ML20207Q458

Summary of 860327 Meeting W/Util in Bethesda,Md to Discuss Discrepancy Between Seismic Response Spectra Peak Broadening Listed in FSAR Vs Broading Actually Used in Design of Unit 2.Attendance List & Related Documentation Encl

ML20207Q458
Person / Time
Site:	Hatch
Issue date:	01/21/1987
From:	Rivenbark G Office of Nuclear Reactor Regulation
To:	Muller D Office of Nuclear Reactor Regulation
References
NUDOCS 8701270360
Download: ML20207Q458 (16)

Text

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January 21, 1987 Docket No. 50-366 DISTRIBUTION

- Decket' File i NRC PDR/L PDR MEMORANDUM FOR: Daniel R. Muller, Director PD#2 Memo BWR Project Directorate #2 DRMuller/GRivenbark Division of BWR Licensing OGC-Bethesda BGrimes/EJordan FROM: George Rivenbark, Project Manager ACRS (10)

BWR Project Directorate #2 NRC Participants Division of BWR Licensing

SUMMARY

OF MEETING WITH GEORGIA POWER COMPANY, (

HATCH UNIT 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 March 27, 1986 to discuss the discrepancy between the seismic response spectra peak broadening listed in the Hatch Ur.it 2 Final Safety Analysis Report versus the broading actually used in the design of Hatch Unit 2. The history of the issue was discussed and GPC presented information (Enclosure 1) that it believes demonstrates that the value of 115% for peak broading listed in the FSAR is an administrative error and that it was never intended to be other than the 10% that was used in the actual design and that was approved at the construction permit stage.

There did not seem to be a reasonable approach to proving that the staff did not base its approval on the 115% broadening listed in the FSAR. The staff stated that it would resolve the issue by determining if the current seismic design is acceptable. This will entail evaluation of the methodology and criteria that GPC has used and proposes to use in its reevaluation of Hatch Unit 1 & 2. A list of meeting attendees is enclosed (Enclosure 2).

Od.pml sQned by

w. . . < o , #, ,

George Rivenbark, Project Manager BWR Project Directorate #2 Division of BWR Licensing

Enclosures:

As stated cc w/ enclosures:

See next page DBlih2 DBL.f GRi bnbark/cd DRMdl er

\/ /87 / /w /87 0FFICIAL RECORD COPY hh P

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v ,

Mr. J. P. O'Reilly Edwin I. Hatch Nuclear Plant, Georg1a.,,

Power Company Units Nos. I and 2 k-cc: l.

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

!!r. Louis B. Long Southern Company Services, Inc.

Post Office Box 2625 -

Birmingham, Alabama 35202 Resident Inspector U.S. Nuclear Regulatory Commission Route 1, Post Office Box 279 Daxley, 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 Chaiman Appling County Commissioners County Courthouse r Baxley, Georgia 31513 f

e ap e ENCLOSURE 1

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ATTACHMENT 1 e

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• 5 HNP-2-FSAR-1 k-TABLE 1.3.2 (SHEET 1 OF 4) 1 -

' SIGNIFICANT CHANGES (HNP-2 ONLY) 9 Change Discussed Item in FSAR Section Reason for Change

1. Core mass flow 4.4.4.5 increase Increased flow due to increase in bypass leakage from 7% to 10%

2. Reactor water cleanup system 5.5.8.2 Changed logic for cleanup system isolation valves isolation topic to close only the outboard isolation valve on high-temperature nonregenerative heat exchanger outlet or i start of standby liquid control system

3. Low-pressure I

coolant 6.3.2.2.4 Provides for closing only injection valve the valve in discharge logic side of unbroken recirculation loop to

• provide for continued depressurization--

permitting core spray

4. Control rod 4.2.3.2.2.3 j drive SCRAM Increased minimum scram discharge volume volume due to BWR operating experience T

increase indicating that leakage

from the seals during scram requires scram discharge volume to be increased

5. Runback of 5.5.1 Changed recirculation

' recirculation logic to runback pump on feedwater pump recirculation pumps when a i trip feedwater pump is tripped '"

6. Onsite 2.3 meteorological Added recorders in main

(( measurements control room to indicate wind direction, wind speed, and vertical

( -

temperature difference in accordance with Reg.

Guide 1.23

, - - - - - , .~,--...-n- . . , - . - - . - -

HNP-2-FSAR-1 TABLE 1.3-2 (SHEET 2 OF 4)

L Change Discussed '

Item in FSAR Section Reason for Chance

7. Rod sequence 7.6.5.2.2 Added RSCS to prevent control system operator from moving

( (RSCS) an out-of-sequence control rod in the 100% to 50% rod density range and from 50% rod density point to 30% of rated power

8. Deleted 1.5.1.2.5

9. Liquid radwaste 11.2, 11.4 Added automatic isolation monitor of radwaste discharge

10. Control room 6.4 Added automatic isolation air intake of control room air intake

(

• from main steam line high flow and radiation signals to prevent detectable airborne radioactivity.

from reaching main room prior to valve closure

11. Chlorine 9.4.1 Added chlorine detectors

detectors in control room intake duct with automatic isolation to ensure that

, an accidental chlorine

' release will not prevent f- continued safe occupation

( of control rocm

12. Seismic 3.7A.4 Compliance with Reg.

instrumentation Guide 1.12

( 13. Jockey pump 2.2.5 Added jockey pump system

( system for residual heat removal (RHR) and core spray to .

prevent water hammer j 14. Plant service 9.2.1 Added standby diesel l

(( water system generator service water pump for supplying cooling i

water to emergency diesel generator 1B

. - ~ _ - _ - - _ - - - . _ . - . . , . _ . . . _ . - - - . - - . _ . . . _ _ . _ _ - , . , , . - , , _ . - - - , - - . - - - - . .

- s HNP-2-FSAR-1 TABLE 1.3-2 (SHEET 3 OF 4) i .

Change Discussed -

Item in FSAR Section Reason for Change

15. Fuel channels 4.2 Changed fuel channel from 80 to 100 mils to provide increased strength

16. Reactor core Chapter 4 Changed core fuel design fuel design from 7 x 7 to 8 x 8 to decrease linear heat generation rate

17. Main steam 6.5 Added MSIV sealing system isolation valve to conform to regulatory (MSIV) sealing requirements to keep doses systam- _ . . _ .

as low as practicable ,

18. Torus drainage 9.3.7 Added torus drainage and

(

and purification purification system to system provide facility for cleanup of suppression pool and to provide installed permanent system for reducing or completely draining the suppression

.. pool -

19. Reactor 7.6.5 Added prompt relief overpressure trip logic to enhance protection reactor control at the end-of-core life (This

, has subsequently been removed.)

C .

5.2.2 Replaced safety valves Supplement SA with safety relief valves (L.

-- . . ~ , -- - , _ . . - _ . - - _ - - - - . . - _ . - . - , , - _ . . , - . . - , - -

HNP-2-?SAR-1 C . . .

TABLE 1.3-2 (SHEET 4 OF 4)

Change Discussed C Item in FSAR Section Reason for Change

20. High-energy Supplement 15A Changes resulting from the breaks outside evaluation of high and

( containment moderate energy line breaks are discussed in Supplement 15A.

21. Primary 6.2.1.2.1.8 Redesigned primary containment containment purge system purge system so that all of gases purged from primary containment will be filtered by standby gas treatment system

22. Low reactor 7.3.1.2.2 Added second reactor

( water level confirmed for automatic depressurization . . -

water level permissive to ADS logic to confirm that water level in reactor vessel -

system (ADS) is low and to provide protection against inadvertent depressurization should an instrument line fail

2.

23. Dryke11 9.3.6 Added drywell pneumatic pneumatic

~

system to supply the system gas-operated valves in the drywell

. 24. Instrument air 9.3.1.2 Divided instrument air system system in reactor building into interruptible and noninterruptible sources to improve reallability and redundancy of the system L

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ATTACHMENT 2 i

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~Bechtel Power Corporation .

Engineers -Constructors 15740 Shady Grove Road Gaithersburg, Maryland 20760

301-948 2700 October 26, 1977 Mr. F. C. Downey General Electric Company (MC-381)'

175 Curtner Avenue San Jose, California 95125

Dear Mr. Downey:

E. I. Hatch Nuclear Plant Unit 2 '

Bechtel Job 6511-020 FSAR Revised Seismic Analysis Reactor Building and Internals File: A-19.1/0150/B-GE-1916 Enclosed please find one advance copy of the revised seismic analysis.

We went to emphasire that the preli=inary seismic analysis was and still is the basic design document. However, if data obtained from the preli=inary analysis causes an overstress problem which can not be relieved by any other means than this revised seimsic analysis may be used after consulting the Hatch Project Civil Group Supervisor.

We waat you to verify the adequacy of all GE's analysis on equipment, components, and systems, that used the final seismic analysis dated April 15, 1975, to sustain the effects of the revised seismic analysis.

Your response is required on or before December 1,1977.

If you have any questions, please advise'. i .

Very truly yours,

W C ~

T. J. Mcdonald Project Engineer CYW:RADW: din Enclosure 1-j ec: V. C. Valekis, w/o enc 1. .

i J. R. Jordan, w/o encl.

K. M. Gillespie, w/o encl. .

p0. Batum, v/o-encl.

J. C. Thornton, w/o enc 1. '

W. A. Widner, w/o encl.

J. A. Arn, w/o enei. '

Bec7tel Power Corporation Engirieers - Constructors 15740 Shady Grove Road

( Gaithersburg, Maryland 20700 301-948 2700 September 26, 1977 Mr. W. A. Widner

• Georgia Power Company I.O. Box 4545 Atlants, Georgia 30302

Dear Mr. Widner:

E. I. Hatch Nuclear Plant Unit'2 '

Bechtel Job 6511-20 ,

seismic Analysis File: A-29.3/A-57.0/0150/B-CP-4796 The Ratch Unit 2 Preliminary seismic Analysis was issued on April 15 & 17, 1972 for the Reactor Building and internals. This docu=ent was to be used as the basis for design. On May 16, 1975 we issued a final seismic analysis for the Reactor Building and internals. The reasons for generating this final analysis are:

1. To upgrade the model to include structural revisions which occurred after the preliminary design stage. -

2. To provide a confirmatory check of the preliminary analysis.

3. To consider composite Modal damping of the soil and struetures (This was an FSAR commitment).

We vant to emphasize that the pre 31minary seismic analysis was, and still is, the basie design document. However, if data obtained from the preli=inary analysis caused en overstress problem which can not be relieved by other practical means then this final seismic analyis may be used af ter consulting the Hatch Project Civil Group Supervisor.

While performing the seismic analysis for the Radvsste Building, a recent

, request of the Client, the final seismic analysis was used as a for=at guide.

l At this time, an error was discovered in the development of the soil spring coefficient.

l This particular calculation had been perfor=ed and checked on project and reviewed by ot:r staff. However, this error was not detected. , .

Since the preliminary seismic analysis was the basic design document none of the structures. are affected by the error. In addition, we have reviewed

Bechtel Power Corporation September 26, 1977 '

Mr. W. A. Widner '

Page Two {,

B-CP-4796 ,

i most of the affected components and conclude that only a minimum number of  !

recently purchased items of equipment were analysed or tested using the i i final seismic analysis data. These are being evaluated based on the corrected '

. final report data. In addition, we have taken the following measures to minimise the probability of a similar re-occurrence:  ;

. 1. A general review of future seismic calculations will be made by the Civil Staff.

2. Any significant deviations from previous calculations will be reviewed in detail by the Civil Staff.

Under the requirements of sub-paragraph (e) (2) to 10CFR50.55, "Conditi as of Construction Permits", a potential reportable deficiency exists concerning the final floor response spectra generated for the seismic analysis of floor supported equipment, components and systems in the reactor building.

The following is a description of the potential reportable deficiency.

l Final seismic analysis of the reactor building used equivalent soil springs and dampers to account for soil-structure interaction. An error was made in the calculation of soil damping coefficients, resulting j in an excessively ds= ped system and consequently, lower structural responses and floor response spectra were generated.

The error in the final seis=ic analysis was corrected and structural responses and floor response spectra were regenerated. The revised final analysis indicates higher responses than the final analysis at some locations, but lower than the preli=inary results which were used in the structural designs and for most of the equipment, component and system designs.

Corrective actions are presently in progress. All equipment, ce=ponents and systems affected by the final anslysis will be identified and their adequacy to sustain the revised final analysis results will be assured.

We expect to submit a final report to your office by December 30, 1977. -

Very truly yours, ,,, A.

g Q -

.l' T. J. Mcdonald , '\

Project Engineer - ..'

CYW: din cc: V. C. Valekis C. H. Burson '

• J. R. Jordan R. W. Staffa F. C. Downey C. R. Miles g W. Majors C. E. Bald J. A. Arn

h e e 0

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ATTACHMENT 3

m - -- w - . _. . . . . . _ . .

Comparison of Response Spectra Peak Broadening Hatch 2 ed for O. L. Received O. L.

App'1g07/28/75 06/13/78

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Plant OL Date FSAR (orig) FSAR (rev)

Ginna Sep 19 69 no mention + 15%

Oconee 1 Feb 6 73 no mention I 10%

Brunswick 2 Dec 27 74 no mention I~ 10%

Millstone 2 Aug 1 75 1 10%** l Trojan Nov 21 75 + 10% l Indian Point 3 Dec 12 75 I 10% + 10%

Beaver Valley 1 Jan 30 76 no mention no mention St. Lucie 1 Mar 1 76 no mention no mention Salem 1 Apr 6 77 no mention + 10%

Davis Besse 1 Apr 22 77 1 10% i10%

Farley 1 Jun 25 77 + 10% + 10%

North Anna 1 Nov 26 77 +

~

15%

Cook 2 Dec 23 77 *

• Three Mile Island 2 Feb 8 78 + 0.02 see no mention Arkansas Nuclear One 2 Jul 18 78 no mention

• Sequoyah 1 Sep 17 80 + 10%

Farley 2 Oct 23 80 + 10% + 10%

McGuire 1 Jun 29 81 7 10% I 10%

Diablo Canyon 1 Sep 22 81 I 10% I 10%

LaSalle County 1 Apr 17 82 I 10% T 10%

~

. Grand Gulf 1 Jul 1 82 I 15%

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Susquehanna 1 Jul 17 82

• Summer Aug 6 82 + 10% + 10%

McGuire 2 Mar 1 83 ni mention i T 10%

~

St. Lucie 2 Apr 6 83 + 20%

i WNP 2 Dec 20 83 + 15%

Palo Verde 1 Dec 31 84 I 15%

Enrico Fermi 2 Jul 15 85 {20% no update

• Peak broadening was used but no value was given in the ?SAR.

• • Currently do not have a copy of VSAR page (obtained verbally) e

. .s.. .-

Response Spectra Peak Broadening History for E. I. Hatch-Unit 2 i TABLE I E. I. HATCH UNIT 2 FRS

SUMMARY

r COMPUTER PROGRAM DATE OF USED TO BROADENING ITEMS EVALUATED OR CATEGORY I FRS BROADEN VALUE QUALIFIED USING BUILDING GENERATION FRS ACHIEVED CORRESPONDING FRS Reactor 1972 Note 1 Note 1 Original design of most equipment, piping and subsystems 1975 CE921R 15%(2) None 1977 CE921 10% Equipment, piping, and subsystems evaluated after 1975.

Control 1971 No. 8046 10%C4) All equipment, piping and subsystems within.

Main Stack 1972 Note 1 Note 1 All equipment and subsystems attached.

Intake Structure 1971 No. 8046 10%C4) All equipment, piping and subsystems within.

Diesel 1971 Note 3 Note 3 All equipment, piping and subsystems within.

NOTES:

(1) Hand enveleping of FRS data performed. Intended broadening value was

+ 10%.

l (2) This value applied to major FRS peaks only. Other FRS points enveloped by hand.

l i

(3) Hand envelope of data from 3 analyses performed. Effective broadening exceeds + 10%.

(4) Program raises all spectra points within the specified range to the peak value, however, new frequency pairs are not generated.

0644A

Q. O ENCLOSURE 2 2

lf. ATTENDANCE LIST NRC/GPC MEETING TO DISCUSS HATCH UNIT 2 SEISMIC RESPONSE SPECTRA PEAK BROADENING

~

MARCH 27, 1986 NAME ORGANIZATION i

G. Rivenbark NRC/NRR R. Herman NRC/NRR R. Goodard NRC/0 ELD E. Reiss NRC/0 ELD L. Gucwa GPC A. Domby GPC-Consultant t

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