTMOVEMENT e
STRESSES ARE SECONDARY AND AFFECT T FATIGUE LIFE OF THE PIPING e FATIGUE FAILURE CAN BE PREDICTED BA 3
L MARKL'S WORK AND TEST RESULTS:
o i SNo.2 = C e COMPARISON OF 1/6" VS.1/8" THRESHOLD F SAM ANALYSES USING MARKL'S APPROACH i SNo.2 = C i x 5 = INTENSlFIED STRESS = Se NORMALLY, Se <
SA HOWEVER, TO EVALUATE EFFECTS OF 1/8" SAM &
CONTAINMENT GROWTH, TAKE Se = 2.5 S '
A COMPUTED CYCLES TO FAILURE = S0,000 DESIGN BASIS CYCLES = 400(CONSERVA l
FACTOR OF SAFETY = 125 P
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i SEISMICAt/CHOR MOVEMEllTS AND CONTAINMENTMOVEMENT v
- SAM between buildings combined by
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SRSS i
Similar to method of Re 1.92 for combination of.g. Guide modes NUREG 1061 suggests elimination of the closely spaced mode consideration Building movements would act in a similar fashion - maximum!
displacement in opposite directions is unlikely to occur at the same instant in time l'
SEISMICANCHOR MOVEMENT u
AND CONTAINMENTMOVEMENT r
Containment expansion due to LOCA pressure o
Approximately 1/8" worst case location Single event in life of plant Effect on Pipe j
e Secondary stress on pipe-Not considered in analysis and accepted by audit team Effect on Pipe Supports o
Not considered in support loads (similar to thermal expansion loads)
Thermal considere(similar secondary load) is not d for emergency or faulted cases per FSAR Table 3.9-2
)
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i SEISMICANCHOR MOVEMENTS AND CONTAINMENTMOVEMENT e
INDUSTRY SURVEY SHOWS THAT CONTAINMENT ST CONTAINMENT TYPE IN PIPING ANA VCSNS 7 of 7 PLANTS OF OUR CONTAINMENT DESIGN CONTAINMENT STEADY STATE THERMAL MOVEM PIPING ANALYSIS e
INDUSTRY SURVEY SHOWS THAT CONTAINMENT LOC MOVEMENTS WERE GENERALLY NOT CONSIDERED FOR CONTAINMENT TYPE IN PIPING ANALYSIS PRIOR TO 198 VCSNS 6 OF 7 PLANTS OF OUR CONTAINMENT DESIGN D CONTAINMENT LOCA PRESSURE MOVEMENTS IN T ANALYSIS e
INDUSTRY SURVEY SHOWS THAT SAM THRESHOLDS W CONSISTENTLY CONSIDERED FOR PIPING OTHER PLANTS VCSNS 2 PLANTS ANALYZED SAMs > 1/8"IN THEIR PIPING ANA S PLANTS ANALYZED SAMs > 1/16" IN THElR PIPING A 4 PLANTS IDENTIFIED SAMs AS BEING INSIGNIFICAN CONSIDER THEM IN THEIR PIPING ANALYSIS 1 PLANT ANALYZED ALL SAMs IN THEIR PIPING ANALYSIS 1 PLANT COULD NOTIDENTIFY THElR SAMs CRITERIA ANALYSIS e
INDUSTRY SURVEY SHOWS THAT SAMs BETWEEN BUIL FOR PIPING ANALYSIS PRIOR TO 1982 GEN VCSNS 7 PLANTS USED SRSS BETWEEN BUILDINGS 1 PLANT USED ABSOLUTE SUMMATION (ASUM) BETWEEN
)
BUILDINGS l
3 PLANTS USED A COMBINATION OF SRSS AND ASUM B BUILDINGS 1 PLANTIDENTIFIED SAMs BETWEEN BUILDINGS AS NO 1 PLANT COULD NOTIDENTIFY THEIR SAMs BETWEEN B CRITERIA 1
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INDEPENDENTSEISMICDESIGN\\
VERIFICATION e
ZPA RESPONSE SPECTRA ANALYSIS OVERPREDICTS
)
ACCEPTED TES[TELEDYNE] ANA AGREED THATITS IMPLEMENTATION WOULD LE ACCEPTABLE DESIGN e
SAMs TO RIGID BODY MOTIONSAMs AT SUPPOR BOX TYPE SUPPORTS GENERALLY HAVE A T GAPS i
RIGID BODY MOTION WOULD NOT CAUSE AN PIPING FOR SOME OTHER PLANT ANALYSES OF THIS
[
WERE NOT EVEN CONSIDERED L
DECOUPLING INERTIA RATIO OF 1:10 OR 1:7 INDUSTR L
THE AREA OF COMPUTER TECHNO CONTAINMENT THERMAL GROWTH e
SMALL THERMAL GROWTH AT CONTAINMENT WO CAUSE AN OVERSTRESS CONDITION PIPING STRES$ ANALYSIS WAS NORMALLY GO 111, NC-3600, EQUATION 9, WHICH DOES NOT INCLUDE THE THERMAL EFFECTS L
- THE EFFECTS ON SUPPORT LOADS SHOULD ALS BECAUSE THERMAL AND DESIGN BASIS EARTHQ ARE NOT REQUIRED TO BE COMBINED "SCE&G EXERCISED PRUDENT DESIGN CONTROL TH
. SELECTION OF A COMPETENT DESIGNER, TES WITHOUT BEING PRESCRIPTIVE OF THE ANALYTICAL AND DESIGN TE k
SCE&G PRACTICES ZPA Original Analysis:
ZPA not considered on TES systems, ZPA considered on G/C and Impell Systems.
New and Re-analysis:
ZPA will. be considered on all systems, and handled as a p tant up-grade.
REMAINING ISSUES DECOUPLING Original Analysis:
15% movement ofinertia.
SAM-THRESHOLD Original Analysis:
SAM <1/8 inch can be neglected.
I SAM-BETWEEN BUILDINGS Original Analysis:
SAMs between buildings were combined b the square. y square root sum of CONTAINMENT MOVEMENT p
Original Analysis:
Thermal growth not considered.
New and Re-analysis:
We will evaluate our original practice and the current industry standards to determine criteria.
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SUMMARY
AND CONCLUSION Adequate Technical Management Through gal e
Use of Various Contractors with Different Mode e
Different Modeling Techniques Was And is An Acceptable Practice Design inputs Were Adequately Coordinated By gal e
Rigorous Design Verification Was Performed e
e SWEC Independent Design Verification Supports The Adequacy of Our Program From A Technical and Quality Base u
o e
Used Industry Accepted Practices Engineering Staff Adequate e
Specific Compliance Deficiencies identified Will Be e
corrected and Evaluated as We Feel Appropriate For Genericimpact No significant Safety issues Were identified e
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Retrofiting To Current " State of the Art" Neither Warranted Nor Justified 1
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4 SEISMICANCHOR MOVEMENTS !
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L AND i
AINMENTMOVEMENT i
Conclusion u
L L
The 1/8 inch threshold for SAM ' analysis and-the 1/8 inch,
L containment thermal growth not included in the analysis do ~not reduce safety margins. Adeq margins are demonstrated by the very e L
fatigue eval'uation p' resented; and ther' Wogg, tiv r engineering judgment, that the industry acaetiaej L.
acceptable,is demonstrated valid.
The SRSS method of combining SAM between building is consistent with the industry practice and with the methodology recommended and employed for similar '
loadings. -
1 t movement due to LOCA-pressure. is t
. pipe stress an(support design loads in
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ith the intent,of the commitments in the y
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(
Hr. 0..S..Bradham-
, South Carolina Electric & Gas Company Virgil C. Sumer Nuclear Station cc:-
Hr. R. - V. Tanne r Executive Vice President-S.C. Public Service Authority P.. O. Box 398)
Moncks Corner, South' Carolina 29461-0398 J. 'B. Knotts, Jr., Esq.
Bishop, Cock, Purcell and Reynolds 1400 L Street, N.W.
Washington, D. C.
20005-3502 i
Resident Inspector /Sumer NPS J
c/o U.S. Nuclear Regulatory Comission Route 1,-Box 64
.Jenkinsyille, South Carolina 29065
. Regional Administrator, Region II U.S. Nuclear Regulatory Comission, 101 Marietta Stmet, N.W., Suite 2900 Atlanta, Georgia 30323-Chairman, Fairfield County Council
-P. O. Box 293
. Winnsboro, South Carolina. 29180 i
Mr. Heyward G. Shealy, Chief
. Bureau of Radiological Health South Carolina Departant of Health
'and Environental Control 2600 Bull. Street Culurbia, South Carolina 29201 South Carolina Electric & Gas Company Mr. A. R. Koon, Jr., Manager Nuclear Licensing Virgil-C. Sumer Nuclear Station P.
0.- Box 88 Jenkinsv111e, South Carolina 29065 b
.