ML13311A219

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Summary of 791114 Meeting W/Util in Rosemead,Ca Re SEP Seismic Review
ML13311A219
Person / Time
Site: San Onofre Southern California Edison icon.png
Issue date: 01/03/1980
From: Levin H
Office of Nuclear Reactor Regulation
To: Crutchfield D
Office of Nuclear Reactor Regulation
References
TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8001160653
Download: ML13311A219 (98)
v • d • e


Text

JAN 0 3 1980 MEMORANDUM FOR:

D. M1. Crutchfield, Chief Systematic Evaluation Program Branch, 00R FROMI:

H. A. Levin Systematic Evaluation Program Branch, DOR

SUBJECT:

MEETINb SUHMARY - SOUTtiERN CALIFORNIA EDISON C0lHPAY On November 14, 1979, representatives of the Southern California Edison Company (SCE) met with the NRC Staff and NRC staff consultants -in Rosemead, CA to discuss details of the Systematic Evaluation Program Seismic Review of tne San Onofre Nuclear uenerating Station Unit 1 (SOiGS 1).

Participants are listed in Enclosure I and the meeting agenca is provided in Enclosure 2.

The purpose of the meeting was to discuss the following specific items:

1.

The NRC Review of a report entitled; "San Onofre Nuclear Generating Station Unit 1, NRC Docket 50-206, Seismic Reevaluation and Modification," April'29, 1977, Southern Cali fornia Edison Co.

and San Diego Gas & Electric Company.

2.

Details pertaining to the scope, criteria, and schedule for SCE's seismic reevaluation program (Phase 2) for the balance-of-plant and items inside containment riot evaluated in Phase 1.

SCE provided an overview of Phase 1 of their SONGS 1 seisnic reevaluation program which was started in Mlay, 1973.

Phase 1 included an evaluation of the containment sphere, reactor building andithe reactor coolant system.

Certain pieces of attachea piping which are part of the reactor coolant pressure boundary were riot inclucied. In April, 1975, SCE submitted a seismic reevaluation methodology report to the NRC which was later endorsed by the NRC staff in July, 1975. Hodifications to the SONGS 1 facility were made during an extended Cycle VI refueling and ma-intenance outage in 1976.

In April 1977, SCE submitted a final report documenting Phase 1 of their seismic reevaluation program.

A detailed summary of the overall technical effort was made Dy SCE and their consultants.

This presentation was followed by an open discussion of all parties including a aetailed question ana answer period.

The Tol owing open is ues were icencified Dy the ARC staff for resolution SUNME i...,,

1 1hat is the i hluence onn inear behavilor as a funct on of OAT5jt ty~ChSO t ' --

1 l

l j

ltII u I-Eill c

iti LI dy I= IORM 318 (9.76) NRCM 0240 1Wu.o..oosn.ft~tJ eITING OPC c

-8

- as

-7e

-2 envelope the design response spectra, increased response may be possible as a function of the frequency, amplitude and energy content of alternate time histories.

2. Quantify the influence of the Enclosure Building on the dynamic response of the containment/reactor building considering approp riate soil-structure interaction effects.

-3. Provide an example design calculation for both a drilled and grouted anchor bolt installation. Clarify which anchoring methods were utilized in the modifications.

4. Provide an example stress calculation including the method of deriving the rebar stress from the finite element analysis results showing how a representative concrete section was either checked or designed.
5. Clarify how the stress quantities Pm, PI and Pb were determined for nozzles and various plates and shells in the NSSS.

SCE committed to provide answers to as many of these issues as possible within a couple weeks. An action plan will be developed for any issue which requires more detailed evaluation.

SCE indicated that the nonlinear dynamic system analysis of the reactor coolant loop was performed with Westinghouse Computer Code WECAN using the method of director integration and that the verification of the WECAN code was provided in the Westinghouse topical report WCAP 8929.

However, the HRC staaff stated that the review of report WCAP 8929 has not been completed by the NRC and a confirmatory analysis for the reactor coolant loop to be performed by the NRC might be necessary to confirm the structural integrity of SONGS 1 reactor coolant loop system, components, and their supports under seismic excitation.

The staff agreed to inform the SCE within a couple of weeks whether or not a confirmatory analysis would be carrid out by the staff.

SCE stated that an evaluation of the remainder of the reactor coolant pres sure boundary and systems necessary for warm shutdown would take approxi mately two years.

SCE estiamtes that an evaluation of the additional systems needed to attain cold shutdown and accident mitigation would take an additional 1-1.5 years.

The design of any modifications has not been included in these estimates.

SCE stressed that these estimates are preliminary and that they hope to provide more detailed information by the end of 1979.

SEE PREVIOUS CONCURRENCE 6 FC

)O.

URNAME............ I

.Systematic Evaluation P ogram Branch DATE~

Division CfOperating R actors/

NRC FORM 318 (9-76)

NRCM 0240 U. S. GOVERNMENT PRINTING OFFICE: 1976-626-624

DISTRIBUTION:

Central Files SEPB RDG HLevin PKuo

-2 envelope the design response spectra, increased response may be possible as a function of the frequency, amplitude and energy content of alternate time histories.

2.

Quantify the influence of the Enclosure Building on the dynamic response of the containinent/reactor building considering approp riate soil-structure interaction effects.

3. Provide an example design calculation for both a drilled and grouted anchor bolt installation.

Clarify which anchoring methods were utilized in the ioifications.

4.

Provide an example stress calculation including tihe 3etnod of deriving the rebar stress fromi the finite element analysis results showing-how a representative-concrete section was either checked or designed.

5.

Clarify how the stress quantities Pm, P1 and Pb were determined for nozzles and various plates and shells in the WSSS.

SCE committed to provide answers to as many.of these issues as possible within a couple weeks.

An action plan will be developed for any issue which requires more.detailed evaluation.

Discussions later focused on Phase 2 of SCE's seismic reevaluation program.

SCE stated that an evaluation of the remainder of the reactor coolant pres sure boundary and'systems necessary for warm shutdown would take approxi mately two years.

SCE estiamtes that an evaluation-of the additional systems needed to attain cold shutdown and accident mitigation would take an additional 1-1.5 years.

The design of any modifications has not been included in these estimates.

SCE stressed that these estimates are preliminary and that they hope to provide more detailed information by the end of 1979.

Howard A. Levin Systematic Evaluation Program Branch Division of Operating Reactors

Enclosures:

As stated

.S.E.PB.:

D~

$EPB.;DQR.A

..Q R.

SUNM_.I..e jn.Thij......

Tirq.........

lpfmnyer...... nr f ~ e 12 -79 2479 12

-79 12 -79 t--=__

NKC FORM 318 (9-76) NROM 0240 uM*

U

  • ovenuouse Parnn1es a CUe tees -

7ee 69

ENCLOSURE 2 MEETING AGENDA SCE/NRC MEETING November 14, 1979 I.

Introduction W. C. Moody II.

Bechtel Work P. Koss III.

Weastinghouse Work W. LaPay IV.

Additional Discussion of NRC P. Koss/W. LaPay Questions V.

Conclusion W., C. Moody

SAN ONOFRE NUCLEAR GENERATION STATION L

UNIT 1 SDIES.L GENERATOR BUILDING A DMINISTRATION-CONTROL BUILDING ENCLOSURE BUILDING F.W. HEATER PLATFORM CONTAINMENT SPHERE rn ri rJ TURBINE NORTHS (jTURBINE PEDESTAL EXTENSION r, Li Li

-i0 L.)

(A c

STACK FUEL HAE t

TU RBIN STORAGE PLATFORM SOUTH VENTILATION BUILDING EXTENSION BUILDING-I PIPE TRENCH CONDENSATE STORAGE REACTOR-AUXILIARY INTAKE STRUCTURE REFUELING WATER STORAGE TANK STOP GATE SALT WATER

& MOV-9 SEAWALL COOLANT PIPES a

Ium I

Rraomtc P-rhdu jc-We rY6lrmOri mm CA T N-O. 25 3-201 e

~~~

I~..

VI i-LlJ "I~.EidTTN i 0 *N C O-N 11.

SAI i VA I llo IVI

1k~~~~~~.

I.

I ANPIS FIGURE 1.4-4 AOdaE

N

4.

ii

  • COMPLETE h

SUPPORT STRUCTURE AND PRWf~ARY PIPI~JG, LOOP A 1

'K I, J I I

.1 I

4 4

I

COMPLETE SUPPORT STRUCTURE,

LOOP B 7-

.COMPLETE SUPPORT STRUCTURE AND PRIMARY PIPING, LOOP C

.1

SPECTRA HORIZONTAL TRACE A AT=

0.01 SEC 2000 PTS MAX ACC.67 1.82 1----

DAMPING =.0200 1.56 1.30 --

1.04

.7 8

.67 S.52

.26 0.01

.10 1.00 10.00 PERIOD (SEC)

/SPECT3A HORIZONTAL TRACE B AT 0.01 SEC 2000. PTS MAX ACC.67

______DAMPING

=

0200 1.56 1.30-

.0 4

~.78-LU

-1.67 0.01

.10 1.00 10.00 PERIOD (SEC)

SPECTRA VERTICAL DT=

0.01 SEC 2000 PTS MAX ACC.33G 1.0 DAMPING =.0200

.g

.8

.6

.4 cm:

3

.2

.1 0._

.01

.10 1.00

0. O PERIOD (SEC)

GENERAL STRUCTURAL ACCEPTANCE CRITERIA A.

TO MAINTAIN STRUCTURAL INTEGRITY AS REQUIRED FOR SAFETY FUNCTION WHEN SUBJECTED TO DBE LOAD COMBINATIONS.

B.

TO PREVENT STRUCTURAL DEFORMA TIONS FROM IMPAIRING THE SAFETY FUNCTION OF CATEGORY A SYSTEMS AND EQUIPMENT.

SEISMIC REEVALUATION STRUCTURAL ANALYSIS COMPUTER PROGRAMS A.

BECHTEL STRUCTURAL ANALYSIS PROGRAM (BSAP)

B.

SPECTRA COMPUTER PROGRAM (SPECTRA)

C.

SYMBOLIC MATRIX INTERPRETIVE SYSTEM PROGRAM (SUPER SMIS)

D.

OPTIMUM CONCRETE DESIGN PROGRAM (OPTCON)

E.

REINFORCED CONCRETE DESIGN FOR AXIAL FORCE AND BIAXIAL BENDING (BIAX) o+ ~ tweden ru -jII RLIIII Qv~I 1 1-7Eflr-OVJ..

METHODS EMPLOYED TO COMBINE STRESS RESULTANTS FROri ORTHOGONAL EARTHQUAKE COMPONENTS METHOD FOR COMBINING STRESS RESULTANTS SYSTEM BEING ANALYZED TIME HISTORY METHOD SRSS METHOD*

CONTAINMENT SPHERE NO YES CONCRETE REACTOR BUILDING FOR CHECKING ONLY YES PRIMARY COOLANT LOOP YES NOT APPLICABLE

  • SRSS METHOD:

oxt =

Xy +z oxt= TOTAL X STRESS o x=

X STRESS DUE TO X DIRECTION EARTHQUAKE ox= X STRESS DUE TO Y DIRECTION EARTHQUAKE a

= X STRESS DUE TO Z DIRECTION EARTHQUAKE

4c,, 500 60,

< 4uu G (100 Km um ) 2/3 om=2/3 GO 30 Um/

3 U

300 0= OVERBURI)EN PRESSURE 20_-_UNITS LB/FT 2 m -

200

_jj 100 10- 5 10-4 10-3 10-2 10-1 MAJOR PRINCIPAL STRAIN, ePERCENT 15 13 LA.

CD 7

EXTRAPOLATED 5

MEASURED uj 3

= 1 104 10-3 10-2 10-1 MAJOR PRINCIPAL STRAIN, e, PERCENT MODULUS AND DAMPING VS STRAIN SAN MATEO FORMATION SAND 12186-002 T 1317 F31

PLOT DEAD LOAD "Z" STRESS (LBS/FT 2) 120.0 50.0 20.0

-20.0 5 0

-20.0

3000, 100 1500

--100.0 1000 200 300 500 400

-200.0 EV. (FT)

EFFECTIVE U Mil IVIU U LU S KSF FOR THE SAN ONOFRE SITE*

X103 60 6.0 KSF

~CONFINING PRESSURE 50 5.0 KSF

.40 4.0 KSF 3.0 KSF 30 2.0 KSF 20 10 10

-mmasum 10-4 10-3 10-2 10-1 MAJOR PRINCIPAL STRAIN (%

  • BASED UPON WOODWARD McNEILL SOILS REPORT

0"r FDYNA~MiC MOnemDULUS ON ROCK(ING MODE FREQUENCY 8.0 7.0 6.000

~5.0 0

2.0 5,000 10,000 15,000 20,000 25,000j K...

SOIL DYNAMIC MODULUS (KSF)

CONTAINMENT (EQUIVALENT BEAM)

E-1C COMPARTMENT E-1B3 COMPARTMENT OPERATING DECK REFUELING CANAL E-1A COMPARTMENT PRIMARY SHIELD FOUNDATION NODE

MSSS MACRO ELEM ENT 460 404 579 390 293 295 387 294 524 521 458 384 459 381 391 289 389 308 28 493 292 287 392 388 385 454 199 300 383 455 301 386 382 198 200 201 305 202 309 317

NODE 6

NODE 2 yx NODE 41 NODE 1 NODE 5 NODE3 N

FREQUENCY COMPOSITE MODE (CPS)

MODAL DAMPING DESCRIPTION AND PRINCIPAL MOTION 1

1.81 2.88 PRESSURIZER NORTH-SOUTH 2

1.85 3.0o PRESSURIZER EAST-WEST 3

2.96 2.70 G-2A NORTH SOUTH, E-1 BEAST WEST 4

3.06 2.36 G-2A NORTH SOUTH, G-2B EAST WEST 5

3.36 4.68 G-2A NORTH SOUTH, G-2B EAST WEST STRUCTURE 6

3.41 15.97*

VERT., SOIL/STRUCT/INTERACTION G2A, G2C 7

3.42 4.52 G-2C (NS/EW) E1C (VERT) SOME 8

3.57 8.71*

STRUCTURAL ROCKING (NS) G2A (NS) G2C (EW) 9 3.70 6.61*

STRUCTURAL ROCKING (EW) G-2C(EW) REACTOR INTERNAL (NS) 10 3.71 3.25 REACTOR INTERNALS (EW, NS) 11 3.72 3.53 REACTOR INTERNALS (EW, NS) 12 4.95 2.16 G-2A (EW) TOP REACTOR (NS) 13 5.32 2.15 G-2C (NS, EW, VERT) G-2A, G-2B 14 5.42 2.05 TOP REACTOR (NS, EW) G-2A, G-28 15 5.45 2.10 TOP REACTOR (NS, EW) G-2B 16 5.55 2.12 TOP REACTOR (NS) G-28 (EW) 17 5.76 2.18 G-2B (EW) TOP REACTOR (NS) E-1A (VERT)

MODE(CS 19 18

.962.20 TOP REACTOR (NS) G-2A (EW), E-B (VERT) 19 6.20 2.26 G-2 20 65

.0 20 653

.30G-213 (NS) E-113 (EW) 217.022

  • 21 7.0 2.20G-2A (NS), G2C (EW), E-2CEW) 227.921
  • 227.892.15G-28 (NS), EB (EW) REACTOR AND OTHER LOOPS 23 83 23 8.362.30 E-IC (NS) G-2C (EW) 2489725
  • 24 8.7 2.56G-2A (IS,EW,VERT), E-JA (N-S) 25 1.182.90 E-1A (N-S,EW, VERT), G2C (VERT) E-IC (E-W) 26103
  • 6 0.46.47*

STRUCTURAL TORSIONAL MODE' 27 10.44 2.31 G-2A (N-S; VERT), G2C (VERT), E1IC (NS) 28 1-224 28 10.2 2.43E-IA (N-S,E--W) G-2A (N-S) 29111 2911196.17*

G-213 (VERT) G-2A (EW,Ns, VERT) 30113 3011314.48 E-JC (E-W) E-113 (N-S) 31119 3111993.02 ElB (N-S), G2A (N-S,E-W), G-2B (VERT) 32 1.421 32 1.34

.10G-2A (E-W, VERT), E-IC (E-W), REACTOR (E-W) 33137 3313792.18 G-2C (N-S), E-1C (N-S,E-W) REACTOR (N-S) 34 1.827 34 1.982.7 E-B (N-S), G-2C (N-S,E-W, VERT), E-11C (N-S) 35 14.25 2.74 G-2A (N-S'E-w) E-IA (E-W) PRESSURIZER (VERT) 36 1.628 F

UG-2E (E-W), PRESSURIZER (VERT C) 37 14.55 2.25 P REACTOIZR-(NS) G-2A3 (EW), E-8 V)

FREQUENCY COMPOSITE MODE (CPS)

MODAL DAMPING DESCRIPTION AND PRINCIPAL MOTION 38 15.39 4.09 ROCKING 39 15.55 2.05 TOP OF REACTOR (1 VERT), G-2C (E-W) 40 16.49 4.31 TORSION 41 17.06 4.20 E-1C (N-S) 42 18.69 4.46 E-1C (N-S) 43 19.16 4.44 E-1C (N-S) 44 20.28 3.11 G-2A (E-W) G-28 (N-S) 45 20.65 3.01 G-2B (N-S,E,W), G-2A (E-W) E-1C 46 21.24 2.97 E-1C (N-S) G-2C (N-S, E-W) E-1A (N-S) 47 22.25 2.56 E-1A (N-S), G-2A (E-W) E-1B (N-S), G-2B (N-S) 48 22.38 3.23 E-1A (N-S, E-W) G-2A (N-S, E-W) E-18 (N-S) 49 22.94 4.00 G-2A (E-W), E-1B (N-S) 50 23.69 2.76 G-2B (N-S), G-2A (E-W) REACTOR, N-S) 51 24.31 4.28 G-2A (E-W), G-2B (N-S) 52 24.52 2.49 G-2C (E-W,N-S) G-2B (N-S) 53 25.34 2.81 G-28 (N-S-E-W) E-1A (E-W) E-1C (E-W) 54 25.88 3.53 E-1C (E-W) 55 26.20 3.62 E-1C (E-W) 56 28.03 4.93 G-2B (N-S) 57 28.59 4.78 ROCKING

FREQUENCY COMPOSITE MODE (CPS)

MODAL DAMPING DESCRIPTION AND PRINCIPAL MOTION 58 30.12 5.04 TORSION

- 59 32.38 5.32*

TORSION 60 33.93 4.85 ROCKING 61 34.45 5.58*

STRUCTURAL TORSION 62 35.47 5.81*

ROCKING TORSION 63 36.76 5.04*

ROCKINGj SHEARING 64 37.60 5.03*

ROCKING, TORSION

AM

-I

.71 AVINUfi.

--RE LAT Pf

_~

-7

.EH T E L LIHiEAR I..7

.81'

.-3.45 8

x

.6 5

.J4 15

_5_

.62"2 fA:~L.1.33

.A3 G-2 ~2.20-,

1.70 2 235

.0

  • -~Z 7-K:.53

-. 20 OPERATING x

.9.4 D MKy.38 3.

NE Z

.71

.5 FOUNDATION, Y

.00.0

.401

I 5.13X10 0

04 2

1.02XI0 4

2.05X10

'I 04 5

2.59X 10 04 6 3.OOX 0'

04 04 4

'04 3

6 5.3I

%A L4 3

WEST WALL 01+11-IEQI M(YY) g~t 3*

0p4 2

.5'1 04 3 2.78XI 0I 4

3.71X/

(NORTH WALL DL+LL+EQ. M(YY) 16.94xIo 0 0'4 3

1.-75X 10 0 4

2.27X 10 04 6

3.33X10 04 6

3.33X 10 04 87 3.86X10

04.

9 14.9X10 04' 9O 4.91 x10 04 30514 33 44 5

7S 5/

n

OPERATING DECK DL+LL+EQI M(XX),

I 6.62XI102 03 2

3.41X10 6

03 3

6.16X10 03 64 4

8.91X10 04 7

04 6

1.44X10 04 7,

t.71X10 673 04 a

1.99X 10 9

2. 26X 10 04 10 2.54lsX1 5

~6 55 4

LONGITUDINAL DIRECTION 177 --- 226 1

-1.5 2

-6 1624 46 t

225 108 EARTHQUAKE EARTHQUAKE DEAD LOAD PRESSURE LOAD HORIZ N-S VERTICAL FORCES (KIPS/FT)

.2

.1

-. 12

.48.27 1.0 1.3

.48.1.

.51 EARTHQUAKE EARTHQUAKE DEAD LOAD PRESSURE LOAD HORIZ N-S VERTICAL MOMENTS (KIPS-FT/FT)

L-164 aw a4Q CO

t.

t,*

HOOP DIRECTION 17 2

.8

-.1 8

-.3 25 6

-307 46~

F o263 50 240 EARTHQUAKE EARTHQUAKE DEAD LOAD PRESSURE LOAD HORIZ N -- S VERTICAL FORCES (KIPS/FT)

.16 13

-.1.13 3

.15

-15

-38.3

.4 EARTHQUAKE EARTHQUAKE DEAD LOAD PRESSURE LOAD HORIZ N-S VERTICAL MOMENTS (KIPS-FT/FT)

SUMMARY

TOTAL(a)

STRIESS AT MOST CRITICAL POINJT NODE POINT NODAL POINT 19 DESCRITION LONG'L STRESS PSI

.HOOP STRESS PSI MEMBRANE BENDING MEMBRANE BENDING STEEL SPHERE PRESSURE 18909 2763 19366 820 DEAL LOAD

-260 62 372 18 SEISMIC 1154 216 1462 63 PIPE DL + THERMAL AXIAL

-483 495

-474 830 OVERT MOMENT, 322 1183 168 1839 SEISMIC AXIAL 498 504 482 845 OVERT MOMENT 264 963 136 1498 TOTAL 20404 6186 21512 5913 TOTAL MEMBRANE

+ BENDING 26590 27425

a. TOTAL STRESS MEMBRANE:

as

+ as + as

+ op

+

DL P

EQ DL+T EQ BENDING:*

P

+ o

+

+ o STRES WAD P

S E, UP PE

+T E

  • THE ABSOLUTE VALUE OF THE SUM OF THE BENDING STRESS WAS CONSIDERED AS AN UPPER BOUND.

-~~~~~~.

.~4

PLOT DEAD LOAD "Z" STRESS (LBS/FT 2

- 120.0 50.0 20.0 00

-50

-20.0 3000 2500.

2000, 100 1500

.-100.0 1000 200 300 500 400

-200.0 LEV. (FT)

REACTOR COOLANT SYSTEM TIME HISTORY SEISMIC ANALYSIS 0 Time History Using 3 Orthogonal Loading Functions-Statistically independent

  • Elastic Analysis with Geometric Nonlinearities included in the Model
  • Component Internals are Included in the Analysis 8153-65
  • I

ELEVATION OF SEISMIC MODEL forth Bldg. Corners South Bldg. Corners 40'6" Elev.

Building Pt. Reactor Foundation O'4" Elevation Support 10'10" Elevatio 8153-1 Soil Rocking Stiffness 1

I, e FSoil Vertical Stiffness Free Field 4

iL

REACTOR COOLANT SYSTEM DYNAMIC MODEL X

Elbow Loop C Loop B Elbow Reactor Pressurizer Loop A Elbow Coolant Pump Steam Generator 8153-3

LOOP B SYSTEM MODEL RCP Stiffnesss of Stiffness ofl II Surge Line Ig Added/l Press.

8163-4p

II SYSTEM STEAM GENERATOR MODEL Stiffness of Steam Line Added El. 49'-6 3/4" Stiff rass of Feedwater Line Added El. 31'.7" El. 17'-5" 8153-29 LO

NONLINEARITIES INCLUDED IN SYSTEM MODEL

  • Non-Linear Stiffness in Pump Hangers
  • Reactor Lift-Off o Pumps Lift-Off

TABLE C.2-1

SUMMARY

OF GAP CONDITIONS Units - Inches Loop A Steam Generator Hot Standb*

Upper seismic support North-West 0.20 South-East 0.00 Gap opposite Snubber North-East 5-1/4 Lower seismic support North-West 0.03 South-East 0.03 Loop B Steam Generator Upper seismic support West 0.05 East 0.22 Gap opposite snubber South 6-1/8 Lower seismic support east 0.15 East 0.2 0.28 Loop C Steam Generator Upper seismic support North-East 0.04 South-West 0.20 Gap opposite snubber North-West 5-7/8 Lower seismic support North-East.

0.03 South-West 0.24 ot Standty Condition Date January 13, 1974 Temperature of coolant loop 535*F Pressure 9 pil Fluid level of steam generator 501 Fluid levP1 nf nes

.Il

STEAM GENERATOR SYSTEM MODEL Stiffness of Steam Line Added El 49'-6/4" Gap Gap Hanger Stiffness of Feedwater Line Added Ga p, Loop B El 17'-5" Plan at El 49'-61" Gap.

Cap El 17'-5" 8153-5

I.0v~

PLATES tADOED 1976, REFPLACE EXISTING N MEM9ERS TYPICAL IADO 1971 LOOP ASHOWN LOOPS 24 C MMILMI OOYERl PLATES (AOOEO 1976)

IREACT09VEIE L i.

UM 1ATEALLSUPPORTS

~ ~~Uid MOEL LOOP B Plan at El. 22'-4"j East Wall Ncii

-doll El. 22'-4" Col.

Hanger Dn.,

Dn..

CaCol Dn Plan at El. IV'.090 Saulwfng Canal Wall

Spr I

K!

D ard c* Upward W.2 for cotof Negative CAP.2'O of CostantForc.

0.173 11-489 "Cn~rsanor Mvc Donar anger Pf.od P~u W

SM' 0J N 0 RE U

IT t1 E gACTO To CRDM Seismic Support To Head Lift Mechanism &

CRDM Support CRDM (45)

Head Litt Lug (3)

CL Top Supt. Plat Hold Down Spring RCC Cde(45~)

Control Rod Drive Shaft (45)*

"Upr-.r. Support Columns (29)

[

COre Barrel Thre a ShPlat Upper Core Plate.-,

FuLl AeCembe P Lower Core Plate Lower Core Support Cotumns Lower Core

-S i

Secondary Core Support

~Core Support Castittt

REACTOR SYSTEM MODEL Up-per Hca-a p7 Mating Line Supt Cola and

~ 1Guie-Tubes Upvo.. Core Plei41.

suport coumnF Nwlzontal St. ort VI

-(3 Typ)

'Fue-Carel LAv Lower Cr lt Cof a Svoport C=Nt1 Lowm hf a

PRES SUP'-71=

SYSTEMA MODEL El. 44'-O I

Plan at El. 448-0" Prowurizur Co.

n Col. Dn.

Col. Dn.

Col. Dn.

El. 14'-0" Pan at El. 14'-O" 4-44 Pressurizer and Framing

11 p~&RZEA u-.PE LAT: AL IlL.

UPPER IATZAL SUPP'ORT 3.83-i2

DAMPING VALUES USED IN SSE DYNAMIC ANALYSIS Percent Critical Damping Main Coolant Loop - Structural 4

Main Coolant Loop - Impact 8

Reactor Fuel Assembly - Structural 10 Reactor Fuel Assembly - Impact 25 Reactor Internals - Structural 2

  • ,S

FLOW CHART NONLINEAR SEISMIC TIME HISTORY ANAL YSIS OF NSSS 9

Detail RCL

+

Simplified

"~d-soil IW atornsal 8153-113 mmaturam-am

SIMULTANEOUS THREE-DIMENSIONAL 20 SECOND EARTHQUAKE Son spring 8153 -23

.Vert ca.45G

UPPER NORTHWEST SEISMIC STOP GAP 0.2 INCH STEAM GENERATOR "A"9 Force(KIPS) 100.

-100

-200

  • 300.

-400 10.0 10.5" 11.0 11.5 12.0 12.5 Time (Seconds) 8153 -19

SEISMIC.STOP FO Cso.01IC

  • STEAM GENERATOR- "A"g Force (Kps) 10

-400 l0ob 10.5 11.0 11.5 12.0 12.3 TO=_____-.--

-C 915 M

SEISMIC RESPONSE Impact Force (KIPS) 300 200 Force In East Seismic 100 Stop 49'6" Elev.

0 I I I

l i

f f

l Displacement (Inches) 0.4 II 0.4 East-West Displacement 0.2 0.22 East Steam Generator "B" GAP Relative to NE Bldg. Corner 0.0 0.05 West GAP

-0.2 II I1 1I l

Impact Force (KIPS) 0 7

100 Force in West Seismic Stop 49'6" Elev.

200 300 Curves Show Correlation Between Li LStearn Generator Motion and Force 12.5 13.5 1 14.5 In Seismic Stops 13.0 14.0 15.0 8153-13 Time (Seconds)

SEISMIC RESPONSE Displacement (Inches) 4.0 North-South Displacement 2.0 Top of Pump "A" 0.0 V

2.0 Torque (106 In.- Lb) 0.Torque in Cold Leq Pipe at Elbow 10 0

-10 Moment (106) In.-1b)

20.

Moment In Cold Leg, 0

-10

-20

  • Curves Showing Correlation

.0 -Between Pump Motion 10.0 11.0 12.0 and Pump Loads Time (Seconds) 8153*.14 1I

TABLE D.5 DISPLACEMENTS AND ACCELER ALUES FROM THE REACTOR COOLANT LO ALYSIS Displacements* -inches Acceleration - g COMPONENT LOOP NODE PT.

X Y

Z X

Y Z

Pump A

128

.40

.82

.66 127

.39

.60

.66 1.5 3.0 2.0 Pump B

265

.95

.70

.77 264

.70

.50

.77 2.0 2.0 2.1 Pump C

405

.74

.60

.76 404

.57

.45

.76 2.0 2.5 2.0 Steam Generator A

96

.56

.46

.58 2.2 1.7 0.9 Steam Generator B

235,

.45

.71

.62 1.2 2.5 1.0 Steam Generator C

372

.59

.48

.61 2.8' 1.7 0.9 North-East 450

.36

.30

.55 Building Corner Foundation 48

.00

.00

.40 RC STEAM NORThEAST CORNER

. NS PUMP GENERATOR OPERATING DECK1 Y *EW 128,265,405 96,235, Upper 3

etclJ72

-Ltea NE 2

  • ertcal127,264,404 L

72ateral N

Support 1G1 soo Deflections relative to free-field ground motion.

SA

-4.

-STEAM11GEM~

D'!SPLACEMENTS 0

.2.*

TIM~ SEGM[NT 13.5 TO 15.5 SECONDS-I 7

0-.1 2

13.5

.14.5 15.5 TIM9,SECONDS 0

k I

1~~~

~.i i

i

.2.

1.14515.5 13.5 14.5

.15.5 TIK,SECOH~DS TIME,SECONDS

tSTEAM GENERATOR B 0

U PPER SEISMIC STOP LOADS If.__ jTI r-i SEGUENT 13.5 TO 15.5 SECONDS I

'~

-100~

-200..

w-02 7

Cl

-~~

-300.

  • 13.5 14.5 s

~STEAM GENERATOR TIME, SECONDS.

100 4

It-,.-

'GAP-005

-200-11I 13514..5-15.5

.FOR.I.

PM;2 A LATERAL SUPPORTS TORKE SEMET

.5 TO 11.5 SECON4DS ARZEb AL GAP AR ZER 200 10i TIME SE ID 9.5 105

11.

I-4

-120 9.5 11.5 T11E. SECONMs

-300 (Gap donotclse 200.

-J 200 100 l

11.5 12.5 1.5 01

.5 1 T.D Z51.

TIME$ SECONDS TIME, SECONDS TIME, SECONDS NO RTHEAST

  • PULW HANIGER LOADS 10TOTAL TINESEGRENiT 11.5 TO 13.5 SIC.

PUMP HANGER LOADS 10 so-so:

-200 TlSCtD F.

-12AM6 TIME,~~TME SECONDSIM ECND

STEAM GEN WR B TM HN)GER LOADS 100 TIME SEGMENT 11.5 TO 13.5 SECONDS

-60 0

-100

-100

-140

-200 I

I v-180 11.5 12.513.5 NSE1.5

12.

13.5 TIMESECONDS

TIME, SECONDS NANGER CONNECTION HANGER CONNECT ION 20 0 C

w 200 100 F

STEAM GENERATOR B

100 0 10 a10 A

-20 O A.p00

-200 115 TIME, SECc13.5 11.5 12.5 13.5

TIME, SECONDS

NE LIFT-OFF OF NEc VESSEL COLUMN RPV COLUMN? LOADS 0

TIME SEGMENT 11.5. SECONDS TO 13.5 I

SECONDS

1.

-NE 9

-)SW 12' NW 15 E.

-400ON

~ 600~

-J 2800--

! iT11-S 11.5ECND 1251.

LIFT-OFF OF SW VESSEL COLUMN FT

~~600_-----

0~~~~~~~

4001 II 2001

-200

~-20

~

{

~

?A-II-f ~-7~~

~!-200 A

1Al 1f

-800 OO I

-600 TWK, SECONDS

.TIME,SEC0lliDS

System (or Subys-im) caPonts Stress Ml+/-i :s ZG?

lney!i

_________l.,r Ccam~ent Support$

(o

+(or?) 3 FL)

Elst,.Mn1lla of e:t~

f (4)

Larger of Largor of (4) 10 Y (S) I.S(2)

((3)

(1 rn,;.t to c

G.70) SU.

(2)

EL; 3t niot to c

~

1 L.01 SL'.

(3) Li c-A L2 E-'

uadi It 11 1cnd with cu- -- ylalhl Point er~&1 to 2. 3 222 eid 8~ (but Mot to.

encead 0.70 1.,). r~otvy (4) h -a s Izm-tte mv~e based an c w dm hp o ector f I.$.

For Oriole beatHaS casee with different

LOADS USED FOR EQUIPMENT EVALUATION

1. OPERATING PRESSURE
2. FLUID FLOW
3.

OPERATING DEADWEIGHT

4. FLUID SLOSHING
5.

DESIGN BASIS EARTHQUAKE

6.

MECHANICAL LOADS

7.

MOTOR MAGNETIC PULL

8. FLUID PUMPING IN BEARING AREAS Ll

FEACTO". CWUM LC firl "Ed REPCi ECU A C0OX.ARIS011 X DIECTIO, IPPER E VATI(34

17. 5 UN4l DSPE;TRU

'*1 ODIFIED SPECT~rUM 4% AMPING 4% 4~AP ING

'a NODE 126X (LCUOP A

8.0 ND 2X (O P A

15.

NOD 12X(L A

Ip 6.0 7.r.

LOOP A

5.

2.0 14 I

PERIO (SEC~

PERIOD (SEC)*

MODIID PEIU MODIFIED SPECTRUM

4.

NOE 263Y (LOOF B7)

NODE 403Y (LOOP C) 3.C Tv6.0-u2.]0, L4.L0kr.K

0. Li~

Inl

)

C3 C

CD r.-

tC)C

-D prpr-fN o

I

.RESPONSE SPECTRA, COMPARISON I

REACTOR COOLANT PUMP

1.

X-DI RECTION 4% DAMPING:

LOWER SUPPORT ELEVATION UPPER SUPPORT FLEVATION

15.

yO KE~Y

.1.5.

UNr*ODI FI ED.

.....1.6 UNMODI FI ED LOOP A X LOOP A x LOOP B y LOOP B '~

LOOP CY LOOP C y

~

1

==I

-3L----,

.... z.z.

7_

~~~Qfl~~~

7 i77

-0

.02F

-0

.6

-. 6

.10 02..

PERIOD__

I IOD)

E

RESPONSE SPECTRA COMPARISON REACTOR COOLANT PUMP Y -DIRECTION 4% DA14PING

  • LWERSUPOR ELVATONUPPER SUPPORT. ELEVATION 15.C KEY KEY 3.1 UJNMOfDIFIED 1.8.ri

.01M~

FIED LOOP A Y.

z:z-OOA

~

LOOP B X--------

LOOP B X I

-- LOOP C X_

LOOP.C Y

__~!

4 77----_

74-_

7 7

7 2

.08

  • ER 10 (S__roDS_.

-J

____3 4~__V7

a RESPON4SE SPECTRtA Co~pRTSN REACTOR' COMMANT PLff' 2

S Z

- DIRECTioW LOWER SUPPORT ELEVATION UPPER.SUPPORT ELEVATI(M4

~

.LT U N~ D IF iE D 1-U-OIFE

__-0_

1.0

-II LOOP A ZLO LOOPB 8 z LOOP B Z 1 LOOP C ZMOCZ

~

-t7E7 77--_

-~ ~ ~ ~ ~ 7 M71:__ __

-1 057 7=I

ACCELERATION(G) o CD C)

0.

C 0.2 0.4 0.6 0.4 O,

20I C.

05.

0.8 co

o.

1.0

--- -e-

.rn ACi LAI

)

A

)nV

e.

.C

/,'

r-m Cm 1rn CD t/)

!0 L')

.0 ar CD a

o a

o ro rrio

.5 m6

."3.

1" rri"Cw m

00

ACCELERATION?1( G)

.N~

ACLRTONG MIS, c CD-2k

-M

I 17 i

rr Ct ro-

.0 I

  • s-21

~CD I~di ACCELRATIe~(o r

1. 9_

CS r"i

RESPONSE El ECTRA F(PR P 'EATOR

-Z DIRECTION1

.......................WER SUF EL:?VATJCHJ 4'1 Dunping 10.0 STEAM GENIERATOR MODEL

-L O O P. P.L NOD: 187 N

28.0l

-1 01 F IJ LOFIEP-C LOOP B K

SPECTi A LOCATION

4. 0 I

C) m

) u IRFACOR COLANT LOOPAU PEAIOI,'gC.)o C-1 MODFIE LOOP C 4.EC 5.C 8.0.

LI HOi tOD CZ35C DEi cn

-O3FE

RESPONSE SPECTR(A FOR_18 GF ELATOR

-Y DIRECTION LOGWER clJPW LEATION 10.04% Damping HOOE 187STEAM GENERATOR MODEL o

NODEFED

.1 LOOP

__j 6.0OoPIBM I

~4 SPECTRA LOCATIONLiI 4.04 10.0

8.

N~ODE 187 3.9*7 UNI -O D I F IFE D

-4.0 NODE 335 LOO B3.6

-~b~7-

-MODIFIE 6)- j6..

C%

LOOPC L.A 6.

J-1ODIFIED Li2.4 2

J

-j S2.1 LI (

1.2 I

Cl

~OD~

5n 0

..REACTOR SUPPORT. MAXIMUM STRESS ANALYSIS

SUMMARY

Description Stress or Surplus Load Allowable Marin Column Buckling 1137K 2496K

. 2.20 Stress in Bolts for Short Angles Associated with Support Shoe 29.6 ksi 35.7 ksi 1.21 Stress in Bolts for Long Anqles Associated with Support Shoe 16.8 ksi 26.2 ksi 1.55 Suprort Shoe Stress 9.70 ksi 9.0 si20.16 ksi 2.10-7 Reactor Vessel Support Attachment Stress 3i 23.52 ksi support Shoe Bolts 27 ST p Sh oe Blt s

20.6 ksi 26.4 ks; 1.28 11.7 ksi 40.25 ksi SUPPORT SHOE BOLTS COLUMn LONG ANGLE oI

.T-I SHEAR rL.

PLATES O

00000000000 00 0

iSHORT AGLE 00 ANCHOR BOLTS ooV

REACTOR COMPONENT ANALYSIS MAXiMUM RESPONSE Surplus Location

Response

Allowable Margin Fuel Assembly Grid Fuel-Barrel Impact 5.27k 7.3k 1.3 Fuel-Fuel Impact 2.87k 7.3k 2.5 Control Rod Guide Thimbles 61.9 ksi 91.1 ksi 1.4 Upper and Lower Core Plate Fuel Assembly Guide Pins 2.44 ksi 19.7 ksi 8.0 Rod Control Cluster Guide Tubes 0.23 in.

1 in.

4.3 Cure Barrei Flange Weid 12.9 ksi 50.7 ksi 3.9 Girth Weld at Neckdown. Region 4.90 ksi 50.4 ksi 10.3 Lowe r Corei Radial Support(a)

Ve!sel-Block Junction, Point "A" 22.0 ksi 55.9 ksi 2.5 VIock, Point "B" 16.6 ksi 41.e ksi 2.5 Ckevis, Point "C" 14.6 ksi 41.9 ksi 2.Q Clevis, Point "D" 37.8 ksi 41.9 ksi 1.1 Key, Point "E" 11.0 ksi 29.5 ksi I

2.

Cievis, Point "F" 13.5 kfsi 55.9 ksi 4.1 jC'trc! Rod Drive Mechanisms (CROM)

Hazd Adaptors 25.1 ksi 52.1ki 2.3 Rod Travel Housings 12.6 ksi 59.4 ksi 4.7 C DM Supt. Frame Tie Rod 6.09 ks.i 38.5 ksi 6.3 Tie Red Clevis at Min Cross Section 8.82 kri 38.5 ksi 4.3 iBolts in Base Attaching Tie Rod Clevis to Supt. Frame 18.56 ksi 33.5 ksi 2.0 Weld of Tie Rod Clevis to Bs Plate 5.0 ksi 19.25 ksi 3.8

AA

('op R di-ifSun or Stes Evluaic Pont

REACTOR COOLANT PUMP "COMPONENT ANALYSIS STRESS RESULTS MAXIMUM STRESSES Type of Stress Allowable Surplui Item Stress Intensity Stress Limit Margin Main Flange Bolts 6P11 8,800 1.44 Main Flange Shaft PM + P86 5,150 32,150 6.2 Seal Housing Shell PtM + P8 3, -C0 2, 150 8.2 H ousiflg Bolts PM10,650800 1.69

~t tain Flange PM P

19,000 54,000 2.3A j otor Stand Bolts I

fMain Flange 9, -,n rupMotor Stand I

S~ell at Upp,..r EndII (t = 10.4225 sac)

PM + P a 21,7:54 54000

.43 pfXlip Mvotor St~tmd

-it U,,ier End kt 10.4225 scc)

PM

  • 7,0 1.i3

-,."no C- :aier Aserrnbly F.-'

I 23&25 px!

I M.:;

r,

PUMP CASING AND FEET MAXIMUM STRESSES (Units.

PSI)

Surplus Description Stresses Allowable

Margin, Feet 14,254 32,150 2.25 Casing at Feet PM 2,064 21,450 10.4 Pm PB 3,032 32,150 10.6 FEET CASING FEET

REACTOR COOLANT PUMP MOTOR MAXIMUM RESPONSE Calctlated Surplus Descriptsi

Response

Allowable Margin Rotor Shaft Bending Stress 12,275 psi 52,500 psi 4.27 Rotor Shaft Center Deflection 0.0378 in.

0.125 in.

3.30 Vertical Loading on Rotor Core Assembly 34,1041 60,038 1.76 Flywheel Bolt Stress 570 psi 20,300 psi 35.6 00 ra:

IN Loopcc A shown) fm

~ ~<

J1

Maximumj, StLIbr Snub~erAl iwaLe Sup~V

SUMMARY

OF MAXIMUM LOADS ASSOCIATED WITH STEAM GENERATOR SEISMIC STOP STRUCTURAL MEMBERS Maximt.m Allowable Load Load Surplus (K)

(K)

Margin Loop A Upper Seismic Stop NW 440 562 1.28 Upper Seismic Stop SE 459 562 1.22 Lower Seismic Stop SE 493 552 1.14 Lower Seismic Stop NW 233 500 1.75 Loop B Upper Seismic Stop E 128 450 3.52 Upper Seismic Stop W 438 57r 1.32 Lower S2ismic Stop E 430 628 1.46 Lower Seismic Stop W 554 628 1.13 Loop C Upper Seirmic Stop SW' 510 870 1.1 Upper Seismic Stop NE 555 870 1.54 Lower Seismic Stop SW j

227 500 2.20 Lower Seismic Stop NE 691 920 123 F,

F~

STEAM GENERATOR HANGER SUPPRT SURPLUS MARGINS(&)

Item Stress/

Surplus Wrln 25.5-ksi 86.8 ksli 3.4 Nuts 6.9 ksi, 45.0 ksi 55 Hanger Support Structure Columns 3.6 ksi 15.1 ksl 4.17 Hanger Support Stnictura Beam Connection 227 kips M4 kirs1-:

SUPPor Structure 8em 20.0 k-e

33. 3 4
a. Load's include eir~mic nm

7771.

UPPER LUG UPPER ADDITIONAL LUG

.,-UPPER LUC3S VELDS.ADDED 1976 Lovwr Lug CLtS IC

.LA LV M MA YI, UM

,LLcWA6Zat1P L:

CrX-~O~INT LAO~'(~:LOAD tk? W noi-..

ROD)S 218 3 22 4

U-$'EPi LUGS 2 1 8 1.3 U-' 7R LUG WELDS 218

  • 351 1.81 CR" LUC IGS 218 263 1..1 CLE ViS PiNs2137 1.2 LOAD~~E~S 18 46 1.13 LOAD :.EEVE ~'~

218 232 1.2a LOAD S LE E VE IN fS 21640J 218 333 1.5

,70? C~4' 210 344 1438 TC~CA~WSLD,:,

213 246 1.13' ReaCtor CGCo'.

Piurp Hai,!zr (TYP -9)

~~~~~~~~~~...

ELUI EY1D V~ nq IrA1C~

I 11CI 67t ONEI C12 1-0, f

I (32 GO 1fs G,

MlACTOR COOLANT PUMP G-2A FRAME ANALYSIS. RESULTS.

0-ontro4Iing.

Controlling.

Load Surplus LOWd

Szrples, Mobr 1

t~ombination Miargin

[albe cibnation Margmn 132.25 24

8.

4.13 2

8 2.17 25 92.74

3.
3.

4.12 26 4-3.12 4

5 4.11.

27

9.

2.G?7 5

8

2. 60 23 7

7.42 6

8, 2.62

29.

3 7.36 7

3 3.31-30 9

3.55 8

8 2.29 31 9

1.73 9

8 2.25 32 3

7.14 10 4

1.61

.33 929 11 81.43 34 3

3.5 1 12 4

1.31 3333 13

.4 1.129 335 4

221. 2 I~~e 3

4 1.60 33 52=

16 3

2.29 981 1h 7 9

2.41 40 4

1.

189 2.23 41 4

1587 19 9

1.47 42 5

24. 3 219

.1.24 44 5

22

2. 01 7

2a.17 23 3

AP14Lo&1 2

3 4

8 9j Pump WcI~rs V

V V

V V

V V

V S.G. Sf&-Imic StopV V

S. Lwo~ra!, Supt.

N. Laturai Supt.V V/-

1D42 5

47 4

CP 5

6 olTI 3

46 51" 5

321 I o f C~

AL0 5

y.~y

.. ~zAJELtEATI 38 TYPICA!

%actr ~C-i~Pumnp G-23 Supjc+.r-W Prame kMojei

PEACTOR COOLANT PL'MP G-2B FRAME ANALYSIS RESULTS Controlling Controlling Load Surplus Load Surplus Member[aJ Combination Margin 1.lemberla]

Combination 1.argin 1

9 12.69 21 5

3.44 2

5 3.06 22 4

2.76.

3 5

3.44 23 6

2.88 4

6 1.82 24 6

2.71 5

4 1.85 25 8

1.14 6

4 10.17 26 8

1.48 7

6 3.81 27 8

1.12 8

6 4.03 23 9

1.98 9

6 2.81 29 9

1.C3 10 6

1.73 30 9

11.i 11 6

1.73 31 6

5.G2 i2 13 4

2.49 33 8

14 6

1.94 38 9

19.42 1@

6 2.62 37 9

.5.43 16 6

1.48 47 G

7

1.

1.-740 43 6

2.31 7

2.39 49 4

2.77 19 7

2.39 57 8

20 6

2.20 Lond ng Combnations Applie-d Loads 1

2 3.-

4 5

6 7

8 9

Pump Hangers

/V S.G. Seismic Stop

-1/

N.W. Lateral Supt.

N.E. Lateral Supt.

S.E. Lateral Supt.

-v

ELEVATI-1D-0 0

35 ELEVATION C-C D4B

-vj, (f

L Reacto

W REACTOR COOLANT PUMP G-.2C FPAMAE ANALYSIS RESULTS

~A~g~flI oadSur~usLoad Surpl'us Corninaios maginComrwicato 2

. 22 2.65 2

8 5.13

-2 L

3

4.

9.54 233 4

43.58 243 5

42.17 25 431 862.68 26 9

1.51 7

4 2.96 27 4

iso I

8 6

2.64 213 2

9 7

1.02 Z-3 11 4

2.52 14 2

125 1.65 3'2 4

14

2.

I j

8 22 I3' 4.7' 3124 4

A I

7 12 3,

7 4

63:

18 1

.'~4 LcedifI 22 3

4 7

V V4

II..

1.

1

.u 7

vF-4 FAIII Lp 6

i.q.

R~

F-

-SEISMIC EVALUATION STUDY LOGIC OwnerdDoun Westnghouse I

( etaied Buiding Detailed Recto ModelCoolant Latop M.odof

.LEluillj Model j c'Simphiiid RCL

-5'c Fin~ i~~ ti~n FinAl NOci~i.: F CL4 h RCL Model iks I:t C

I ofR -.5.0 Final Repoit 15

7 NOQE 21 NODE 4

Nocz I-.

-Figure 3.7.2-5 NODAL POI.,-

LOCATION'S For, RELLCUD REACTOR yULDI.N' MODEL

f ii M

I-,

z

7.

IN88nt=

AM~

ILM m A

M37 M.-2A-.

Maw.

junad IX0 "W

9=

so 1

SA 4',.EE..EEE WS KM90 3I.W 173 so.0 6

fmapg" fr

""dno oWAimWi n oOmsm

.hii:Iw,1410

~A0 2.3 17.00 1.44"3.1

4w If

funi-

-*1 MOW*

sum.

-f t.B - Regionpe Location of Top Support Plate 8600 48.930 5.68 Bow0 73,400 IL53

a.

Strese are for a combination of the maximum loads, on all three steam genrators.

e

-L

A 4v.

.~brasses IASATOA MAXMU SimL STRES5S W

Sto

-~

El 17-S 1194 3,W33 1,0 99.2 Local Shel Strs eat Upper Lateral El. 49'-6-3/4"11,970

.39,900 3.33 12,100 59,900 4.99 a, Stres. are. for &"combination, of -the maximum loads on all three steam generatorsM MM~I 40tT I

S ALL EL 5-'

U

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