Text

Responds to NRC 790312 Request for Addl Info Re Systematic Evaluation Program Seismic Reviews
	ML17244A450
Person / Time
Site:	Ginna
Issue date:	04/03/1979
From:	White L; ROCHESTER GAS & ELECTRIC CORP.
To:	Ziemann D; Office of Nuclear Reactor Regulation
References
	TASK-02-04, TASK-03-06, TASK-2-4, TASK-3-6, TASK-RR NUDOCS 7904100323
	Download: ML17244A450 (66)
	v • d • e

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:7900100323 DOC SEDATE: 79/00/03 NOTARIZED:

NO DOCKET FACILe50~244 ROBERT EMMET GINNA NUCLEAR PLANTE UNIT ii ROCHESTER G

05000244 AUTH'AME AUTHOR AFFILIATION NHITEiL~ D ~

ROCHESTER GAS 5

ELECTRIC CORP'EC IP o NAME REC IP IENT AFFILIATION ZIEMANNFDSLN 'PERATING REACTORS BRANCH 2

SUBJECT:

RESPONDS TO NRC 790312 REQUEST FOR ADDL INFO RE SYSTEMATIC EVALUATION PROGRAM SEISMIC REVIEWS ~

DISTRIBUTION CODE:

ASSIS COPIES RECEIVEDtLTR i.ENCL' SIZE I TITLEl GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING.

I IC NOTES:~cg:~ FT(l '

(~~rn

'ECIPIENT COPIES RECIPIENT COPIES ID COOK/NAME LTTR ENCl ID CODE/NAME LTTR ENCL ACTION!

05 BC ORB 7

7 INTERNAL:

0 KG FILE 1

I 1

ORE PERF BR 17 ENGR BR 19 PLANT SYS BR 21 EFLT TRT SYS 1

2 2

1 1

02 NRC PDR 10 TA/EDO 16 AD SYS/PROJ 18, REAC SFTY BR 20 KEB 22 BRINKMAN EXTERNAL: 03 LPDR 23 ACRS 0~

NSIC 16 16 1

1 APR l 1 tgyg TOTAL NUMBER OF COPIES REQUIRED'TTR 38 ENCL 38

1' k

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//iRSZW ROCHESTER GAS AND ELECTRIC CORPORATION gaa< Owlsa O

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IIAtt l

o 89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE, JR.

VICE PRESIDENT T E LE P II0IIE AREA CODE 7IE 546.2700 April 3, 1979 Director of Nuclear Reactor Regulation AttentiOn:

Mr. Dennis L. Ziemann, Chief Operating Reactors Branch No.

2 U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Systematic Evaluation Program Seismic Reviews R.E.

Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Ziemann:

In your letter of March 12, 1979 you addressed the review of Ginna Station which is to be performed by the NRC Seismic Review Team.

In order to facilitate your review, you requested that cer-tain information in addition to that found on the Ginna docket be supplied.

With the assistance of Westinghouse Electric Corporation and Gilbert Associates, we are gathering the information necessary to respond to your questions.

The attachment. to this letter pro-vides the information we have gathered to date.

We are continuing our investigation and expect to have further information available for discussion during the planned site visit.

Very truly yours, Attachments

~ 904 y0 0 S>~

uestion 1

A list of specific pieces of mechanical and electrical equipment necessary to ensure integrity of the reactor coolant pressure boundary and to safely shut down the reactor and maintain it. in a safe shutdown condition during and after a postulated seismic event. is requested.

~Res ense:

The following basis are considered appropriate for evaluating the seismic classification of Ginna structures, systems and components.

These bases, and the listing of equipment in the accompanying tables, reflects plant,-specific requirements, rather than relying on the more general LWR standards currently in effect.

A comparison to these standards

can, however, be made.

1.

Seismic classification will be restricted to structures,

systems, and components required for safe

shutdown, and to maintain RCPB integrity, and to prevent other design basis accidents which could potentially result in offsite exposures comparable to the guideline exposures of 1'OCFR Part 100.

These latter systems and components include, for

example, the steam, feedwater and blowdown piping up to first, isolation valve, and the spent, fuel pool, including fuel racks.

(Also included, though not explicitly defined in the list of seismically classified equipment, are all structures, systems and components not required to function, but whose failure could irreversably prevent the functioning of required safe shutdown equipment or cause a

design basis accident.)

Seismic design of these items will ensure a very low probability of failure in the event of an SSE.

System boundaries, for purposes of seismic re-evalua-tion will be considered to terminate at the first normally closed, auto-close, or remote-manual valve in connected piping.

2.

Safe shutdown is defined as the capability to control residual heat. removal under all plant conditions resultant from a seismic event (with the consequential loss of function of non-seismic equipment) and a loss of off-site power.

Safe shutdown may be the maintenance of extended hot shutdown condition, or a gradual cooldown to cold shutdown conditions.

For Ginna, safe shutdown assumes gradual cooldown and depressurization in the event of an SSE.

Safe shutdown would be in the following manner:

a) b)

.,c) d)

e)

Maintenance of pressurizer level to ensure continued natural circulation capability (this does not require use of the pressurizer heaters.

Loss of heaters would result in a net heat loss through the pressurizer insulation, resulting in a slow natural depressurization).

RCS subcooling would be maintained through use of the auxiliary feedwater system and atmospheric dump valves.

Auxiliary feedwater control to the SG's (on-off pump control) and depressurization by local manual control of the atmospheric dump valves.

Inventory and reactivity control is by use of the charging pumps taking suction from the RWST.

Letdown is to be isolated.

Boric acid concentration is such that the reactor is

,maintained subcritical with the most reactive rod stuck out of the core.

Charging is via the reactor pump seal integrity while the RCS is above 450'F.

Below 450'F, the seals do not require either CCW or seal injection.

RCS depressurization is accomplished by shrink-ing of RC due to contraction cooldown, then fillingthe pressurizer.

This would condense the steam in the pressurizer by mixing with the relatively cooler RCS water entering the pressurizer.

Service water (SW) is the seismic source for the auxiliary feedwater system (credit is taken for the standby AFW system only).

Sufficient inventory is available in the steam generators following a loss of off-site power to maintain the RCS within safe limits until the operator can align the SW system to the suction of the standby AFW pumps (time is conservatively estimated at 10 minutes).

Eventually, when the RCS is depressurized and cooled to about 350'F and 360 psig, the RHR system can be placed in operation.

Containment access may be necessary to manually open a

RHR suction valve (700 or 701) and an RHR discharge valve (720 or 721) in the event of a single failure (such as a failed valve or power supply).

Containment access is considered acceptable.

e ~

a There is no need to expedite this operation, since the AFW source is essentially infinite, heat loss from the pressurizer is very gradual, and residual heat can continue to be removed by operation of the steam generators.

g)

Component. cooling water must be established to the RHR cooler and the RHR pump oil coolers concurrent with this operation.

Although categorized as seismic I in the original plant design, the CVCS is not required for safe shutdown except 'as noted in 2.c) above, and therefore, most portions need not be classified as seismic I for purposes of seismic re-evaluation.

As noted in l. above, systems required only for accident mitigation, such as the safety injection, containment

spray, containmen isolation, contain-ment purge, HEPA and charcoal filter systems are not classified as seismic I for purposes of seismic re-evaluation.

The only component cooling water functions required, per 2.g) above, are cooling the'HR HX and pumps.

All other CCW functions will not be classified as seismic I for purposes of seismic re-evaluation (such as cooling the SI and charging pumps, CI valves, etc.).

However, system integrity must be maintained to the extent needed to perform the safety function.

The waste disposal

system, though largely considered seismic I in the original plant design, will not be classified as seismic I for purposes of seismic re-evaluation.

This is because of the relatively low potential consequences that could result from failures in the system.

Ventilation systems for safe shutdown equipment rooms are designated seismic I in the FSAR.

However, for purposes of seismic re-evaluation, no ventilation systems in the plant except the standby AFW building coolers are designated seismic I.

Other cooling

systems, such as for safety equipment

rooms, the control room, the battery rooms, etc.

are not needed immediately, since the rooms will heat up slowly. It is considered acceptable to provide for the use of portable air conditioning/ventilation units to effect cooling of safe shutdown equipment in the-time required.

Containment access might be required only for a short time; protective clothing and breathing apparatus could be worn by personnel for the time required.

8..

The FSAR has designated portions of the fire protec-tion system as seismic I.

However, for purposes of seismic re-evaluation, and consistent with Appendix A to BTP 9.5-1 for operating reactors and RG&E's "Fire Protection Evaluation",

GAI Report No. 1936, Narch 1977, no credit for seismic design of the fire protection system is needed or claimed.

These criteria are presently under review by the NRC SEP Branch (under Topics III-1 and "Safe Shut-down").

The attached tables list the specific individual equipment, which sould be classified Seismic Category I, consistent with the above criteria.

Figure 1-1 provides the area designations for equipment.

FIGURE l-l Screen House Area;".5 D/G Annex Area 84 AVT Building Turbine Building Area 87 Service Building intermediate Building Area

.-'. 3 Reactor Containment Building Area Nl Control Building Area 88 Auxiliary Building Area

-.'.2 Aux.Bldg.

dd-;tion Area

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5 SYSTEM NAME 8I NUMBER MANUFACTURER gI HODEL NO.

E UIPMENT DESCRIPTION SIZE 8 WT LOCATION FUNCTION SEISHIC INPUT LEVEL QUALIFICATION OBE H ~

SSE REFERENCES load load Charging LCV 112B Pump 1A, 1B, 1C 286 323 348B 356 Continental (Butterfly) w/ASCO Solenoid LBX 831616 (Globe)

(Globe)

(Diaphragm)

(Diaphragm) 4II 60 gpm 2ll 2lI 2 II Area 92 Area g2 235'"

Area g2 Area g2 Area g2 Area 82 Suction from RWST Inventory and Reactivity Control Hanual valve used to isolate normal charging Hanual valve used to isolate alternate charging Manual valve used to isolate unneeded portion Manual valve used to isolate unneeded portion 52/ 52 E Spec 676370, Rev 0

3.0/2.0 E 676241, Rev.

1 3.0/2.0 E-676241, Rev.

1 RWST Pulsation dampener Pittsburgh-Desmoines Steel Co.

335,000 gal Area 82 26.5'35 I 8'I 81.0'igh Area g2 integrity only integrity only

.08/.08

.20/.20 Pittsburgh-DesHoines Steel Co. No. 56063-02 2/26/78 Seal injection Commercial Filter filters Dwg. 9 AOV427 2ll Area 82 integrity only integrity only

.60/.60 E-Spec 676436, Rev.

1 Regen heat Exchanger Letdown Orifices Sentry Dwg j7 2ll Area g2 235'"

Area f/I 235'ntegrity only

.52/.52 E-Spec 676224 (ASHE III)

='Horizontal/Vertical

V tV o

SYSTEH NAkE S NUHBER HANUFACTURER 6 HODEL NO.

E UIPHENT DESCRIPTION SIZE Sr VT LOCATION FUNCTION SEISHIC INPUT LEVEL QUALIFICATION OBE N V SSE REFERENCES load load Charging AOV 200 A Copes-Vulcan (Globe)

Zll Area gl 235'solation letdown 3.0/2.0 E-676270 AOV 200 B

AOV 202 AOV 310 Copes-Vulcan (Globe)

Copes-Vulcan (Globe)

Copes-Vulcan (Globe) 21 ~

Zll 3/4o 236'36 Area gl isolation letdown isolation letdown integrity (fail closed) 3.0/2.0 E-676270 3.0/2.0 E-676270 3.0/2.0 E-676270

4.

SYSTEH NAHE 6 NOSER HANUFACTURER

& HQDEL NO.

E UIPHENT DESCRIPTION SIZE 6 WT LOCATION FUNCTION SEISHIC INPUT LEVEL QUALIFICATION OBE H/~

SSE REFERENCES load load RHR HQV 700 MOV 701 1QII 1 QADI Area 81 Elev. 236 Area gl Elev.

236 RHR Suction from A Hot Leg RHR Suction from A Hot Leg 3.0/2.0 E-676258, Rev.

2 3.0/2.0 E-676258, Rev.

2 Pump A, B Pump:

Pacific SVC-6L 2500 gpm Area g2 Hotor:

Westinghouse 445 TS Elev.

219'BDP, Class B Insula.

ME Suction from A Hot Leg

.52/.52 E-676228, Rev.

0 Heat Exch A, B Jos.

Oat.

Dwg. 4807, Rev.

4 24 HBTU/hr Area 82 Elev 236'"

RHR Suction from A Hot I,eg

.52/.52 E-676228 (ASHE III-tube)

(ASHE VIII-shell)

HCV 624 HCV 625 HOV 720 HOV 721 MOV 850A Continental (Butterfly) v/ASCO Continental (Butterfly) w/ASCO Solenoid)

Limitorque SHB-00 Reliance Operator, Class B Insul.

8lI gll 1QII 10n IQII Area 82 Area /f2 Area 81 Elev 246 Area gl Elev 236 Area g2 Elev 230'HR Discharge to B Cold Leg MR Discharge to B Cold Leg RHR Discharge to B Cold Leg ME Discharge to B Cold Leg Integrity only req'd for ME System Operability 3.0/2.0 E-676270 3.0/2.0 E-676270 3.0/2.0 E-676258, Rev.

2 3.0/2.0 E-676258, Rev.

2 HOV 850B Limitorque SHB-00 Reliance Operator, Class B Insul.

10" Area 82 Elev. 230'ntegrity only req'd for MIR System Operability 3.0/2.0 E-676258, Rev.

2

SYSTEH NAHE & NUMBER HANUFACTURER Sr HODEL NO.

E UIPHENT DESCRIPTION SIZE 8 Vl'OCATION FUNCTION SEISHIC INPUT LEVEL OBE N/~

SSE load load QUALIFICATION REFERENCES MiR HOV 704A HOV 704B HOV 857A HOV 857B FCV 626 Limitorque SHB-00

Reliance, Class B Insulation Continental (Butterfly) v/ASCO Solenoid)

Limitorque SHB-00

Reliance, Clas B Insulation 101 l 10'I 6tl 6 II 6 II Area 0 Kiev 226'"

Area 0 Kiev 226< 5n Area g2 Elev 236'rea g2 Kiev 236'rea 82 Integrity only req'd for MR System Operability Integrity only req'd for ME System Operability Integrity only req'd for MR System Operability Integrity only req'd for RHR System Operability Integrity only req'd for RHR System Operability 3.0/2.0 E-676258, Rev.

2 3.0/2.0 E-676258, Rev.

2 3.0/2.0 E-676270 HOV 852A HOV 852B Limitorque SHB-1

Reliance, Class B

Limitorque SHB-1

Reliance, Class B

Insul Insul 6II 6lt Area gl Elev 236'rea gl Elev 236'ntegrity only req'd for RHR System Operability Integrity only req'd for MR System Operability 3.0/2.0 E-676258, Rev.

2 3.0/2.0 E-676258, Rev.

2 HCV 133 Copes-Vulcan (Globe) 2II Area gl Integrity only req'd for RHR System Operability E-676270 SYSTEM NAME S NUMBER MANUFACTURER 8 MODEL NO.

E UIPMENT DESCRIPTION SIZE 8 WT LOCATION FUNCTION SEISMIC INPUT IEVEL QUALIFICATION OBE SSE REFERENCES load load Component Pump 1, 2

Cooling Ingersoll-Rand SSD Motor:

Westinghouse 444TS TBDP, Class B

Insulation 3000 gpm Area g2 270<

Cold Shutdown Cooling

.52/.52 E-676370, Rev.

0 CCW Heat Exch A,B Atlas, Dwg. B-1260 25 MBTU/hr Area g2 281'"

Cold Shutdown Cooling

.52/.52 E-676454 Surge Tank MOV 817 MOV 738A MOV 738B 772A 772B 2000 gal gtl 10" 10" 3II 3u Area g2 Area g2 Area 82 Area g2 Area g2 Area g2 Maintain CCW Pressure Boundary Isolate CCW to RCP, reactor supports, excess letdown heat exch.

CCW to RHR Heat Exch.

CCW to IUE Heat Exch.

Isolate CCW from Sample Heat Exchangers Isolate CCW from Sample Heat Exchangers 3k.O/2.0 E-676258, Rev.

2 3.0/2.0 E-676258, Rev.

2 3.0/2.0 E-676241, Rev.

1 3.0/2.0 E-676241, Rev.

1

1.

>+% L E UIPHENT DESCRIPTION NAME 6 NUMBER MANUFACTURER g> HODEL NO.

MOV 4616 HOV 4734 HOV 4735 MOV 4670 Service Pump 1A, 1B, Worthington 20M-500-WZ 1C, 1D Hotor:

Westinghouse 509-UPJ{ ABDP, Class F Insul HOV 4615 SIZE g> WT 20" 20>>

14>>

14" 10" LOCATION Area 95 Elev 253'"

Area iJ4 FUNCTION Isolation Valves Between Two Loops Isolation Valves Between Two Loops To CC Neat Exch To CC Heat Exch Isolation to ensure diesel cooling SEISMIC INPUT LEVEL OBE M V.

SSE load load QUALIFICATION REFERENCES HOV 4664 MOV 4663 4651A 4652A HOV 4609 Containment Fan Cooler 3 Reactor Com-partment Cooler 1A>

1B>

1C>

1D 41>

4>>

Area JJ5 Isolation between seismic/

nonseismic portions Isolate air conditioner water chiller from required seismic portion Isolate discharge of water chiller Isolate supply to travelling screens Integrity only Integrity only

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)av 3$

Vrf 3

lu 3

3 ~v 12 SYSTEM UIPMENT DESCRIPTION NAME Sv NUMBER MANUFACTURER 8 MODEL NO.

Service MOV 9629A Water MOV 9629B SIZE Sc Wf 4II 4l 3 LOCATION Area g6 Z71 3

Oll Area 86 271'"

FUNCTION Supply to SAFW System Supply to SAFW System SEISMIC INPUT LEVEL OBE

@Ve SSE load load QUALIFICATION REFERENCES Spent Fuel Pit Neat Exch Cool Spent Fuel

v I

I,

13 SYSTEM NAME 8 NUMBER MANUFACTURER 8 MODEL NO.

E UIPMENT DESCRIPTION SIZE 8 WT LOCATION FUNCTION SEISHIC INPUT LEVEL OBE

@Ve SSE load load

(}UALIFICATION REFERENCES Hain AOV 3516 Steam AOV 3517 Valve:

Atwood Morrill 30" Operator:

Chicago Fluid A3I Solenoids:

Laurence

110114W, 125434W Valve:

Atwood Horrill Operator:

Chicago Fluid A3I Solenoids:

Laurence 110114W1 125434W Area d3 Elev 278'rea g3 Kiev 278'ain steam isolation Valves Main steam isolation Valves 3410 3411 HOV 3504 HOV 3505 3508-3515 Fisher 476AA Fisher 476AA Rockwell 604NY Rockwell 604NY Crosby (HB-BP86) 6ll 6 II 6II 6 II 4"x6" Area 93 Elev 278'rea 93 Elev 278'rea d3 El. 278 Area g3 El. 278 Area g3 El. 278 Power Operated Relief Valve used for Cooldown by Hanual Operation of Handwheel Power Operated Relief Valve used for Cooldown by Manual Operation of Handwheel Isolation from TDAFP Isolation from TDAFP Hain Steam Safeties E-Spec 676279

SYSTEM E UIPMENT DESCRIPTION NAME S NUMBER MANUFACTURER 6 MODEL NO.

Feedwater AOV 5737 AOV 5738 SIZE 6 WT LOCATION Area g3 Area g3 FUNCTION Blowdown Isolation Blowdown Isolation SEISMIC INPUT LEVEL QUALIFICATION OBE

@Ve SSE REFERENCES load load

l, lv

a'

~ SYSTEM NAME 8 NUMBER MANUFACTURER Sr MODEL NO.

E UIPHENT DESCRIPTION SIZE 8 Vf LOCATION FUNCTION SEISMIC INPUT LEVEL OBE H/~,

SSE load load QUALIFICATION REFERENCES Standby HOV 9629A Aux Feed-Water System HOV 9629B 4II 4I~

Area d6 27 1 I Oll Area g6 271'"

Supplt SAFW System Supplt SAFW System Standby AFW System Seismic Analysis is Referenced in 10/6/78 (letter)

See also 7/28/78 RGE Submittal to NRC Standby AFW System Seismic Analysis is Referenced in 10/6/78 (Attached)

See also 7/28/78 RGE Submittal to NRC Standby Aux Feedpump 1C>

1D HOV 9701A MOV 9701B Ingersoll Rand w/GE Motor 200 gpm Type R Class B Insulation 3 II 3II Area 96 271'"

Area g6

," 271'"

Area g6 271'"

Aux Feed to S.G.

1A, 1B Standby AFW System Seismic Analysis is Referenced in 10/6/78 (Attached)

See also 7/28/78 RGE Submittal to NRC Standby AFW System Seismic Analysis is Referenced in 10/6/78 (Attached)

Response

Spectrum (RRS).

He then includes a require-ment, in all Class IE equipment specifications for qualification of the equipment, in accordance with IEEE 344-75, and the RRS.

The qualification may be in the form of analysis, particularly in the case of cable tray and conduit supports, and equipment mounting anchors.

Complex electrical equipment however is usually qualified by test.

This general method of qualification depends on the existence of published industry standards (e.g.

IEEE 344-75) which establish acceptable qualification procedures, together with appropriate quality assurance standards which can meaning-fully be referenced in the specifications.

At the time that the Ginna Nuclear Plant was designed and con-structed no industry framework of standards for seismic qualifi-cation existed.

Certain critical components were required by specification to be capable of withstanding the maximum seismic loads postulated for the plant site as shown in Table 1-1.

Nost components in the Class IE electric power distribution system are designed to withstand forces due to electrica1 faults, which are much larger than the inertial forces due to a severe seismic event.

In addition, many components used at Ginna have undergone subsequent seismic testing or analysis using inputs equal to or greater than those postulated for their location.

It has been the policy of RGB Engineering, when making modifica-tions at Ginna Station, to require seismic qualification in accordance with the current standard when possible.

In practice this means that when major Class IE components, which are inde-pendently anchored to Category I structures, are designed and procured, it is done in accordance with the current seismic standard.

This has resulted in an evolution of seismic qualification in Ginna electrical equipment to increasingly severe standards including IEEE 344-1975.

The attached table (Table l-l) shows the major components in the Class IE electrical system and the basis for seismic qualifica-tion.

S

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TABLE 1-1 CLASS lE ELECTRICAL SYSTEMS S stem Com onent I.

EMERGENCY POWER SYSTEM:

Basis for Seismic uglification A.

B.

C.

D.

Low Voltage (600V) Switchgear (excluding unit transformer)

(Westinghouse DB-15, 25, 50 & 75 Breakers).

Motor Control Centers (Westinghouse type W)

MOV operators (ac/dc)

Vital 120 VAC Post construction testing Post construction testing and analyses in accordance with IEEE 344-1971.

Upgraded by analysis to IEEE 344-1975.

Post construction testing.

E.

1.

distribution pnl 1A & 1C 2.

inverters Solidstate Controls 3.

const voltage xformers Inc.

125 VDC Power System 1.

125V-60 cell batteries (Gould) and racks 2.

battery chargers Diesel Generators (Alco/Westinghouse)

Post construction testing Installed in 1978, qualified by test in accordance with IEEE 344-1975.

Design specification; 0.52g simultaneous horizontal and vertical.

Currently under review.

Design specification; 0.47g simultaneous horizontal and vertical acceleration.

G.

Reactor Bld. Cable Penetrations (Crouse-Hinds)

Post construction testing.

r p",

h'l q I

c'

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1-~:

P

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23 TABLE 1-1 CLASS 1E ELECTRICAL SYSTEMS (Cont'd)

S stem Com onent H.

Conduit Supports

& Tray Supports I.

Electrical Equipment Anchors II.

SAFEGUARDS INSTRUMENTATION AND CONTROL:

1.0 Transmitters (Barton, Foxboro) 2.0 Reactor Trip Switchgear (DB 50) 3.0 Main Control Board (Wolf and Mann) 4.0 Reactor Trip System Racks (A/D conversion)

Basis for Seismic uglification Currently under review.

Currently under review.

Post construction testing.

Design specification; 0.52g simultaneous horizontal and vertical acceleration.

5.0 Protective Relay Racks.

(SI and reactor trip logic)

Design specification; 0.52g simultaneous horizontal and vertical acceleration.

6.0 Safeguards Racks (ESF actuation output) 7.0 Control Switches (Westinghouse type W2 and OT2) 8.0 Under Voltage Relay Cabinet (Westnghouse CV-7 Sc MG-6) 9.0 Safety Related Indication Design specification; 0.52g simultaneous horizontal and vertical acceleration.

Post construction testing.

Currently under review.

Question Ra Describe in detail the proceduze for calculating earthquake structural reponse of containment shell, internal structures and other category I structures, including dynamic model, determination of natural periods (or frequencies),

combinations of horizontal and vertical responses, method of analysis, calcula-tion of stresses in concrete and steel at base of containment building for given shears, forces and

moments, etc.

~Res ense:2 The following discussion presents the design details or appropriate Ginna FSAR references for the Category I seismic structures at the Ginna facility.

The area designators refer to the areas shown on Figure l-l attached to the question 1 answer.

Reactor Containment Buildin Area ¹1 Analytical models and methods used to determine the seismic response of the containment shell, distribu-tion of stresses through the structure, combination of seismic response components, and evaluation of structural element capacity to substain applied loads are presented in the following sections of the FSAR:

<<a~f n RRRRsttjt tprwe-s2see>> p',"2'~'e~+~~~4 W~JRst s

a) Section 5.1.2.3 b) Section 5.1.2.4 c) Section 5.1.2.5 d)

GAI Report 1683 Vol. 4, FSAR Dated Sept.

6, 1968 Interior Structures Mechanical Design Bases Seismic Design Classification Detailed Design "Design of Large Opening Reinforce-ment for Containment Vessel" i

J s

s The natural period of simple harmonic motion for a cantilever beam was determined for the interior concrete structures of the reactor building and the corresponding acceleration obtained from the seismic response spectra of the FSAR (Fig. 5.1.2-7 and 5.1.2-8).

This acceleration value multiplied by the total mass acting at the center of gravity, produces base moment and shear.

Flexural properties of the wall systems were used for evaluation of stiffness properties.

The resultant shears and overturning moments were distributed through the reinforced concrete structure in proportion to the stiffness of the structural element and structural element design capacity evaluated.

Individual walls and floor slabs were

designed for the concentrated seismic reactions of the attached major components.

Overhead crane support structures within containment were evaluated to obtain natural periods of simple harmonic motion in the two horizontal directions.

Equivalent horizontal seismic forces were then obtained by applying the corresponding acceleration from the seismic response spectra to the mass of the crane.

Vertical response of the crane and crane support structure was taken as the peak response acceleration of the response spectra.

Vertical forces were obtained by applying the peak acceleration to the mass of

crane, crane support structure and lifted load.

Auxiliar Buildin Area 52 Steel superstructures above elevation 271'-0" were evaluated for equivalent horizontal seismic loads based upon either the maximum spectral response or the spectrum value corresponding to the first harmonic frequency of the structure.

(Ref:

FSAR Section 1.2.9)

The spent fuel pool is a Class I Structure as referenced in FSAR 9.5-6.

As part of the spent fuel storage rack modification, the east wall of the spent fuel pool has been re-analyzed for seismic forces.

Seismic forces con-sidered the hydrodynamic response of the fluid within the pool and the response of the wall as a plate.

Response

accelerations were obtained from seismic

~

response spectra developed for the Auxiliary Building, in accordance with USNRC Regulatory Guide 1.60 and 1.61.

Structural capacity of the wall was evaluated against, the load combinations specified in SRP 3.8.4 and found satisfactory.

Since the east wall of the spent fuel pool is the thinnest concrete section, the remaining pool walls are also structurally capable.

Intermediate Buildin Area 03 The bracing system of the Intermediate Building is common to the structures of the Turbine Building, Service Building, Auxiliary Building, and Facade Structure.

The bracing was checked to demonstrate that it was sufficient to resist equivalent seismic load components from the above structures.

Equivalent static loads were based on either the maximum spectral response or the spectrum value corresponding to the first harmonic frequency of the structure (FSAR 1.2.9).

26 Diesel Generator Buildin Area 44 The diesel generator building has concrete shear walls and steel framed roof structures.

The seismic design of the concrete shear walls considered both inplane and normal equivalent seismic loads.

Seismic accelerations were taken as the peak acceleration of the seismic response spectra (Ginna FSAR) for 5%

damping.

The steel roof framing was designed for a horizontal equivalent SSE seismic load.

The equivalent load was taken as the mass of the roof structure and superimposed loads times the peak seismic response spectra for 2.5% damping.

Column foundations were designed for 'additional 20% of axial load to include seismic effect.

Screen House Area 45 Since the screenhouse/service water building was not addressed in detail in the FSAR a more detailed description of the design is presented here.

General Descri tion The screenhouse/service water building supplies plant water needs for the circulating water system, service

.water system and the plant fire water system with the exception of yard hydrants.

Mater is brought into the screenhouse through a submerged intake structure and a reinforced concrete tunnel excavated in bedrock.

If necessary a redundant water supply is available by

-utilizing the discharge canal and opening the sluice gate on the north wall of the screen house.

Service water would be discharged to the discharge canal or through the emergency service water discharge on the south side of the auxiliary building.

The entire building (both below and above grade) is a

seismic Category I structure.

The building is com-prised of two super structures, one for the service water system and one for the circulating water system (the screen house portion).

The service water portion houses four Class I service water pumps and Class I electric switch gear.

The screen house portion houses the traveling water screens and circulating water pumps.

Below Grade The entire screenhouse-service water building is founded in or on bedrock with the exception of the basement of the electric switchgear portion which is founded approximately four feet above bedrock. 'ince the building is founded in bedrock the basement will not realize any spectral acceleration and the seismic

27 loading is equivalent to the ground motion of 0.08g and 0.20g.

The basement is designed to be dewatered.

The full height of the wall is designed for an external hydro-static pressure plus a seismic load equal to a per-centage of the dead load of the wall and the hydro-static pressure.

For the portion of the wall below grade and above bedrock an active earth pressure based on a saturated soil weight is applied.

Internal walls, such as pump baffles and the wing walls between the traveling screens were designed for a full height hydrostatic pressure on either side plus a seismic load due to the water movement during a seismic event.

Above Grade 1)

Design loadings:

a)

Dead Loads 1)

Built-up roof 2)

Piping hung from roof 3)

Siding 14 psf 10 psf 5 psf b) Live loads 1)

Snow load to roof (New York State 40 psf Bldg.

Code Para.

C-304-5) 2)

Wind load a)

Walls (SBC Para.

304-4) b)

Roof (SBC Para.

304-5)

Down Up 3)

Crane loads (inc-live, dead impact) 9 support points a)

Horizontal to runway 20%

4,000¹ b)

Horizontal

= to runway 10%

2,000¹ c)

Seismic loadings 1) 0.08g ground motion 9 working stress 2) 0.20g ground motion 6 yield stress 3)

Using welded steel connections a

1% critical damping was assumed 20-psf 5 psf 17 psf 20,000¹ 2)

Design loading combinations for analysis:

a)

Dead loads

+ snow load + crane load 9 working stress b)

Dead load + snow load + crane load + wind load 9 1.33 times working stress c)

Dead load

+ snow load

+ crane load + 0.08g vert.

seismic

+ 0.08g E-W horiz. seismic 8 1.33 working stress d)

Dead load + snow load

+ crane load

+ 0.08g vert.

seismic

+ 0.08g N-S horiz. seismic 6 1.33 working stress

28 e)

Dead load

+ snow load

+ crane load

+ 0.20g vert.

seismic

+ 0.20g E-W horiz. seismic I yield stress f)

Dead load

+ snow load + crane load

+ 0.20g vert.

seismic

+ 0.20g N-S horizontal seismic I yield stress 3)

Design Analysis:

The service water portion of the screen house consists of four rigid frame bents in the east-west direction with bracing for wind and seismic loads in the north-south direction.

The roof system is designed as a horizontal truss to transmit horizontal seismic loads to the frame columns and through the bracing to the foundation.

a)

Member stresses due to all loading combina-tions were calculated using equations for statically indeterminate rigid frame analysis and virtual work analysis for roof trusses and bracing.

b)

Seismic loads were calculated by analyzing both the 0.08g and 0.20g ground accelerations for vertical load, horizontal east-west load and horizontal north-south load.

c)

The vertical seismic loads were determined as follows:

1)

Member sizes for the bents were assumed and the vertical static deflection of'the bent was calculated with the entire dead and live load contributing to the bent applied on the girder.

2)

Calculate the period.

3)

Knowing the

% damping (FSAR Table 5.1.2-1) and the period the spectral acceleration was obtained from the response spectra (FSAR Figure 5.1.2-7

& 5.1.2-8).

4)

The applied seismic load is then the product of the structural mass and the spectral acceleration.

d)

Horizontal Seismic loads:

East-West Direction.

The total horizontal seismic load is carried at the mass center of the structure and if the mass center

& center of rigidity concide, each bent. will deflect an equal amount and carry a force directly proportional to its relative rigidity, provided the roof acts as a diaphram.

However in this case none of these conditions existed and the following was done.

29 e)

A total seismic l'oad was calculated using the total-mass and four times the stiffness of one bent.

This force was distributed in direct proportion to the relative mass carried by each bent.

(The seismic load was calculated as in c above assuming the equivalent mass applied horizontally at, the top of the bent)

The center of mass and the center of rigidity were located and an applied moment was calculated equal to the total force times the distance between the mass center and center of rigidity.

This moment would then be changed to an additional force on each bent which is proportional to its relative rigidity and distance from the center of rigidity.

The combined loads acting on each bent represent the applied loads for the east-west horizontal seismic event.

Since the total structure must act as a

unit the final deflection of each bent must be the same and since each bent has the same columns and girders, they each have the same rigidity.

Therefore, the resisting force at each bent will equal the total force divided by the number of bents.

~

Because of the distribution of forces and the fact that the mass center and center of rigidity do not coincide, the resisting truss will have a tendency to twist.

This requires additional resisting forces per-pendicular to the bent.

These forces are transmitted through the roof by means of a bracing truss and transmitted to the founda-tion through the columns.

Horizontal Seismic Loads:

North-South Direction.

The north-south horizontal seismic loads were calculated in the same manner as item d above with the exception that. the loads were applied to the roof truss at the panel points of the roof purlins.

Also as

above, the loads were distributed through the horizontal. roof truss and transmitted to the foundation through the vertical column bracing.

SAFS Auxiliar Buildin Addition Area C6 The SAFS Auxiliary Building addition was analyzed to obtain the seismic response to three simultaneous, independent, mutually perpendicular acceleration time

30 histories which enveloped the response spectrum of USNRC Reg.

Guide 1.60.

The analysis considered soil/caisson interaction and soil liquefaction potentials.

Equivalent seismic forces obtained from the analysis were distributed through the reinforced concrete structure in proportion to the stiffness of the structural elements.

For complete. discussion of this analysis refer to RGSE's letters to Mr. A. Schwencer

,dated 5/20/77 and to Mr. D. L. Ziemann dated 7/28/78.

Control Buildin Area 48 Steel framing for the floor and roof is supported by concrete shear walls on two sides (West S South) with the remainder of the structure framing into the turbine building.

Original design considerations for seismic loads were based on Operating Bases Earthquake as follows:

a)

Structural steel columns were designed for flexural moments resulting from horizontal load equivalent to 10% of the axial load applied at mid-span of the column.

b)

Concrete walls above grade were subjected to a horizontal reaction normal to the wall and applied at mid-span treating wall as cantilevel fixed at base.

Magnitude of the equivalent seismic load was 10% of the wall weight.

Emer enc Dischar e Structure Walls of the emergency discharge structure below grade were designed for an equivalent seismic lateral earth pressure for OBE and SSE conditions.

Zero period acceleration of the ground level seismic response spectra was used to obtain equivalent. vertical seismic forces for design of the structures.

31 uestion 2b Explain in detail the dynamic coupling between all Category I and non-Category I structures.

~Res ense:

, Dynamic Analysis of the buildings (other than the screen house) were performed by isolating individual plane frames from the adjoining structures.

The first harmonic period of the framing system was determined and the appropriate acceleration value was obtained from the plant ground response spectra.

This accelera-tion value was later used to determine equivalent static loads for analysis.

The service water and screenhouse superstructures are immediately adjacent to one another but have separate column lines where they meet.

The only physical connec-tion between the two superstructures is the architectural metal siding.

It was assumed that a seismic event.

would rupture the siding liner and that the two super-structures move independently of each other.

During a seismic event it was assumed that. the super structures would move out of phase of one another and the column lines between the two buildings are separated by an amount equal to the sum of the calculated deflection for each building.

uestion 2c Summarize the procedures used for determining the seismic input to piping and equipment.

R~es ense:

Seismic input for original plant piping and equipment was based on the ground seismic response spectra given in the FSAR, Figures 5.1.2-7 and 5.1.2-8.

A peak seismic acceleration was extracted from the curves based on the damping coefficient applicable to a com-ponent and that component',s natural frequency.

If natural frequencies were unknown, as with vendor designed equipment, peak accelerations were specified.

The methods of analysis are summarized in the NRC Phase I report enclosed in NRC correspondence from D.L. Ziemann to L.D. White dated 3/9/79.

The original piping design and*piping modification analyses are summarized on Tables 2c-1 and 2c2.

- 33 PIPING SYSTEM DESCRIPTION TABLE 2c-l Ori inal Class I Pi in PLANT LOCATION GROUND RESPONSE (see fig. for area location)

SEISMIC INPUT LEVEL OBE SSE g load g load SEISMIC ANALYSIS METHODS CVCS CCW Sampling System S.W.

M. Steam F.

W.

Stm.

Gen.

Bldn.

Class I Portion Class I Portion Class I Portion Class I Portion Class I Portion Class I Portion Class I Portion Class I Portion 0'8 0.08 0.08 0.08 0.08 0.08 0'8 0.08 0.20 0.20 0.20 0.20 0.20 0.20 0.20 0.20 Methods of analysis described in Phase I

Report Methods of analysis described in Phase I Report Methods of analysis described in Phase I Report Methods of analysis described in Phase I Report Methods of analysis described in Phase I Report Methods of analysis described in Phase I Report Methods of analysis described in Phase I Report Methods of analysis described in Phase I Report

QPanisc ". PIPING SYSTEM TABLE 2c-2 Pi in Modifications DESCRIPTION PLANT LOCATION GROUND RESPONSE SIZE & WT (see fig. for area location)

SEISMIC INPUT LEVEL OBE SSE g load g load SEISMIC ANALYSIS METHODS Pressurizer Safety Valve Valve Discharge Piping 1972 4u 6n From Press.

to 8" Header Containment Area Ol 0.08 0.20 Max. horizontal level 8 elev. 278.4 Stress limitations per B31.1-1967 Main Steam 1973 SIS Mod.

1974 Main F.W.

1974 10" 14 II Reanalysis of M. Stm.

from Steam generators 1A 6 1B to isolation valves in header From Accum.

B to Primary Loop "A" Modification to Anchor SI-9.

From Header to Stm Gen.

1A 6 1B - outside cont. reanalysis Int. Bldg.

Area 53 Containment Area 51 Int. Bldg.

Area 57 0.08 0.08 0.08 0.20 0.20 0.20 Seismic analysis of a lumped mass model by response spectra method.

Response

curves used were Flormo response spectra (6/

/72) with 0.5%

damping.

Max. horizontal G level 4.5 9 elev. 227-11.

Stress limits per B31.1-1967 Seismic analysis of a lumped model by response spectra method.

Response

curves used were Flormo response spectra (5/23/72) for elev. 312.0, max horizontal G level 0.9 Stress limits per B31.1-1973.

See Main Steam (1973)

4

WpWA, 4'k.4 PIPING'YSTEM SIZE & WT DESCRIPTION PLANT LOCATION GROUND RESPONSE (see fig. for area location)

SEISMIC INPUT LEVEL OBE SSE g load g load SEISMIC ANALYSIS METHODS S.W. - 1974 20" Seismic analysis of buried S.W.

piping 0.08 0.20 Analyzed to meet re-quirements of ASME III-1974.

CVCS - 1975 SIS - 1975 2 II 3/4 II Letdown Line Relocation of S.I.

6( W.D.

lines after valves 882 6

884 Containment Area ¹1 Auxiliary Bldg.

Area ¹6 0.08 0.08 0.20 0.20 See SIS Mod (1974)

Seismic analysis of a lumped mass model by response spectra method.

Response

curves used were response spectra (10/31/74) for elev. 253'-0",

Max.

horizontal G level 1.1 with 2/ damping per Reg.

Guide 1.61.

Stress limits per ASME III 1974.

S.W.

1975 20n 121I 6ll 3 II Redundant S.W.

Return in Aux.

Bldg.

Auxiliary Bldg.

Area ¹6 0.08 0.20 Seismic analysis of a lumped mass model by response spectra method.

Response

curves used were response spectra (10/31/74) for elev. 282'-0" max.

G level 4.1, damping per Reg.

Guide 1.61.

Stress limits per ASME III - 1974.

Stm.

Gen.

Blowdown 1976 Aux. F.W.

1976 2 It 3 II Relocation of Stm.

Gen.

1B Blowdown Valve Addition of 4 new valves Int. Bldg.

Area

¹7'nt.

Bldg.

Area ¹7.

0.08 0.20 See Main Steam (1973)

Stress limits per ASME III - 1974.

See Main Steam (1973)

uestion 2d Described the method of seismic analysis for major equipment (e.g.,

steam generator, pressu'rizer, reactor internals,

tanks, pumps, etc.).

~Res ense:

a)

STEAM GENERATOR The series 44 Steam Generator has been evaluated to a set of generic loads including seismic.

The seismic loads are based on envelope horizontal response spectra for 1% equipment damping shown on Figure 2d-1 for the OBE and SSE.

The generic curves are based on an envelope of floor response spectra of eleven (11) plants with W NSSS.

The analysis was performed to update the 44 Series Steam Generator stress report to show compliance of Section III of the ASME Code for a revised set, of external loads.

The analyses in the update revision of the Series 44 Steam Generator are either by direct comparison with the 51 Series Steam Generator or revision of the original 44 Series Steam Generator report for the revised external loadings.

The dynamic analyses are by the response spectrum method with the staem generator idealized by lumped masses interconnected by three-dimensional beam elements (Figure 2d-2).

In addition, generic nozzle loads for the primary and secondary nozzles were developed.

Table 2d-1 compares the primary nozzle loads initially specified in the E-Spec with the loads used for the generic analyses.

37 TABLE 2d-1 SEISNIC NOZZLE LOADS FX FZ NZ E-Spec Update Seismic Analysis 109 107 75 6221 6353 9650 OBE DBE 800 800 300 15500 13500 13500 1000 1000 400 20500 18000 15000 All loads 2

Units or KIP, IN-KIP

I

'I

'I ~

C II' II ~

~I I)

~

gi t I, II, I

~

~ 'I1

~

0 cps 1

2 8

9 20 30 35 OBH 1 g 1

3.

3 2

2

~

1 1

SS.E 2

g, 2

6 6

2 2

For vertical motion use 2/3 Damping =

1%

SSE V.

2

'Q OBE

0 0

10 15 20 25 30 FREQUENCY (CPS) fij Figure 2d-3.- )forizontal Response Spectra

r -~

(

1 L

1 IIIII II

/

I

'H

~

I I'I II I

I I

I 16 I\\

~

I20 II22

(~25 I'27 l30 l32

,'35 I 37 s40 42 45 48 12 14 19 29 34 44 47 52 II54

)

I c:I:

'57 J

r C

N~d"~

)r 18

~50 II, I

l I

I 10 13 IIII 17 I

21/

I 23,'

26<

I 28l' 31l 33,'6ll 38t l

I 431 40 53 58 I

,) 't le FS.pur.e 2d-2 See;im CI'ne> stol H;lctiuiv:<t$.:a] llv<lel a,

Node numbers

~'39-

b)

CONTROL ROD DRIVE MECHANISMS There are 29 model L-106 full length control rod drive mechanisms and 4 model 121J001 part length control rod drive mechanisms attached to the reactor vessel head adapters for the plant.

The seismic analysis of both the full length and part length mechanisms consists of two phases.

The first phase involves comparing the results of a computer analysis of the'mechanism for internal pressure and thermal loads with the ASME Section III stress allow-ables.

The results of the comparison are used to derive the alllowable seismic bending moment for the mechanism.

In the second part of the analysis, a

lumped mass beam model of the CRDM system is used to calculate the bending moments at various locations on the assembly resulting from seismic loading.

These calculated bending moments are compared with the allow-able seismic bending moment for the mechanism.

The seismic design basis used for the plant's drive mechanisms was 0.8 g's in both the horizontal and vertical directions.

For these loads the bending moments throughout the mechanism were below the allow-able bending moments.

The results of the second phase of the analysis were used in the design and analysis of the CRDM seismic support scheme.

c)

REACTOR INTERNALS The RG&E reactor internals is a standard 12-foot, 2-loop assembly (Figure 2d-3) that has been qualified on a generic basis.

The qualification analysis uses a

linear response spectra analysis with a lumped mass beam finite element model as shown on Figure 2d-4.

The input for the analysis is the response spectra for the Kanasi Plant. as shown on Figure 2d-5.

Two cases are considered in the analysis:

first, where the modal contributions are combined by the square root of the sum of the squares and;

secondly, where the modal contributions are summed by the absolute method.

Two models were evaluated, one horizontal and rotational stiffness representing the soil as shown on Figure 2d-4 and the second model where a fixed base is assumed.

In both cases, 5% damping has been used for the concrete and 1% for the internals.

In the vertical direction, a

single degree-of-freedom

model, uncoupled from the horizontal direction, was used.

Stresses were obtained by adding horizontal and vertical responses absolutely.

~

~ ~

I

~

P I

I CONTROL ROD DRIVE MECHANISM UPPER SUPPORT PLATE CORE BARREL SUPPORT COLUMN INLET IIOZZLE UPPER CORE PLATE THERMAL SHIELD REACTOR VESSEL c) 0 O 0 0 O

~ca a tier',o <y ro

~~

go leg HOLD-DOWN SP RING ALIGNMENTPIN CONTROL'OD GU ID E TUB E CONTROL ROD DRIVE SHAFT CONTROL ROD CLUSTER

( WITHDRAW N )

OUTLET NOZZLE BAFFLE RADIAL SUPPORT SUPPORT PLATE Figure 2d-3:'Reactor Vessel and Internals

'OWER CORE PLATE FLOW MIXER PLATE CORE SUPPORT COLUMNS SUPPORT PLATE INSTRUMENTATION THIMBLE GUIDES V

PPPI<<% ~I VP

BUILDING K

= RADIAL SUPPORT SPRING CONSTANT rs K

~ = ROTATIONAL GROUND SPRING CONSTANT rot K

= TRANSlATIONALGROUND SPRING CONSTANT t

VESSEL BARREL FUEL ASSEMBLIES

ELEV.

165.0 INCHES K

= 592.E 5 LB/IN rs K

= 2.1337E14 LB/IN rot.

ELEV.

~

139. 9 INCHES ELEV.

106.1 INCHES K

= 17.967ES LB/tN t

Not to Scale ELEV.

0.0 INCHES Figure 2d-4: Mathematical Model 0

1 0.8

~t 0.6 V

~e 0.4 0

NEN KANSAI CURVE (.3g) NOV. 10, 1967 0.2 0.1 0.01 0.1 1.0 PERIOD (SECONDS)

Figure 2d-5=;

Seismic Average Acceleration Spectrum Design Earthquake for 1% Damping

d)

REACTOR VESSEL Seismic analysis of the reactor vessel was peformed by applying a steady-state acceleration to the piping and calculating the resulting nozzle reactions.

A stress analysis was then conducted by Babcock 8 Wilcox to verify that the nozzle stresses were acceptable.

Loads were provided to BSW by Westinghouse for stress calcula-tions for the following three (3) cases:

1.

Design Seismic plus thermal loads 2.

No-loss of Function Seismic plus Thermal Loads 3.

Design Seismic plus Thermal plus Interaction Loads Accelerations of.08g and

.20g were used for design and no-loss-of function seismic, respectively.

e)

PRESSURIZER Westinghouse manufactures five (5) different pressurizer

models, ug to a maximum of 1800 ft capacity (Ginna has an 800 ft pressurizer).

All five models rest on identicial support skirts.

For conservatism,3the seismic analysis was conducted fyy the 1800ft (heaviest) model.

Two cases were analyzed:

an OBE and DBE, however, for both cases the DBE acceleration of.48 g horizontal and

.32 g vertical were for evaluation.

- The accelerations were applied statically at the center of gravity of the pressurizer.

The heaters for the pressurizer have been qualjjjed on a generic basis for the 51 Series Pressurizer.

The heaters in the 800 CF pressurizer are shorter than those qualified in Ref. (2), but are otherwise identical.

The qualification procedure uses an equivalent starting load of 37.5 g for the DBE and 30 g for the OBE.

The fundamental frequency of the heater rods are greater than 33 Hz.

Ref.

(1) F. J. Mendolla and P.

P.

DeRosa, Rochester Gas and Electric Pressurizer Stress

Report, September

1969, Westinghouse Electric Corporation Tampa Division, Tampa, Florida.

Ref.

(2) P.

G. Smith, 51 Series Pressurizer Generic Seismic Analysis, August 1974, Westinghouse Tampa Division, Tampa,.Florida.

45 f) REFUELING WATER STORAGE TANK Analysis of Refueling Water Storage Tank was performed in accordance with TID-7024, taking into account the possible dynamic effects resulting from the slashing of the water.

Shell stresses and support stresses were limited to those permitted in the pressure vessel codes and the structural steel standards of AISC.

~

I

~.

uestion 2e Provide a summary of the changes to the facility as described in the FSAR which affect seismic Category I structures and systems.

~Ree once:

All modifications to the facility as described in the FSAR have been reported to the NRC in Semiannual or Annual Operating Reports in accordance with 10CFR 50.59.

However, many of the modifications do not significantly affect the seismic capability of Category I structures and systems.

The following is a listing and brief summary of modifications which the seismic review team may be interested in reviewing further.

In

general, modifications are performed in accordance with codes and standards in effect at the time the design is performed.

Additional information on these, and other modifications, is available and can be discussed during the site visit.

EWR 1026 Penetration Testing Piping at penetration 332 was modified to facilitate leak rate testing and was resupported.

EWR 1057, Relocation of Charging Pump Filter Vent A 3/4" vent line off the charging pump filter was relocated.

EWR 1079 Instrument Bus Supply Modification The inverters were replaced, static switches were added and regulated backup supplies were provided for instru-ment buses 1A and 1C from the safeguards buses.

A regulating transformer was replaced in instrument bus 1C.

A regulatory transformer was replaced and the normal supply was changed to a safeguards bus for instrument bus 1D.

EWR 1100 Battery Room Air Conditioning Ductwork through the battery rooms was added and seismically supported.

EWR 1431 Auxiliary Feedwater Pump Flow Control A flow-control bypass is being added in the discharge line of each of the two motor-driven auxiliary feed-water pumps.

EWR 1594 Spent Fuel Pool Cooling System A new cooling loop, including pump and heat. exchanger, will be added to the existing cooling system for the spent fuel pool.

EWR 1601 Turbine Driven Auxiliary Feedwater Pump Valves Valves will be added and piping will be modified and resupported in the steam supply line for the turbine driven auxiliary feedwater pump.

EWR 1657 Control Building Ventilation The outside air intake and ductwork to the battery rooms and mechanical, equipment.

room will be modified and resupported.

EWR 1660 RCS Overpressurization Protection System An overpressure protection system was added to provide pressure relieving capability while the primary system is in a water solid condition.

This modification is being reviewed by the NRC.

EWR 1661 Boric Acid Piping Modifications The modification consisted of installation of a new section of boric acid piping between the boric acid tank room and the pipe header above automatic isolation valves

826A, B,

C, and D in the Auxiliary Building.

Also included was replacement of the remaining sections of 8 inch, schedule 10S pipe in the boric acid tank room with schedule 40S pipe.

EWR 1675 Letdown Diversion Line Modifications The modification consisted of installation of a new section of 2 inch piping in the letdown diversion line to the CVCS holdup tanks between valves LCV 112A and 1103D.

Also included was a new section of pipe to tie the concentrates tank transfer pump piping to the letdown line.

EWR 1682 Penetration Testing Piping at five penetrations was modified to facilitate leak rate testing and piping at the penetrations was resupported.

EWR 1836 Pressure Shielding Steel Diaphragm Additional barriers are being constructed between the control building and the turbine building and between

the diesel generator annexes and the turbine building.

Included in the design analysis is a seismic evaluation of the control building, the diesel generator annexes and the three column lines of the turbine building adjacent to the control building and diesel generator annex.

(See RG&E letters of February 6,

1978, August 25, 1978 and October 11, 1978.)

EWR 2273 Steel Shielding, East Side of Control Room Steel plate was erected on the east wall of the control room.

Seismic analysis was performed in accordance with FSAR methodology.

GSM-1 Addition to the Auxiliary Building GSN-16 Standby Auxiliary Feedwater System Two additional auxiliary feedwater pumps were added in a new structure located to the south of the auxiliary building.

These modifications are described in a letter dated Nay 20, 1977 from L.D. White, Jr. to Mr. A. Schwencer, USNRC and in a letter dated July 28, 1978 from L.D. White, Jr., to Mr. D.L. Ziemann, USNRC.

GSN-3 Piping Support Modification for CVCS Letdown Line Inside Containment The letdown line inside containment was reanalyzed and resupported.

GSM-5 Reheater Relief Valve Discharge Piping Modifica-tions Supports were added for the relief valves and the affected, portion of the facade structure was reanalyzed.

GSN-7 Addition of Four Valves to the Auxiliary Feedwater System Two check valves in the discharge line to the feedwater line and two motor-operated valves in the cross over line between the two motor driven pump discharge lines were added.

Piping was modified and resupported.

GSM-21 Station Service Water System Modification-Addition of Redundant Return Line A redundant, discharge line was added for the service water system.

The line services the component, cooling heat

exchangers, the spent fuel pool heat exchanger and the auxiliary building (safety related pump) motor coolers.

The line extends south from the'uxiliary building to a seismic category I discharge structure.

GSM-34 Blowdown Isolation Valve Relocation The easternmost blowdown isolation valve and associated piping (about 8 feet) were relocated.

Check valves in the High Head SI Lines A second check valve was installed in each of the four high head safety injection lines to the coolant legs.

(See the RG&E letter of June 3, 1971).

- Steam Generator and Reactor Coolant Pump Anchor Bolt Replacement The subject, anchor bolts were replaced (see the RG&E letter of December 21, 1970).

Main Steam Safety and Relief Valve Supporter The supports for the main steam safety and relief valves were reanalyzed and modified (See RG&E letters of June 5,

1972 and March 15, 1973).

Pressurizer Safety Valve Supports The support for the pressurizer safety valves were-reanalyzed and modified (See RG&E letters of June 19, 1972 and March 15, 1973).

- Charging Pump Filter A pressure pulse filter was installed in the common discharge of the three charging pumps.

SM74-12 B Loop Accumulator Support Modification The B loop accumulator line (to the A reactor cold leg) was reanalyzed and resupported.

SM74-15 Addition of Valve in Service Water Line A 6" valve was added to the B loop service water line in the B diesel room.

An amplified response spectrum (0.5% damping ratio) is shown in Figure 2 of "Additional Information on Seismic Design of Class 1 Piping," 6/9/79, RG&E Unit. No. 1, USAEC.

Explain how this response spectrum was developed and used.

R~es ense:

The response spectrum shown in Figure 2 of "Additional Information on Seismic Design of Class 1 Piping," dated 6/9/79 for RGSE Unit No.

1 was modified from the design ground response spectrum to factor in building effects.

Specific description of the procedure used in its construction is not available.

However, a comparison of this response spectrum which is contained in a W

Proprietary report with a similar response spectrum developed in the same report indicates that the following formula may have been used:

where s max maximum acceleration of structure

,which is given by the ground response spectrum.

resonant amplication of structural response.

This may be read off Figure 2f-1*.

For periods less than 0.08 seconds, C is dependent of period and a reasonable assumption is to alter it to C's follows:

C' 1

forT< 002

0. 06 The response is assumed to be at. resonance with the fundamental mode of the building.

The response spectrum was applied to the piping analysis using the normal mode, response spectrum technique.

32 M

ze CO 0-C/>

cO bJ ~

Cc 0

IcQ.

DV l6 D~

CO O LU I-CD U

8 O

NOTE:

I0.0$

0.5$

THESE CURVES APPLY FOR PERIOD T > 0.08 SECOHDS

( f < l2.5 cps).

THE MAXIMUM ACCELERATIOH OF THE EgllPMEHT AT RESOHAHCE IS GIVEN BY:

(A,);= C(p, P,) (A,)

WHERE Ae

= ACCELERATIOH OF E)UIPMENT-As ACCELERATIOH OF STRUCTURE STRUCTURAL DA%'IHG (P s)

~ + ~

0 0

6 EQUIPMENT DAMPING (Pe)

(PERCENT CRITI CAI )

8 lo ll Ffg<<~2'~=,l-hIaximum Response. of Equipment in Resonance a@i tie a One Degree of Freedom Structure y ~

~

~ ~ ~i

~ e~

r av khlQA

(s

~

e 52 uestion 2

Provide in detail the seismic design of the RHR system performed in 1969.

R~es ense:

With the exception of the RHR line from the Reactor Coolant Loop A to the containment penetration, the seismic analysis of the RHR system is based on a static analysis approach using as input the peak of the 0.5%

design ground response spectrum for 20% g.

The allowable stress for a combined normal load and the seismic load is 1.2 S,

where S is taken from the power piping code, USAS B31.1.1.0-1967, para.

119.6.4, For the RHR line from the Reactor Coolant Loop A to the containment penetration, two response spectrum analyses were performed.

One analysis is based on the input of the 0.5% design ground response spectrum discussed above in the answer to question F.

Both analyses have the same mathematical model.

Since the location of this line runs low in the containment, the use of the design ground response spectrum was considered to be adec{uate.

Also, the analysis using the amplified response spectrum has shown that the portion of RHR line when assumed to be located in the containment at the steam line elevation is not considered to be over-stressed.

A detailed description of the RHR line analysis is contained in the AEC Docket No. 50-244 "Additional Information on Seismic Design of Class I Piping" dated June 9,

1969 and a summary of the analysis is as follows:

I I A

Summary of "Additional Information on Seismic Design of Class I Piping" Dated June 9,

1969 The reactor coolant system was qualified by tests using a 1/4 scale plastic model.

The tests were conducted with a sinusoidal input at 0.2 g level for the vertical direction and each of the two hori-zontal directions, independently.

All results on dynamic characteristics were evaluated on the basis of the model parameters.

These included natural frequencies, mode

shapes, magnification factors, and dynamic stresses.

Conversion factors and estimates were then made to establish the dynamic characteristics of the actual system which includes the reactor

vessel, steam generator, reactor coolant pump, reactor coolant piping, prssurizer, and accumulator.

In addition, each subsystem was also evaluated.

This includes the steam, feedwater, and residual heat removal systems.

Based on the conversion factor calculated for the model versus the prototype, the reactor coolant system natural frequencies were found to be 1.2, 1.6 and 3.58 Hz.

These frequencies correspond to motions primarily from the steam generator and the reactor coolant pump.

The stresses determined for these frequencies are 2,880 psi, 7,680 psi, and 790 psi, respectively.

For the auxiliary piping systems, the predominate modes are 3.06, 3.42 and 4.94, respectively, for Main Steam, Feedwater and Residual Heat Removal.

The stresses for these modes were computed to be 19,500; 19,300; and 80,600 psig, respectively.

Snubbers were-recommended to minimize the high stresses.

\\

As for the reactor vessel, the natural frequency was calculated to be 46 Hz for the prototype, using the conversion factor deter-mined in the report.

The rz.gidness of the component, was also con-firmed in the test.

Finally, accumulator and pressurizer have been both determined in the test report to have natural frequencies of 23.72 Hz and 3.04 Hz, respectively.

No stresses have been determined in the report.

uestion 2h Summarize the design criteria and methods used for cable trays and duct work including lateral supports, etc.

~Res ense:

HVAC ducts, duct supports and air handling equipment supports were analyzed.

Stresses and deflections of the components were analyzed and the supports designed using the 8% and 20% seismic response spectra with 1%

of critical damping.

Simultaneous horizontal and vertical ground motion was considered.

Class I Systems within containment, the auxiliary building and inter-mediate building were evaluated.

Response

of the duct systems were analyzed by consider-ing the most conservative situation.

Cable trays were designed for a maximum sag of 1/4" when designed as a simple beam with a uniform load of 100 lbs/ft and a 200 lb point load at any location.

55 uestion 3

It is requested that you review the Category I structure

.listing and verify that the systems and components identified in 1 above are not. housed in any other structure, (e.g., turbine building, service building, etc.)

~Res ense:

The NRC Category I structure listing is correct with the following changes and additions:

1)

Change."control room" to "control building" 2)

Add "screenhouse and house service water building" 3)

Add "cable tunnel" The design of the Category I structures

'is addressed in the response to question 2a.

ML17244A450

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