ML20150B554

Forwards Photo of Fault on Property Adjacent to Proposed Site.Engineering Study Requested.W/O Encl

ML20150B554
Person / Time
Site:	05000524, 05000525
Issue date:	09/05/1978
From:	Keiler E KEILER & BUCKLEY
To:	NRC COMMISSION (OCM)
Shared Package
ML20150B553	List: ML20150B548 ML20150B554
References
NUDOCS 7811010389
Download: ML20150B554 (1)

Text

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griler & fBudirg 9 hffsTEfg5 It fEb 9047 Jefferson Xybway .

River Ridge, Bouisic.na 7o123 h cX GE/

737 7424 737-7423 9 y, cAu c/ ../u/7 September 5, 1978 United States Regulatory Commission Washington D.C. 20555 Re: Alabama Power Company (Alan R.

Barton Nuclear Plant, Units 1 & 21 Docket No. 50-524 and 50-525 Gentlemen:

Enclosed is a photograph of a fault which we uncovered on our property which is adjacent to the proposed site of the Alan R. Barton Nuclear Plant.

This East-Uest fault appears to a significant geological anomly and we suggest that it would be appropriate to undertake an Engineering Study to determine if it is active, or inactive, especially since the area is subject to earthquakes.

Yours very truly, KEILER & BUCKLEY EHK/pab w 'I4'I EDWARD E. KEILER 78110103Fr9

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.p. 4,'g NUCLEAR REGULATORY COMMISSION o y y WASHINGTON, D. C. 20555

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% . . . . . ,o! October 3, 1978 Docket Nos. 50-277 and 50-278 Philadelphia Electric Company ATTN: Mr. Edward G. Bauer, Jr. , Esquire V'ee President and General Counsel 2301 Market Street Philadelphia, Pennsylvania 19101 Gentlemen:

The Commission has issued the enclosed Amendments Nos. 46 and 46 to Facility Operating Licenses Nos. DPR-44 and DPR-56 for the Peach i Jottom Atomic Power Station Units Nos. 2 and 3. The amendments revise the Technical Specifications in response to your requests of May 6, 1977 and March 14, 1978, as supplemented July 11, 1978. I These amendments revise the Technical Specifications to (1) revise the l surveillance requirements associated with suppression pool temperature logging, (2) revise the Tables listing Jrimary Containment Isolation Valves (PCIV) to reflect the addition of a controlled bypass heatup line on the High Pressure Coolant Injection (HPCI) steam supply, and (3) revise the identification of certain valves to reflect plant unique designations between Units 2 and 3.

Copies of the related Safety Evaluation and Notice of Issuance are also enclosed. j Sincerely, a A Thomas

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.V M ppolito, Chief Operating Reactors Branch #3 Division of Operating Reactors )

Enclosures:

1. Amendment No. 46 to DPR-44

2. Amendment No. 46 to DPR-56

3. Safety Evaluation

4. Notice cc.w/ enclosures:

See page 2 N .I,b01~0 N

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Philadelphia Electric Company October 3,1978 CC:

Eugene J. Bradley Philadelphia Electric Company . Chief, Energy Systems Analysis Branch (AW-45F Office of Radiation Programs Assistant General Counsel V. S. Environmental Protection Agency 2301 Market Street Room 645, East Tower Philadelphia, Pennsylvania 19101 401 M Street, S. W.

Washington, D. C. 20460 Troy B. Conner, Jr.

.1747 Pennsylvania Avenue, N. W. V. S. Environmental Protection Agency l Region III Office Washington, D. C. 20006 ATTN: EIS COORDINATOR l

l Raymond L. Hovis, Esquire Curtis Building (Sixth Floor) 6th and Walnut Streets 35 South Duke Street Philadelphia, Pennsylvania ,

York, Pennsylvania 17401 19106 l M. J. Cooney, Superintendent Warren K. Rich, Esquire Generation Division - Nuclear Assistant Attorney General Philadelphia Electric Company Department of Natural Resources Annapolis, Maryland 21401 2301 Market Street Philadelphia, Pennsylvania 19101 Philadelphia Electric Company ATTN: Mr. W. T. Ullrich Government Publications Section State Library of Pennsylvania Peach Bottom Atomic Education Building l

Power Station  !

Delta, Pennsylvania 17314 Commonwealth and Walnut Streets Harrisburg, Pennsylvania 17126 Mr. R. A. Heiss, Coordinator Pennsylvania State Clearinghouse Governor's Office of State Planning and Development ~

P. O. Box 1323 l Harrisburg, Pennsylvania 17120 Albert R. Steel, Chairman Board of Supervisors i Peach Bottom Township R. D. #1 Delta, Pennsylvania 17314 I

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PHILADELPHIA ELECTRIC COMPANY PUBLIt- 3IAVICE ELECTRIC AND GAS COMPANY .

DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No. DPR-44

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The applications for amendment by Philadelphia Electric Company, et al, (the licensee) dated May 6,1977 and March 14, 1978, as supplemented July ll,1978, comply with the stand-ards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regula-tions set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application,

' the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assu*ance (i) that the ac'*vities authorized by this amencment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license i

amendment, and paragraph 2.C(2) of Facility Operating License l No. DPR-44 is hereby amended to read as follows:

(2) Technical Specifications f

! The Technical Specifications contained in Appendices 4

A and B, as revised through Amendment No. 46, are 4 hereby incorporated in the license. The ifcensee j shall operate the facility in accordance with the

! Technical Specifications.

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3. This license amendment is effective as of the date of its issuance.

a FOR THE, NUCLEAR REGULATORY COMMISSION j Thomas A. I polito, Chief Operating Reactors Branch #3 Division of Operating Reactors

Attachment:

2 Changes to the Technical Specifications

Date of Issuance

October 3, 1978 l

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ATTACHMENT TO LICENSE AMENDMENT NO. 46

' FACILITY OPERATING LICENSE N0. DPR-44 DOCKET NO. 50-277 Replace the following pages of the Appendix'"A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 165 165 181 181 185 185 186* 186*

187 187 190 190

• No change on this page

'LXMITING CONDIT10NS FOR OPERAT10N SURVEILLANCE REQUIREMENTS i 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS Applicability: Applicability:

Applies to the operating status Applies to the primary and secondary

of the primary and secondary containment integrity.

containment systems. l Objective: Objective: 1 1

To assure the integrity of the To verify the integrity of the primary and secondary containment primary and secondary containment, system.

Specification: Specification:

A. Primary Containment 1. The suppression chamber water l level and temperature shall be 4 1. At any time that the nuclear checked once per day. .

system is pressurized above atmospheric pressure or work 2. a. Whenever there is indication i is being done which has the of relief valve operation potential to drain the vessel, (except when the reactor is

the pressure suppression pool being shutdown and torus cool-water volume and temperature ing is being established) or shall be maintained within the testing which adds heat to the following limits except as suppression pool, the pool specified in 3.7.A.2. temperature shall be con-tinually monitored and also

a. Minimum water volume - observed and logged every 122,900 ft3 5 minutes until the heat

b. Maximum water volume - addition is terminated.

127,300 ft3

, 3. Whenever there is indication of

, relief valve operation with the 4

temperature of the suppression pool reaching 160'F or more and the primary coolant system pressure greater than 200 psig, an external visual examination of the suppression chamber shall be conducted before resuming power operation.

4. A visual inspection of the suppression chamber interior, including water line regions, shall be made at each major refueling outage.

Amendmer.t No. 46 -165-

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TABLE 3.'t.1 (Con t ' d. )

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PRIMARY CONTAINMENT ISOLATION VALVES

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$ Number of Power 4

Maximum Action on Operated Valves ' Operating Normal Initiating Inboard , Outboard Time (sec.) Position Signal Group -Valve Identification 2 S O GC 2D Drywell equipment drain dis-charge isolation valves 2 5 O' GC 2D Drywell floor drain discharge isolation valves 5 - NA C SC ,o 2D Traveling in-core probe to O GC D

, ( 4A IIPCI steam line draints 2 NA t/8 2 NA O - GC

.SA HCIC uteam line drains 2 NA C GC SA RCIC condensate pump drain 11PCI condensate pump drain 2 NA C SC 4A Toruu water filter tuimps 2 NA O GC 2D sucti on isolation valves 15 O GC 4B IIPCI Turbine Exhaust Vacuum 1 13reaker Isolation Valve 15 O GC SB 14CIC Turbine Extidust Vacuum 1 Breaker Isolation Valve 2 NA O GC

.t it PC I' uteam line exhaust drain IIPC I steam linie wierm-up* 1 NA C SC 4

• Effective upon completion of installation, approved by Amendment No.

4 PBAPS Unit 2 TABLE 3.7.4 j PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES Pen No. Notes l

7A AO-2-80 A; AO 86 A MSIV (1) (2) (3) (5) (9) a a i 7B AO-2-80B; AO-2-86B 1

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a 7C AO-2-80C; AO-2-86C i

AO-2-80D; AO- 2-86 D a "

j 7D )

i 8 Mo-2-74; MO-2-77 (1) (2) (4) (5) (9)  !

MO-23-19; Mo-23-2 0 ; MO-23-21 a 9A i

MO 38A; MO- 2663 a 1 9A '

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MO 21 ; MO 2 0; a 9B l

MO-13-30; MO-12-6 8;  ;

l MO-2-38B; Mo-2663 a f 9B {

i MO 15; Mo-13-16 a 10 a

i 11 MO 15; Mo 16 ; Ao-4807*

a 1 12 Mo 17 ; MO 18 i

i MO- 10-25B; Mo 154B; SV-4222 a 13A 133 MO 25A; Mo 15 4 A; SV-4221 i

a j , 14 MO 15; MO 18 1

a j 16A MO-14-123; Mo-14-11B; SV-4224 MO-14-12A; MO 11 A; SV- 4225 a 16B a

17 MO-10-32; Mo-10-33 a

18 AO 8 2 ; AO- 2 0- 83 a

19 AO-20-94; AO- 2 0- 95

• Effective upon completiori'of the modification authorized by Amendment No.

Amendment No. 46 '

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PBAPS Unit 2 TABLE 3.7.4 (Cont' d. )

PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES.

Pen No. Notes' 205A Ao-2502B; Check valve 9-26B (1) (2) (4) (5) (9) l "

205B AO-2502A; check Valve 9-26A

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211A MO-10-38B; MO-10-39B; Mo-10-34B (1) (2) (4) (5) (9) 211A SV-4951B; SV-4950B (1) (2) (4) (5) i 211B MO 38 A ; MO-10-3 9 A; Mo-10-34A (1) (2) (4) (5 ) (9) l 211B SV-4951A; SV-4950A (1) (2) (4) (5) 212 Check Valve 13-50; AO-4240; Ao-4241 (1) (2) (4) (5) (9) ,

214 Check Valve 23-65; Ao-4247; "2148 (1) (2) (4) (5 ) (9) 217B Md-4244; MO-4244A (1) (2) (4) (5) (9) 218A Ao-2968 (1) (2) (4) (5) (10) 218B SV-2671A; SV-2978A (1) (2) (4) (5) 219 AO-2511; Ao-2512; A0-2513; Ao-2514 (1) (2) (4) (5) (9) 219 SV- 2671F; SV-2978F; SV-4960A (1) (2) (4) (5)

SV-4961 A; SV-4 966 A 221 Check Valve 13-38 (1) (2) (4) (5) (9) 223 Check Valve 23-56 (1) (2) (4) (5) (9) ,. ,

225 MO 41 ; Mo 3 9 (1) (2) (4) (5 ) (9) 225 MO- 14-7 0 ; MO- 14-71 (1) (2) (4) (5) (9) 227 MO 58 ; Mo 5 7 (1) (2) (4) (5) (9) f l

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Amendment No; 46 1

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PBAPS Unit 2

. TABLE 3.7.4 (Con t' d. ) ,

I-PRIMARY - C0!Q7NMENT TESTABLE ISOIATION VALVES Pen No. Notes 22 AO-2969A; Check Valve (1) (2) (4) (5) (10) 25 Ao-2520; AO-2505; Ao-2519; (1) (2) (4) (5) (9)

AO-2521A; AO-2521B .

25 AO-2523; Check valves (1) (2) (4) (5) 26 AO-2506; AO-2507 (1) (2) (4) (5) (9) 26 SV- 2671G; SV-2978G (1) (2) (4) (5) 26 Ao-2509; AO-2510; AO-4235 (1) (2) (4) (5) (9) 26 SV-4960B; SV- 4961 B; SV-4966 B (1) (2) (4) (5) 39A MO-10-31B; MO 2 6B (1) (2) (4) (5) (9) 39A SV-4949B; SV-4948B (1) (2) (4) (5) 39B MO 31 A; MO 2 6 A (1) (2) (4) (5) (9) 39B SV-4949 A; SV- 4 94 8A (1) (2) (4) (5) 41 AO-2-39; AO-2-40 (1) (2) (4) (5) (9) 42 Check Valve 11-16, XV-14A, XV-14B (1) (2) (4) (5) (10) 51A SV- 2 67 tE; SV-2978E (1) (2) (4) (5) 513 SV- 2671D; SV- 2 978D 51C - SV-2671C; SV- 2978 C 51C SV-4960C; SV- 4 961 C ; SV-4966C "

51D SV-2980; Check valve 52F AO-2969B; Check Valve (1) (2) (4) (5) (10) ,

2 57 AO 316 ; AO- 2-317 (1) (2) (4) (5) (9) 203 SV-2671B; SV-2 978 B (1) (2) (4) (5)

203 SV-496OD;.'SVL 4961D; SV-4966D ,

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PBAPS 3.7.A & 4.7.A BASES (Cont'd).

The maximum allowable volume assures the integrity and functional capability of the Suppression Chamber (torus) during postulated LOCA pool swell effects on the torus support system. The majority of the Bodega tests were run with a sub-merged length of 4. feet and with complete condensation. Thus, with respect to downcomer submergence, this specification is adequate. The maximum tempera- .,

ture at the end of blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken to be the limit for complete con-densation of the reactor coolant, although condensation would occur for temperatures above 170 F.

Should it be necessary to drain the suppression chamber, this should only be done when there is no requirement for core standby cooling systems operability-as explained in basit 3.5.F.

Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160 F dur-ing any period of relief valve operation with sonic conditions at the discharge r exit. Specifications have been placed on the envelope of reactor operating conditions so' that the reactor can be repressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature' changes very slowly and monitoring these parameters daily is sufficient to establish any temperature' trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of testing which add significant heat, the temperature trends will be closely followed so that appropriate action can be taken if required.

Logging is not required during inadvertent relief valve operation since during such periods operator. action is actively and directly involved in operations  :

relating to controlling torus temperature and monitoring of temperature trends is a natural part of the operations. Additionally torus temperature is monitored by a recorder during these periods so that an historical record is available.

Operating procedures define the action to be taken in tne event a relief valve inadvertently opens or. sticks open. As a minimum this action shall include:

(1) use of all available means to close the valve, (2) initiate suppression

• pool water cooling heat exchangers, (3) initiate reactor shutdown, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.

The requirement for an external ' visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to.be the points of highest stress.

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Amendment No 46 -190-

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