ML20207N787

Forwards Proposed FSAR Pages 3.7.4-2,3 & 4,Table 3.7.4-1, Sheet 1,Table 6.3.2-1,Sheet 1,Tables 6.2.1-2,9.1.4-3 & 13.4.2-2 & Table 16.3-4,Sheets 6 & 7.Changes Will Be Included in Next FSAR Amend

ML20207N787
Person / Time
Site:	Vogtle
Issue date:	01/08/1987
From:	Bailey J GEORGIA POWER CO., SOUTHERN COMPANY SERVICES, INC.
To:	Youngblood B Office of Nuclear Reactor Regulation
References
GN-1288, NUDOCS 8701140420
Download: ML20207N787 (11)

Text

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, =, Georgia Power Company

' Pbst Offica Box 282 Waynesbora Georgia 30830 Telephone 404 554-9961 404 724 8114 Southern Company Services,Inc.

Ibst Office Box 2625 Birmingham, Alabama 35202 Telephone 205 870-6011 gp January 8, 1987 Director of Nuclear Reactor Regulation File: X7N16 Attention: Mr. B. J. Youngblood Log: GN-1288 PWR Project Directorate #4 Division of PWR Licensing A U. S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 V0GTLE ELECTRIC GENERATING PIANT - UNITS 1 AND 2 PROPOSED FSAR CHANGES

Dear Mr. Denton:

Georgia Power Company has completed a review of the VEGP FSAR and the proposed Technical Specifications. As a result of this review certain discrepancies were found which will require a change to the FSAR. The changed pages (attached) are: 3.7.4-2, 3 & 4; Table 3.7.4-1, Sheet 1; Table 6.3.2-1, Sheet 1; Table 6.2.1-2; 9.1.4-3; 13.4.2-2; and Table 16.3-4, Sheets 6 & 7. These changes will be included in the next FSAR amendment.

If your staff has any questions concerning the proposed changes, please do not hesitate to contact me.

Sincerely,

. l.

. A. Bailey Project Licensing Manager JAB /caa Attachment xc: R. E. Conway NRC Regional Administrator i R. A. Thomas NRC Resident Inspector J. E. Joiner, Esquire D. Feig ,

B. W. Churchill, Esquire R. W. McManus  ;

M. A. Miller (2) L. T. Cucwa B. Jones, Esquire Vogtle Project File '

G. Bockhold, Jr.

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8701140420 87010a PDR o

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VEGP-FSAR-3 3.7.4.2.1 Strong Motion Acceleroraeters SMAs produce a record of the time varying acceleration at the sensor location. This data is used directly for analysis and comparison with reference information and may be converted to -

response spectra form for spectral comparisons with design parameters.

Each sensor unit contains three accelerometers mounted in a mutually orthogonal array. Accelerometers have their principal g' axes oriented identically, with one horizontal axis parallel to W the major horizontal axis assumed in the seismic analysis.

One SMA is located in the free field at approximately 500 ft from the containment. A second SMA is located in the Unit 1 containment tendon gallery such that it measures the input vibratory motion on the basemat. A third SMA is located on the Unit 1 containment building operating floor. A fourth SMA is located on the pressurizer support in Unit 1. A fifth SMA is installed on the control building basemat. A sixth SMA is located in the auxiliary building floor on level 1. A seventh SMA is installed on the basemat of the auxiliary building. The eighth and ninth SMAs are on the slab floors of the fuel handling and diesel generator buildings.

triaxial TwoVseismic triggers are provided to start the 3MA recording system. One is located adjacent to the SMA in the containment tendon gallery and the other on the containment operating floor. *k- ^ --+

• ri c;er 1 -^r-iti"- te Jerti 21 ::ccler tier

--d the ::::nd tri;;cr cen::: hieri:1 heri:ent:1 metier A magnetic tape recording and playback unit is provided for multiple channel recording and playback of the signals from the SMAs. The data recordings include an additional recording channel which contains a timing signal.

The recording and playback system is housed in a panel furnished for these instruments and devices necessary for system testing, annunciating, calibration, and control. This l()

panel is located in the control room.

3.7.4.2.2 Peak Recording Accelerograph Each SMA sensor unit contains three PRAs mounted in a mutually O orthogonal array. The PRAs that are mounted directly on equip-ment or piping have one axis coincident with the principle equipment or pipe axis. The other PRA has one horizontal axis parallel to the major horizontal axis assumed in the seismic analysis. They are self-powered units.

3.7.4-2

VEGP-FSAR-3 One PRA is mounted on the reactor coolant pump motor at el 210 ft. A second PRA is located on the steam generator at el 185 ft.

The third PRA pump in the auxiliary is on a residual heat removal (RHR) building.

f, The fourth PRA is mounted on

(') the nuclear service cooling water (NSCW) piping outside the auxiliary building at el 220 ft. Data from the PRA must be 19 manually retrieved following an earthquake and are used in the

- detailed equipment.

investigations for particular structures, systems, and O' 3.7.4.2.3 Seismic Switches and ve rb*c.al s efpo *nt of023 one triaxia c switch, with a horizontal setpoint of 0.17 g, s installed adjacent to the SMA in the containment tendon gallery on the basemat. The second seismic switch is located a onsetpoint vertical the containment of 0.32 g.operating floor at el 220 ft and hasj These devices actuate visual / 30 annunciators in the control room. 'a horizonbl set oint of o P 3.7.4.2.4 Response Spectrum Analyzer

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( A triaxial response spectrum analyzer is used to analyze all the SMAs once a seismic event has occurred. It will also initiate an alarm if an operating basis earthquake (OBE) limit has been exceeded. The response spectrum analyzer consists of a microprocessor-based computational unit and a printer unit.

The computational playback system. The unit is operated in conjunction with a tape analyzer computes the response spectrum from the recorded data. The alphanumeric printer unit prints response acceleration versus frequency in a hard copy form.

l 3.7.4.2.5 System Control Panel j

A panel located in the control room houses the recording, i

playback, and calibration units which are used in conjunction l

with the SMA senso'rs to produce a time-history record of the earthquake.

It also contains signal conditioning and display equipment associated with the response spectrum analyzer, ~

visual annudciators associ'ated with the seismi6'_switche's and l30 response spectrum recorder, audible and visual annunciators l wired to display initiation of the SMA recorder, and the power t

supply components for the equipment contained within the ' panel'.

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Amend. 19 9/85 3.7.4-3 Amend. 30 12/86

VEGP-FSAR-3 3.7.4.3 Control Room Operator Notification Activation of the seismic trigger causes an audible and visual annunciation in the control room to alert the plant operator l30 that an earthquake has occurred. These seismic triggers initiate the SMA recording system at acceleration levels which l3 are elightly ir.,.1_ dinghigher then ir.duccd - thetion; citr; e perted fr;; hech;rru .d 1:;::1,tr:ff

ur :: cu:1 @in conformanc.c.

1. ' R GJ 12 cl: vat-r-rs, pscpic. --j t r'intry. These initial setpoints are based on experience in existing plants and may be changed once significant plant operating data have been obtained which indicate that a different setpoint would provide a better SMA system operation.

The peak acceleration level experienced on the containment tendon gallery is available immediately following the earth-quake. This is obtained by playing back the recorded SMA data from this location and reading the peak value for this data from the printer unit.

Response spectra from the free field tendon gallery and the operating floor are available in the control room, following an earthquake, on readout equipment suitable for comparing the measured response spectra with the OBE and safe shutdown earthquake (SSE) response spectra.

As discussed in paragraph 3.7.4.2.3 visual indication is given 1f the seismic switch setpoints have been exceeded. g l30 3.7.4.4 Comparison of Measured and Predicted Responses

. The plan for utilization of the seismic data includes both the i function of the operator and engineering to evaluate the

effects of an earthquake on the plant. For a detailed I

description of the data flow refer to figure 3.7.4-2.

Initial determination of the earthquake level is performed immediately after the earthquake by comparing the measured l()

response spectra from the containment tendon gallery with the OBE and SSE response spectra for the corresponding location.

If the measured spectra exceed the OBE response spectra, the l

plant will be shut down and a detailed anclysis of the earthquake motion will be undertaken. ()

i After an earthquake, the data from the seismic recorders and recording instruments are reviewed. See figure 3.7.4-2. The l data from these instruments are analyzed to obtain the seismic accelerations experienced at the location of major Category 1 structures and equipment. The measured responses from the 3.7.4-4 Amend. 30 12/86 l

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4 TABLE 3.7.4-1 (SHEET 1 OF 2)

SEISMIC MONITORING INSTRUMENTATION REQUIREMENTS Triaulal Time-History Triaxial Response Trioxial Peak IM Lrumente tlon_Lgea t ion Aqpa s --m e1@ Triaxial Seisele

'oct um *ece-dn h eleeuctaufe Sw i to ri senseic t ri aae r RG 1.12 SRP PG 1.12 SRP RG 1.12 SRP Rea. Rh y(Qf RG 1.12 SRP RG 1.12 SRP Rh h YLQf _ Reg. h hP Reg. h YLQf Ree, h yffef 1 Free Field Requiresent - Locetion

1. Free field - 500 ft I free structure 1

1 *I 1( 1 )(e><*

30 < tr$

II. Inside Containment O Requirement - Location 9 i

Prj

1. Besomet - Tendon 1 I I'3 (g)1(2)I I I I*I I I*I 1 (D). (f) gallery OId3 I I*3 1(12) /

1 @I (a) 0 1(13 (g)

E 3 0 y>(f}

2. Structure - Operating ta) (b),(O I floor 1 1 1(31 0 0 0 1( 18a) 0 1(15) E30 w

3. Reactor equip. -

Reactor coolant pg spoto r 1 1 1( 8 )IU i 4. Reactor piping -

Steen gen. , , ig,)

5. Reactor egulp or $30 piping support -

Pressurizer support (83. (dl 1(g) 1 1 0(d) - 30 III. Outelde Containment Requirement - Location

& 1. Selselc Cat. 1 S Equip. - RMR pump c c 1(10)

D 2. Selselc Cat. 1 piping y Q,

e NSCW piping ,

c c 1(11)

W 3. Selselc Cet. 1 esluip.

O support or floor-Control bldg. sieb (el.(4) w 1(5) C C 0 O N

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VEGP-FSAR-6

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TABLE 6.3.2-1 (EHEET 1 OF 3)

EMERGENCY CORE COOLING SYSTEM 7,

COMPGNENT PARAMETERS Nm Accumulators Number 4 Design pressure (psig) 700 Design temperature ( F) 300

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N- Operating temperature (*F)60-120 Normal operating pressure (psig) 650 Total volume (ft') 1350 each Nominal water volume (ft') 9dO 4&O- each Nominal volume N 2 gas (ft') 400 each Boron concentration, nominal (ppm) 2000 l32 Charging pumps (See figure 6.3.2-4. )

Number 2 Design pressure (psig) 2800 Design temperature ( F) 300

(' Design flow (gal / min) 150 Design head (ft) 5800 Maximum flow (gal / min) 550 Design head at maximum flow (ft) 1400 Design head at shutoff (ft) 6200 Motor rating (hp) 600 Required NPSH at maximum flow (ft) (See figure 6.3.2-4.)

Available NPSH at maximum flow (ft) 81 from RWST SI pumps (See figure 6.3.2-5.)

Number 2 Design pressure (psig) 1750 Design tenperature (*F) 300 Design flow (gal / min) 425 Design head.(ft) 2680

.y- Maximum flow (gal / min) 650 Design head at maximum flow (ft) 1650 Design head at shutoff (ft) 3545 Motor rating (hp) 450 Required NPSH at maximum flow (ft) (See figure 6.3.2-4.)

(

Amend. 32 12/86

VEGP-FSAR-6

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TABLE 6.2.1-2 ASSUMPTIONS FOR CONTAINMENT ANALYSIS - PART 1 ir, 10 Service water temperature (*F) 95 Refueling water temperature (*F) X 120 Refueling water storage tank volume (gal) 715,000 Initial Containment.

Temperature (*F) 120 l

Initial pressure (psia) 15 Initial relative humidity (%) 20 Net free volume (fts) 2.75 x los i

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VEGP-FSAR-9 vertical position so that the assembly can be lifted out of the fuel container.

In the fuel handling building, fuel assemblies are moved about

]'.- by the fuel handling machine. When lifting fuel assemblies, the hoist uses a long-handled tool to ensure that sufficient radiation shielding is maintained. Initially, a shorter tool is used to handle new fuel assemblies, but the new fuel eleva-tor must be used to lower the assembly to a depth at which the fuel handling machine, using the long-handled tool, can place the new fuel assemblies into or out of the fuel storage racks.

[]

Decay heat, generated by the spent fuel assemblies in the fuel pools, is removed by the spent fuel pool cooling and cleanup system. After a sufficient decay period, the spent fuel assem-blies are removed from the fuel racks and loaded into a spent fuel shipping cask for removal from the site.

9.1.4.2.2 Refueling Procedure New fuel assemblies received for refueling are removed one at a time from the shipping container and moved into the new fuel assembly inspection area utilizing the monorail on the cask handling crane. After inspection, the accepted new fuel assem-r blies are stored in the new fuel storage racks. For the initial core load, some new fuel assemblies may be stored in the spent fuel pool.

The refueling operation follows a detailed procedure which provides safe and efficient refueling. Prior to initiating the refueling operation, the reactor coolant system (RCS) is borated and cooled down to refueling shutdown conditions as specified in the Technical Specifications. Criticality protection for refueling operations, including a requirement for 4 ili checks of boron concentration, s also specified in the Technical Specifications. The followin gnificant points are ensured by the refueling procedure: once per 7ghours,

(

l A. The refueling water and the reactor coolant contain approximately 2000 ppm boron. This concentration, together with the negative reactivity of control rods, is sufficient to keep the core approximately 5 percent C Ak/k subcritical during the refueling operations. It E} is also sufficient to maintain the core subcritical in the unlikely event that all of the rod cluster control assemblies (RCCAs) are removed from the core.

B. The water level in the refueling cavity is high enough to keep the radiation levels within acceptable limits 9.1.4-3

VEGP-FSAR-13 ,-

)

H. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.

I. Reports and meeting minutes of the PRB.

In addition, periodic audits of plant activities will be performed under the cognizance of the SRB.to evaluate:

A. The conformance of plant operations to provisions contained within the Technical Specifications and applicable license conditions (at least once per 12 l24

. months). E B. .The performance, training, and qualification of the entire plant staff (at least once per 12 months).

'C.- 'The results of actions taken to correct deficiencies occurring in plant equipment, structures, systems, or

~

method of operation that affect nuclear safety (at-least once per 6 months).

17 D. The performance of activities required by the Quality '

Assurance program (at least once per 24 months). 16 4, I E. Th'e Emergency Plan and implementing procedures (at least once per 12 months).

F. The Security Plan and implementing procedures (at least once per 12 months).

G. The Fire Protection Program and implementing procedures (at least once per 24 months).

H. Any other area of plant operation considered appropriate by the SRB .cnr the senior vice president-nuclear operations.

The SRB will report to and advice the senior vice president-nuclear operations on matters related to their responsibilities.

fe've.

The SRB will be composed of a minimum of 4MMeet persons who, as .

.a group, provide the expertise to review and audit.the operation of a nuclear power plant. The chairman and vice-chairman and other members shall be appointed by the~ senior l24 vice president-nuclear operations or other such person as he may designate. No more than a minority of the SRB will be l

members of the onsite operating organization. All alternates Amend. 16 4/85 Amend. 17 7/85 13.4.2-2 Amend. 24 6/86

VEGP-FSAR-16

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, TABLE'16.3-4 (SHEET 6 OF 15) l32 i

Valve 7alve Closure

( Number Function Time (s)

Check Valves (continued) 1206-U6-015 Containment spray supply N/A

! 2301-U4-036 Fire protection water N/A 2402-U4-017 Nitrogen supply to accumulator N/A 8

1208'-U4-021 Excess letdown and seal water leakoff N/A 120f-U6-032 Normal charging line N/A 120f-U4-355 Reactor coolant pump seal water supply N/A 6

1205-U4-354 Reactor coolant pump seal water supply N/A 120h-U4-353 Reactor coolant pump seal water supply N/A 120f-U4-004 Reactor coolar.t pump seal water supply N/A 1204-U6-128 RHR pump discharge to hot leg N/A 1204-U6-129 RHR pump discharge to hot leg N/A 1204-U6-147 RHR loop into cold leg N/A

, 1204-U6-148 RHR loop into cold leg N/A 1204-U6-149 RHR loop into cold leg N/A 1204-U6-150 RHR loop into cold leg N/A-1201-U6-112 Pressurizer relief tank makeup water

supply N/A 1411-U4-043 Chemical addition N/A 1411-U4-044 Chemical addition N/A 1513-U4-001 Containment H 2 m nitor discharge N/A 1513-U4-002 Containment H2 m nitor discharge N/A 4 k Amend. 30 12/26 Amend. 32 12/26

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VEGP-FSAR-16 A

TABLE 16.3-4 (SHEET 7 OF 15) l 32 Valve Valve Closure Number Function Time (s)

Check Valves (continued) 2401-U4-034 Service air and post-LOCA purge air

/- supply N/A

(_)T 2o 24 Dis-U4-049 Instrument air N/A 1302-U4-126 Auxiliary feedwater N/A 1302-U4-118 Auxiliary feedwater N/A 1302-U4-114 Auxiliary feedwater N/A 1302-U4-134 Auxiliary feedwater N/A 1302-U4-128 Auxiliary feedwater N/A

{ 1302-U4-120 1302-U4-115 Auxiliary feedwater Auxiliary feedwater N/A N/A 32 1302-U4-136 Auxiliary feedwater N/A 1302-U4-127 Auxiliary feedwater N/*.

1302-U4-119 Auxiliary feedwater N/A 1302-U4-116 Auxiliary feedwater N/A 1302-U4-135 Auxiliary feedwater N/A

() 1302-U4-125 Auxiliary feedwater N/A 1302-U4-117 Auxiliary feedwater N/A 1302-U4-113 Auxiliary feedwater N/A

( 1302-U4-133 Auxiliary feedwater N/A

5. Remote Manual HV-5280cb' Chemical Addition N/A -

( HV-5281<b' Chemical Addition N/A 3 Amend. 30 12/26 Amend. 32 12/26

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