ML20210A061
| ML20210A061 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/03/1987 |
| From: | Rivenbark G Office of Nuclear Reactor Regulation |
| To: | Muller D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8702060397 | |
| Download: ML20210A061 (40) | |
Text
{{#Wiki_filter:- _ A 3 FEB 1987 a Dockets Nos. 50-321/366 MEMORANDUM FOR: Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing FROM: George Rivenbark, Project Manager BWR Project Directorate #2 Division of BWR Licensing
SUBJECT:
SUMMARY
OF MEETING WITH GEORGIA POWER COMPANY, HATCH, UNITS 1 and 2 A meeting with representatives of Georgia Power Company (GPC) and personnel from the NRC was held at the NRC's offices in Bethesda, Maryland, on September 11, 1986, to discuss details of GPC's seismic reevaluation of Hatch, Units 1 and 2. A copy of the meeting agenda, copies of charts used in discussing agenda items and a list of meeting attendees are enclosed. The charts summarize the reevaluation discussion in considerable detail. The staff requested that additional detailed information be submitted by the licensee for use by the NRC's consultants in reviewing the methodology used by GPC. This request was subsequently augmented and formalized by NRC letter to GPC dated October 1, 1986. GPC's response was submitted by letter dated November 14, 1986. Ofiginal signed by George Rivenbark, Project Manager BWR Project Directorate #2 Division of BWR Licensing
Enclosures:
As stated cc w/ enclosures: See next page DISTRIBUTION Docket File ~ NRC PDR Local PDR PD#2 Reading File DMuller GRivenbark OGC-Bethesda EJordan BGrimes ACRS (10) NRC?yticipants D hDH2 DB TMf2 G iv nbark:ct' d fer 2/S/87 2/7/87 8702060397 870203 PDR ADOCK 05000321 P PDR
= _
Mr. J. P. O'Reilly Edwin I. Hatch Nuclear Plant, Georgia Power Company Units Nos. I end 2 cc: Bruce W. Chruchill, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Washington, D.C. 20037 Mr. L. T. Gucwa Engineering Department Georgia Power Company Post Office Box 4545 Atlanta, Georgia 30302 Mr. H. C. Nix, Jr. , General Manager Edwin I. Hatch Nuclear Plant Georgia Power Company - Post Office Box 442 ' Baxley, Georgia 31513 Mr. Louis B. Long Southern Company Services, Inc. . Post Office Box 2625 Birmingham, Alabama 35202 Pesident Inspector U.S. Nuclear Regulatory Commission Route 1, Post Office Box 279 ' Baxley, Georgia 31513 Regional Administrator, Region II U.S. Nuclear Regulatory Commission, 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303 Mr. Charles H. Badger Office of Planning and Budget Room 610 270 Washington Street, S.W. Atlanta, Georgia 30334 Mr. J. Leonard Ledbetter, Commissioner Department of Natural Resources 270 Washington Street, N.W. Atlanta, Georgia 30334 Chairman Appling County Commissioners County Courthouse Baxley, Georgia 3151:
.9 q ATTENDANCE LIST SEPTEMBER 11, 1986 MEETING WITH GPC ON HATCH SEISMIC DESIGN Name Affiliation GRivenbark NRC DBL:PD#2 CPTan NRC DBL:EB RAHermann NRC DBL:EB JJJhonson EQE PSobel NRC DBL:EB KLWhitmore SCS Structural CLWeaver Bechtel Civil - EGS JDHeidt GPC Licensing DPMoore SCS Civil LBlong SCS Nuclear Safety Robert P. Kennedy Structural Mechanies Consulting .
DBLiaw NRC DBL:EB l l
NRC MEETING j AGENDA Concerning the Evaluation of Floor - Response Spectra Peak Broadening 1 E.I. Hatch Nuclear Plant Units 1 & 2 l September 11,1986, 9:00 A.M.
- l. Opening Remarks - L.B. Long (SCS)
II. Enhancements to Seismic Analysis as part of the 1984 Peak Broadening Evaluation - D.P. Moore (SCS) A. Introduction & Overview B. Update of Seismic Models 3 C. Soil Damping 1 l D. Development of New Time Histories ' ! E. Revised Damping for Cable Tray Supports l and Piping Systems l F. Summary & Conclusions l Ill. General Discussion j IV. Closing Remarks - L.T. Gucwa (GPC) ' l 6 A l
/ '
PART 21 EVALUATION . FLOOR RESPONSE SPECTRA ~ PEAK. BROADENING ,l. HATCH NUCLEAR PLANT UNITS'1 & 2 . NOVEMBER,1984
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STATEMENT OF PROBLEM: i I FLOOR RESPONSE SPECTRA (FRS) DID
~
NOT IN ALL CASES MEET SAR' - f COMMITMENTS FOR PEAK SPECTRA BROADENING. - J
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SUMMARY
IINUINEERING ] 'Il METHODOLOGY FOR SAFETY EVALUATION i 9 1, j 1. REVIEWED ORIGINAL CALCULATIONS 1 f
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- 2. MADE IMPROVEMENTS (ENHANCEMENTS TO MO98L-S .
AND INPUT)
- 3. DEVELOPED CRITERIA THAT WAS FOLLOWED BY.'BOTH 4 ORGANIZATIGdS l
- 4. GENERATED NEW PRS MEETING LICENSING .
COMMITMENT (i.E. 10%) j -
- 5. COMPARED NEW FRS TO ORIGINAL FRS '
- 6. EVALUATED THE EFFECT OF NEW FRS O'N-SUBSYSTEMS ]
- 7. TECHNICAL INPUT WAS PROVIDED BY CONSULTANTS '
'i REANALYSIS RIGOROUSLY PERFORMED WITH i TECHNICAL ADVICE FROM CONSULTANTS TO ~~
J l-EVALUATE PLANT SAFETY IN REGARDS'TO. i SPECTRA PEAK BROADENING. k .......c. . .
. , . , ~
1 j
SUMMARY
- 1. NEW FRS WERE GENERATED USING IMPROVED MODELS, SOIL DAMPING, ETC. WHICH MEET OR IS MORE CONSERVATIVE THAN FSAR REQUIREMENTS.
- 2. NEW FRS BROADENED TO 10 %
- 3. DBE WAS USED TO EVALUATE SAFETY.
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- 4. DETAll CRITERIA DEVELOPED FOR FRS GENERATION, FRS i l COMPARISON, AND SUBSYSTEM EVALUATION. THIS PROVIDED li CONSISTENCY BETWEEN ORGANIZATIONS.
l' l '
- 5. HIGHLY QUALIFIED CONSULTANTS PROVIDED INPUT AND REVIEW FOR THIS SAFETY EVALUATION.
rosimca-si
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SUMMARY
- ccONruNuEDy .
- 6. ONLY A SMALL PERCENT OF THn NEW PRS WERETF00RD. l SIGNIFICANT ENO. UGH TO WARf% ANT AN. EVAL.#AT$OF , i EQUIPMENT AND S.TROCTURAL SUBSYSTEMS; TME-WhMN -
REASON FOR MOST OF TN'ESE NEW PRS EXCE20tMG:TliE i - ORIGINAL PMS WAS DUE TO MODeLING IMPROYEMEfftS.
- 7. EQUIPMENT AND STRUCTURAL SUBSYSTEMS.WEfttfi%fMD 1 ~
NOT TO BE AFFECTED BY THE C'HANGES lN fMt PR$. - 1
- 8. COMPARISON OF FRS POR PIPING SYS' TEM 9.'SMOWED NO j :
SIGNIFICANT INCREASE IN SPECTRA ACDE6 ERAT 10NS, ; THEREFORE, NO IMPACT ON PIPING SYSTEMS. ,
- 9. COMPARISON OF FRS FOR CABLE TRAY StWORTS SMDWED ,
l NO SIGNIFICANT INCREASE IN SPECTRA AGOELERATIONS; l THEREFORE, NO INPACT ON CABLE TRAY SUPPORTS.. ' 70 3144ACM12 ,
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i CONCLUSION: - Y . NO REPORTABLa: CONDITION PER~.TME'ORifEMIA ;')
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OF PART 10 CPR 21 EXIST FO.R TPtE Fle6D54 .- -
~l RESPONSE SPECTRA FOR EDWiN~Is ItATC@ .
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NUCLEAR PLANT-UNIT 1 AND 2.- r TE3144AC006-13 l I l l l
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J t Enhancements To Seismic. Analysis
~
As Part Of The 1984 Part 21 Evaluation i 6
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i 1 . Changes Made To Original Seismic Models i During 1984 Part 21 Evaluation , i f B 1 i l
1 i i - UNIT 1 REACTOR BUILDING:
- Upgraded Properties To Account For Interaction Between Reactor Vessel & Reactor Building Model UNIT 2 REACTOR BUILDING: '
- Upgraded Properties To Account For interaction Between Reactor Vessel & Reactor Building Model i .
l .
s CONTROL BUILDING: UNIT 1 ANALYSIS
- Replace Original Seismic Model With The Unit 2 Seismic Model
- Included Mass Moment Of inertia W/ Rocking D.0F UNIT 2 ANALYSIS I
- Included Mass Moment Of inertia W/ Rocking DOF ,
4 4
B I DIESEL GENERATOR BUILDING: i UNIT 1 ANALYSIS ] e Revised Mean. Gs To Agree With Unit 2 Value 1 e included Mass Moment Of inertia W/ Rocking DOF UNIT 2 ANALYSIS .. l l e included Mass Moment Of inertia W/ Rocking DOF f Y
. 4 1
i i i . i ! INTAKE STRUCTURE: UNIT 1 ANALYSIS AND UNIT 2 ANALYSIS i i e Revised Soil Springs i e Revised Mass & Stiffness Properties t i I l
- Included Mass Moment Of inertia W/ Rocking DOF '
i i ' I........ ; 1 . l l
r l i ! MAIN STACK: ' l l 1 e Model Reformated To Run On Standard l QA Computer Program j ! I 4 l i I i l I
1
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1 1 i
; i I ! SOIL DAMPING - UNIT 1 FSAR COMMITMENT ,
l TABLE 12.3-2 , i i i, i h r i i ! 4 i > l l \' -
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HNP l-FSAR-12 . TABLE 12.3-2 DAMPING FACTORS FOR SEISMIC ANALYSIS IN PERCENT OF CRITICAL DAMPING <a g Operating Basis Design Basis Earthquake Earthquake Reinforced concreti structures 3.0 5.0 Steel frame structures 3.0 5.0 Bolted and riveted assemblies 3.0 5.0 Welded assemblies 2.0 3.0 Vital piping 0.5 1.0 Translation and rotation of foundation soil 4.5 5.5
- a. As of April 4, 1985, damping per figure 12.3-2 for piping systems and figure 12.3-3 for cable tray supports is used for all new and replacement systems and load reconciliation work.
REV 2 7/84 REV 3 7/85 o
i 1 iSOIL DAMPING - UNIT 2 i l IFSAR COMMITMENT ' i t ! TABLE 3.7A-2 ! ! SAFETY EVALUATION: ' l i o ! THE SMALLER OF FOR TRANSLATIONAL SOfL.DAMPlNG j 100% OF TABLE 3.7A-2 i or l 75% OF TABLE 3.7A-2 + 6% SOIL MATERIAL D' AMP 1NG '! ' o FOR DIESEL. GENERATOR BLDG USED THE FOLLOWING - TO ACCOUNT FOR SOIL VARIATIONS ! 30% OF TABLE 3.7A-2 + 6% SOIL MATERIAL DAMPING !, or 20% DAMPING , WHICHEVER WAS SMALLER '
HNP-2-FSAR-3 e TABLE 3.7A-1 DAMPING FACTORS FOR SEISMIC ANALYSIS IN PERCENT OF CRITICAL DAMPING <a [ Operating Basis Design Basis Earthquake _ Earthquake Reinforced concrete structurec 3.0 5.0
-Steel frame structures 3.0 5.0 Bolted and riveted assemblies 3.0 5.0 Welded assemblies 2.0 3.0 Vital piping 0.5 1.0 Translation and rotation of foundation soil' D' 4.0 5.0 .
l I
- a. As of April 4, 1985, damping per figures 3.7A-22 and 3.7A-23 for piping systems and cable tray supports, used for all new and replacement systems and loadrespectively, is reconciliation work.
b. In lieu of using the soil damping values specified in this table, the equations soil damping in table 3.7A-2 may be used to calculate coefficients. REV 3 7/85 e 4
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Q HNP-2-FSAR-3
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TABLE 3.7A-2 (SHEUT1OF2) FORMULAS FOR EQU1 VALENT FOUNDATION SPRING CONSTANTS AND DAMPING COEFFICIENTS l (RECTANGULAR 3ASE) Equivalent Equivalent Motion Spring Constant Damping Coefficient Hori7ontal K = 2(1 + v) C6 g )BL c = 0.576 k Rx 1Qp/G Rocking K = 3 fy 6,B L 8 c =3; k,RYp/G Vertical K = -' l
"' O z BL c = 0.85 k g R Yp/G where:
B = width of the base mat in the plane of horizontal excitation L = length of the base mat perpendicular to the plane of horizontal excitation v = Poisson's ratio of foundation medium G = shear modulus of foundation medium p = density of foundation medium R = equivalent radius of the base mat as defined below i I 6, 6,, S = constants that are functions of the dimensional x g ratio, B/L (from figure 10-16 in reference 3) 3 (1-v) I, B, = 8pR' where: I = total mass moment of inertia of structure and base mat about the rocking axis at the base REV 2 7/84 REV 3 7/85
1 i TIME HISTORY APPROACH TO OBTAIN FRS UNIT 1 l . ORIGINAL 1984 ANALYSIS
- 1. N-S Component Of The 1940 El Centro Earth-quake Scaled To The OBE Level Of 0.08 g
- 2. Used Norminization Ratio OBE Raw Floor Spectra X (Sf)
Max. OBE Floor Accel. From Response Spectrum Analy. g Max. OBE Floor Accel. From Time History Analy. ! 3. DBE FRS = OBE FRS X (A) DBE Synthetic Time j History That Envelops The DBE DRS @ 5% Damping I DBE Max. R. Accel. From Reeponse Spectrum Anoly.
! Where A = Avg Of O.E %. n. w. Fm R spectrum Anai.
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1 TIME HISTORY APPROACH TO OBTAIN FRS UNIT 2 f ORIGINAL 1984 ANALYSIS Modifed Taft 1952 Earthquake Horizontal Component Scale To The OBE Level of 0.08g i DBE FRS = OBE FRS X (A) DBE Synthetic Time . History That Envelops The DBE Max M. Acc.l. From Meeponse Spectrum Analy. DBE DRS @ 5% Damping Where A = Avg Of one m.. n.A i.rrom M ,on s,.ctrum An.i. l l 1 f
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t i REVISED DAMPING FOR CABLE TRAY SUPPORTS: l l
- Higher Damping Than Originally Used Justified By Tests i :
e Used Higher Damping in Part 21 Evaluation i i e A 50.59 Evaluation Was Performed - ) e FSAR Revised To Allow Higher Damping For New Or ' Replacement Systems And Load Reconciliation Work b
*1 24 W TO FULLY LOADED TRAY 20 l
\ 6 l, l c 16 E 15% 12 0 !
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4 5% 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 ZPA OF FLOOR RESPONSE SPECTRA, ACCELERATION ( g ) NOTES: 1. For unloaded tray, use damping values specified in Table 12.3 2 for steel structures. For tray loaded less than 50%, linear interpolation is used to determine l the applicable design damping value. l 2. As of April 4,1985, damping per this figure is used for all new and replacement systems and load reconciliation work. REV 3 7/95 DAMPING CRITERIA FOR SEISMIC ' EDWIN I. HATCH ANALYSIS OF CABLE TRAY SUPPORTS GeorgiaPower nuci.can n.aur - unir i FIGURE 12.3-3 l a.w
i . . ^l 1 l REVISED DAMPING FOR PIPING SYSTEMS: i i i 1 l e Higher Damping Than Originally Used Justified i By PVRC Study l e Used Higher Damping in Part 21 Evaluation ,
- A 50.59 Evaluation Was Performed l
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- NRC Accep'ts GPC Use Of Code Case N-411 W/ Specific Requirements 1 e FSAR Revised Accordingly ,
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5 I I 4 - I I T I 9 i D 3 - l 6 I e i o I E 2 - 1 I I I l 1 l l i l I i l I i 1 0 I l 0 10 20 30 FR EQUENCY (Hz) NOTES: 1. Applicable to both OBE and DBE, independent of pipe diarneter.
- 2. As of April 4,1985, damping per this figure is used for all new and replacement systems and load reconciliation work.
l l l 1 s. REV 3 7/96 E MIM8 HATCM DAMPING CRITERIA FOR SEISMIC GeorgiaPower nuci.ran ei.4=r uuir i ANAI,YSIS OF PIPING SYSTEMS - seso FIGURE 12.3-2 e ao -
SUMMARY
i j e All Modeling Changes Reflect A More Accurate Representation Of l The Structures And Conform To The FSAR Commitments + l e For Unit 2, The Soit Damping Used Was Mora Conservative Than That Permitted in The FSAR But Was Considered A More-Realistic Estimation !
- The New Synthetic Time Histories Meet The FSAR Commitments And More Accurately Represent The Seismic Design Bases Of The Plant Thus Producing More Realistic FRS e The Use Of More Realistic Damping For Cable Tray Supports And Piping System Has Been Shown Not To Reduce The Intended Seismic Safety Margins At Plant Hatch And Therefore Does Not -
Constitute An Unreviewed Safety issue O t,
j . . _ ) II ( -
,h CONCLUSION ii l i
- 1. FRS produced using the criteria developed in
- 1984 are a more realistic representation of the l FSAR commitments.
! i i 2. Use of these FRS coupled with the higher damping
'} for cable tray supports and piping does not reduce the intended seismic safety margins at Edwin I. Hatch .l Nuclear Plant.
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