ML13333B867

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Summary of 821217 Meeting W/Util,Bechtel & Westinghouse Re Analysis Criteria Concerning Seismic re-evaluation of Plant Concerning SEP Topic III-6.Viewgraphs Encl
ML13333B867
Person / Time
Site: San Onofre 
Issue date: 03/15/1983
From: Paulson W
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8303220248
Download: ML13333B867 (28)
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Text

March 15, 1983 DISTRIBUTION Docket ELlordan NRC PDR JMTaylor Docket No. 50-206 Local PDR ACRS (10)

ORB Reading SEPB NSIC SVarga DCrutchfield DVassallo HSmith RClark WPaulson JStolz LICENSEE: SOUTHERN CALIFORNIA EDISON COMPANY OELP FACILTTY: San Onofre Nuclear Generating Station, Unit No. 1

SUBJECT:

SUMMARY

OF DECEMBER 17, 1982 MEETING (SEP TOPIC 111-6, SEISMIC DESIGN CONSIDERATIONS)

On December 17, 1982, members of the NRC staff met with representatives of Southern California Edison Company (SCE). The purpose of the meeting was to discuss analysis criteria regarding the seismic reevaluation of San Onofre Unit No. 1. Enclosure I is a list of attendees.

The NRC staff summarized the results of the December 16, 1982 meeting.

The NRC staff stated that the use of 100 maximum stress cycles in the augmented evaluation is clearly conservative. If the licensee wants to use less than 100 cycles, the licensee will have to justify the

,lower value. The licensee suggested that 10 cycles was adequate.

Further, SCE indicated that they would be willing to shut down if there is an OBE. This issue was not resolved; the licensee will consider the staff's recommendation and will discuss thi$s further at a later date.

The licensee then discussed the basis for the assertion that one set of time histories is sufficient to evaluate the reactor coolant loop.

The viewgraphs used are shown in Enclosure 2. Conservatisms cited in the reactor coolant loop analysis are:

1. Input conservatism because the sphere enclosure was not included in the model.

SO

2. Used minimum values for material strengths
3. Conservative damping factors were used
4. Inelastic behavior was not quantified Bechtel represenlatives estimated that there would be about a 40%

reduction in acceleration if the enclosure structure is included in the analysis.

OFFICE SURNAME&8303220248 830315 DAT pD ADOCK 05000206 NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

2 -

March 15, 1983 The method of combining loads resulting from building motion and inertial loads was discussed. SCE used the absolute sum method of combining displacement of adjacent independent structures used for developing seismic anchor movement loads and stressess The NRC staff stated that this is a conservatism and that it would be ac ceptable to combine such displacements using the SRSS method.

Similarly, the staff indicated that it would be acceptable to com bine inertial loads due to independent seismic responses of adja cent pipes or common supports. However, seismic anchor motion and inertial loads must be combined using the absolute sum method.

Original signed by/

Walter A. Paulson, Project Manager Operating Reactors Branch #5 Division of Licensing

Enclosures:

As stated OFFICE 03

-i.i

83.

DATE4.......

.....3.

/S

/83 0...

NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-335-960

3 -

March 15, 1983 cc Mr. R. Dietch Charles R. Kocher, Assistant Vice President General Counsel -

Nuclear Engineering and Operations James Beoletto, Esquire Sbuthern California Edison Company Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Post Office Box 800 Rosemead, California 91770 Rosemead, California 91770 David R. Pigott Orrick, Herrington & Sutcliffe 600.Montgomery Street San Francisco, California 94111 Harry B. Stoehr San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Resident Inspector/San Onofre NPS c/o U. S. NRC P. 0. Box 4329 San Clemente, California 92672

  • Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego San Diego, California 92101 California Department of Health ATTN:

Chief, Environmental Radiation Control Unit

Regional Radiation Representative 215 Freemont Street San Francisco, California 94111 Robert H. Engelken, Regional Administrator Ruclear Regulatory Commission, Region V 1450 Maria Lane Walnut Creek, California 94596

CLOSURE DECEMBER 17, 1982 MEETING WITH SOUTHERN CALIFORNIA EDISON COrPANY ATTENDANCE LIST Name Affiliation Walter Paulson NRC Pei-Ying Chen NRC Ken-Herring NRC Mark Hartzman NRC Paul Koss Bechtel William S. LaPay Westinghouse Jack Rainsberry SCE Russ Krieger SCE Rick Ellis Bechtel Jerry Gartland SCE Hank Peters SDG&E Eileen McKenna NRC W. T. Russell NRC F. J. Miraglia NRC' Ken Baskin SCE

41.CLOSURE 2 ISSUE SUFFICIENCY OF ONE SET OF TIME HISTORIES TO EVALUATE THE REACTOR COOLANT LOOP

APPLICABILITY OF ONE SET OF SEISMIC TIME HISTORIES ONE SET OF SEISMIC TIME HISTORIES FOR "ELASTIC NONLINEAR SYSTEMS" IS SUFFICIENT IF THE TIME HISTORIES MEET THE FOLLOWING CONDITIONS:

1. PRODUCE SEISMIC RESPONSE LEVELS CONSISTENT WITH THE SEISMIC EVENT DEFINED FOR THE SITE.
2. PRODUCE NONLINEAR IMPACT LOADS CONSISTENT WITH THE DEFINED SEISMIC EVENT.
3.

ARE OF SUFFICIENT DURATiON AND FREQUENCY CONTENT TO-REFLECT THE POSSIBLE DIFFERENT PHASE RESPONSE IN THE NONLINEAR SUPPORTS.

ELASTIC NONLINEAR SYSTEM A SYSTEM WHEN SUBJECTED TO DYNAMIC EXCITATION UNDERGOES STRUCTURAL SUPPORT IMPACTS AND LIFT-OFFS.

THE STRUCTURAL SUPPORT ELEMENTS REMAIN ELASTIC DURING ANALYSIS,

SAN ONOFRE UNIT I STUDIES LINEAR AND NONLINEAR SEISIIC RESPONSE STUDIES WERE PERFORMED TO SHOW THAT THE THREE CONDITIONS NECESSARY TO SHOW APPLICABILITY OF ONE SET OF TIME HISTORIES ARE MET,

SAN ONOFRE UNIT I REACTCR. COOLANT LOOP SEISMIC ANALYSIS.

-ELASTIC NONLINEARITIES CONSIDERED 0

STEAM GENERATOR SEISMIC STOPS O FUEL ASSEMBLY IMPACTS 0

NOpN-TLINEAR STIFFNESS IN PUMP HANGERS 0

REACTOR LIFT-OFF 0

PUMP LIFT-OF-F 0

STEAM GENERATOR LIFT-OFF

CONCLUSION 1 CONDITION 1 IS SATISFIED SLNCE THE TIME HISTORIES USED TO EXCITE THE REACTOR EUILDINHG PRODUCED SEISMIC RESPONSE LEVELS CONSISTENT WITH THE RESPONSE LEVELS S

COCIATED WITH THE SITE DEFINE DBE HOUSNER GROUND SPECTRA.

COMPARISON OF BUILDING SEISMIC DISPLACEMENTS Displacements Inches Tin -listory T

SRS CSM Location Direction Displacement Rage Range Operating Deck X

10).36 0.2 to 0.3 0.4 to 0.5 Northeast Building Y

0.30 0.3 to 0.4 0.5 to 0.6 Corner Z

0.55 0.4 to 0.6 0.7 to 0.8 Foundation Node 1 X

y Z

0.4 0.3

==0.4

CONCLUSION==

2 CONDITION 2 IS SATISFIED SINCE THE TIME HISTORIES PRODUCE NON-LINEAR IMPACT LOADS OF EXPECTED MAGNITUDE ASSOCIATED WITH THE DEFINED SEISMIC

EVENT,

SINGLE MASS MODEL ANALYTICAL MODEL 4o K

ASSUMPTIONS ASSUMPTION 1:

THE SUPPORTS DO POT MOVE AND THE MASS RESPONDS IN TUNE WITH THE HORIZONTAL BUILDING MOTION, ASSUKPTiON 2:

THE SUPPORTS RESPOND IN A SINUSOIDAL MANNER AND THE MASS IS IN RESONANCE WITH THE SUPPORT MOTION,

STEAM GENERATOR UPPER SUPPORT IMPACT LOADS SINUSOIDAL MOTION A____

SG B3 SG C NW SE W

r SW NE No.

Description*

Support Support Support Support Support Support

1.

Impact Forces -

101 112 120 145 129 123 Assumption 1

2.

Impact Forces -

678 74 1536 655 773 738 Assumption 2

3.

Average Impact -Forces 390 429 328 400 451 430 From I and 2

4.

Impact Forces from 410 435 438 400 460 550 Time-History Analysis Assumption 1 Supports do not move and mass responds in tune with the horizontal building motion.

Assumption 2 Supports respond in a sinusoidal manner and the mass is in resonance with the support motion.

Units:

kips

CONPARISON OF ANALYSIS ASSUMPTIONS AND ACTUAL RESPONSE CHARACTERISTICS THE REACTOR BUILDING DOES NOT RESPOND AS A PERFECT SINUSOID DURING THE SEISMIC EVENT, ITS MOTION CAN BE SINUSOIDAL FOR SHORT-TIME SEGMENTS OR SINE-BEAT IN CHARACTER OR RANDOM, MOTIONS OTHER THAN SINUSOIDAL CAN SIGNIFICANTLY REDUCE THE IMPACT LOADS.

2, THE STEAM GENERATORS DO NOT RESPOND IN PERFECT RESONANCE WITH THE REACTOR BUILDING,

3. THE REACTOR BUILDING MOTION DOES NOT HAVE A CONSTANT ACCELERATION THROUGHOUT THE SEISMIC EVENT.

FOR THE IMPACT LOADS DEVELOPED

AEOVE, A MAXIMUM ACCELERATION LEVEL WAS USED,

STEAM GENERATOR UPPER SUPPORT IMPACT LOADS SINE BEAT MOTION SCxA SGB SGC NO.

DESCRIPTION NW SE W

E SW NE SUPPORT SUPPORT SUPPORT, SUPPORT SUPPORT SUPPORT 1

Impact forces from time 410 435 438 400 460 550 history analysis 2

Range of impact force 340 375 270 330 385 370 assuming Sine beat to to to to to to motion 540 595 430 525 620 590 Units:

kips

CONCLUSION 3 CONDITION 3 IS SATISFIED SINCE THE TIME HISTORY SET USED IS OF SUFFICIENT DURATION AND FREQUENCY CONTENT TO REFLECT DIFFERENT PHASE RESPONSE IN THE NONLINEAR SUPPORTS,

CONCLUSION 3 CONDITICN 3 IS SATISFIED SINCE THE TIME HISTORY SET USED IS OF SUFFICIENT DURATION AND FREQUENCY CONTENT TO REFLECT DIFFERENT PHASE RESPONSE IN THE NONLINEAR SUPPORTS.

BASIS

1.

IMPACTS ARE "SIMPLE" IN THE REACTOR COOLANT LOOP SYSTEM MODEL, NOT "COMPLEX" LIKE A "BOX OF MARBLES",

2.

THE T1E HISTORY COMPONENTS ARE SATISTICALLY INDEPENDENT,

3.

THE DURATION OF THE EARTHQUAKE IS 20 SECONDS,

4.

THE TIME HISTORY INPUT TO THE REACTOR COOLANT LOOP IS FILTERED PASSING THROUGH A COUPLED SOIL-BUILDING-LOOP MODEL.

QFT OF SOIL ST1 FFESS TION CONSIDER A30 VARIATION IN SOIL STIFFNESS COMM E NTS

1.

DOMINANT BUILDING MODE PERIODS WILL CHANGE BY +/- 15%.

2.

SEISMIC TIME HISTORY INPUT TO THE REACTOR CODLANT LOOP WILL NOT INCREASE SIGNIFICANTLY SINCE THE BUILDING MODES USED IN THE ANALYSIS ARE NEAR THE PEAK REGION OF THE HOUSNER SPECTRA; 3,

BASED ON SINGLE MASS MODEL FORMULATIONS THE MAXIMUM INCREASE IN IMPACT LOADS IS APPROXIMATELY 5%. THIS IS NOT SIGNIFICANT.

CONCLUSION A SOIL VARIATION OF +30% WILL NOT BE SIGNIFICANT TO THE REACTOR COOLANT LOOP SEISMIC RESPONSE (STRESS/FORCES).

SUPPLEMENTARY CONVENTS

1. A SOIL VARIATION OF +/- 30% WILL NOT BE SIGNIFICANT TO THE REACTOR COOLANT LOOP SEISMIC RESPONSE.
2. THERE IS A SIGNIFICANT RESERVE STRENGTH IN THE REACTOR COOLANT LOOP COMPONENTS AND SUPPORTS.
3.

THE SURPLUS MARGINS REPORTED FOR SAN ONOFRE UNIT I DO NOT REFLECT THE RESERVE STRENGTH CAPACITY,

4. THE REACTOR COOLANT SYSTEM CAN BE SUBJECTED TO HIGHER SEISMIC LOADS WIThOUT LOSS OF SAFE SHUTDOWN CAPABILITY BECAUSE OF THE RESERVE STRENGTH, Frm

SEISMIC DESIGN MARGIN C RESERVE STRENGTH 0 CONSERVATISM "BUILT" INTO A GIVEN STRUCTURAL. SYSTEM IS A FUNCTION OF THE CONSERVATISM IN THE DESIGN OF THE INDIVIDUAL PARTS.

  • RESERVE STRENGTH IS A FUNCTION OF THE TOTAL SYSTEM AND NOT JUST THE INDIVIDUAL PARTS,
  • ACTUAL DESIGN MARGI LWILL BE LARGER THAN THE SUM OF THE "INDIVIDUAL PARTS" OF A TOTAL SYSTEM, C THE MARGIN IN SEVERAL "INDIVIDUAL PARTS" WILL BE TABULATED AND COMBINED TO GIVE A LOWER BOUND ESTIMATE OF THE MARGIN AVAILABLE.

7FICAL FACTURS AFFECTING SEISMIC DESIGN MARGIN 1,

STRESS CRITERIA

2.

MATERIAL STRENGTH 3, DAMPING FACTORS 4,

INELASTIC BEHAVIOR

SU ARY OF MINIMUM SEISMIC DESIGN MARGIN FACTORS

1.

PIPING..

2.0

2.

SUPPORT CRITERIA 2

1,43 3

ATERI AL 1.20 4. D AMI NG 1.t\\Ca.......

.l.........

.... a PIPING SYSTEM..,.

............ 1.25 BUILDING R-ESPONSE a

1.27 OTHER AREAS OF SIGNIFICANT CONSERVATISM NOT QUANTIFIED

1.

SEISMIC INPUT DEFINITION

2.

SEISMIC COMPONENT RESPONSE COMBINATION

3.

DUCTILITY AND INELASTIC RESPONSE

STRESS CRITERIA AND MATERIAL SEISMIC DESIGN MARGIN

REFERENCES:

1, NUREG/CR-2137, RODEAUGH, E.C.: K. D. DESAI, "REALISTIC SEISMIC DESIGN MARGINS OF PUMPS, VALVES, AND PIPING, ORNL/

SUB-2913/11, JUNE 1981.

2. NUREG/CR-1161i *"APPENDIX D" NUCLEAR PLANT SEISMIC MARGINA" RECOMMENDED REVISIONS TO NUCLEAR REGULATORY COMMISSION SEISMIC DFSIGN CRITERIA, JUNE 1979,
3.

BUMPUS, J,,'STRUCTURAL AND MECHANICAL RESISTANCE, UCID 17965, LAWRENCE LIVERMORE LABORATORIES, LIVERMORE, CALIFORNIA, 1979,

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EXAMPLES OF HIGHER DAMPING 1,

NRC LETTER 12/21/77. DP. ALLISON. DOCKET ND50-27S AND SO-323 PROVIDES DAMPING VALUES OBTAINED BY J.P, KNIGHT FROM JAPANESE SEISMIC SHAKING TESTS OT TOKAI - 2 REACTOR BUILDING, BUILD.ING EXCITATION - 10 GAL, DAMPING COEFFICIENT OF FUNDAMENTAL MADE IS ESTIMATED OVER 20%

2.

DAMPING VALUES IN EXCESS OF 10 PERCENT ARE REPORTED FOR-PIPING AND ARE REPORTED IN THE TWO REFERENCES BELOW.

A. MORRONE, A.,

"DAMPING VALUES OF NUCLEAR POWER PLANT COMPONENTS" WCAP-7921, WESTINGHOUSE NUCLEAR ENERGY SYSTEMS,

NOVEMBER, 1971, NES CLASS 3, B. STEVENSON, J.D., "STRUCTURAL DAMPING VALUES AS A FUNCTION OF DYNAMIC RESPONSE STRESS AND DEFORMATION LEVELS", PAPER Kll/1, PRESENTED AT THE 5TH INTERNATIONAL CONFERENCE ON STRUCTURAL MECHANICS IN REACTOR TECHNOLOGY, AUGUST, 1979,

C SEVERAL PLANTS STUDIED IN DETAIL

  • EL CENTRO STEAMI PLANT (PRESENTED AT ASLB, AUGUST, 1980),
  • HUACHIPATO STEEL PLANT, CHILE (1960) (DIABLO CANYON FSAR),
  • ESSO REFINERY COMPLEX, MANAGUA, NICARAGUA (1972)

(DIABLO CANYON FSAR),

C NO SIGNIFICANT DAMAGE IN THESE STRUCTURES AND SYSTEMS, THESE PLANTS DID NOT HAVE A CURRENT DAY SEISMIC QUALIFICATION PROCESS PERFORMED.

ONLY HE TIKE HISTORY ANALYSIS IS NECESSARY IN THEREANALYSIS OF THE SAN ONOFRE UNIT I REACTOR COOLANT LOOP.

BASIVS

1. THE TIME HISTORIES USED EXCITE THE LINEAR PORTIONS OF THE MODEL TO THE PROPER SEISMIC RESPONSE LEVELS AS DEFINED USING THE HOUSNER GROUND SPECTRA.
2. THE TIME HISTORIES USED PRODUCE IMPACT LOADS IN AGREEMENT WITH EARTHQUAKE BUILDING MOTION CORRESPONDING TO THE DEFINED SEISMIC
EVENT,
3. THE TIME HISTORIES USED ARE STATISTICALLY INDEPENDENT, 4,

THE TIME HISTORIES ARE PASSED THROUGH A "COUPLED SOIL/BUILDING MODEL" RESULTING IN A SUFFICIENT NUMBER OF DIFFERENT PHASE CONDITIONS THROUGHOUT THE 20 SECOND EARTHQUAKE DURATION.

5, THE NONLINEARITIES IN THE REACTOR COOLANT LOOP SYSTEM ARE "SIMPL" WHERE PHASING BETWEEN EARTHQUAKE COMPONENTS IS NOT AS SIGNIFICANT, 6,

SOIL VARIATION IS NOT SIGNIFICANT, 7,

THERE IS SUFFICIENT RESERVE STRENGTH IN THE REACTOR COOLANT SYSTEM.

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