ML19256A837

Application for Amend to Licenses DPR-44 & DPR-56 Re Shock Suppressors,Mod to High Pressure Svc Water Sys & Mod of Language Relating to Audits.Certificate of Svc Encl

ML19256A837
Person / Time
Site:	Peach Bottom
Issue date:	01/11/1979
From:	Baver E, Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19256A833	List: ML19256A832 ML19256A837
References
NUDOCS 7901160220
Download: ML19256A837 (15)

Text

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY 50-278 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 6 DPR-56 Edward G.

Bauer, Jr.

Eugene J.

Bradley 2301 Market Street Philadelphia, Pennsylvani.t 19101 Attorneys for Philadelphia Electric Company 7901160 B 0

BEFORE THE UNITED STATES NUCLEAR BEGULATORY COMMISSION In the Matter of Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY 50-278 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 S DPR-56 Philadelphia Electric Company, Licensee under Facility Operating Idcenses DPR-44 and DPR-56 for Peach Bottom Atomic Power Station Unit Nos. 2 and '3 respectively, hereby requests that the Technical Specifications incorporated in Appendix A of the Operating Licenses be amended by revising certain sections as indicated by a vertical bar in the margin of the attached pages 234k and 2341 for Unit No. 2; and pages 127 and 251 for both Unit No. 2 and Unit No.

3.

Pages 234m and 234n for Unit No. 2 are

included due to redistribution of material on the revised pages.

A discussion of each of the requested changes is set forth below.

Change 1 During the recent Unit No. 2 refueling outage (Sept. - Oct.,

1978) shock suppressors were added to the Residual Heat Removal (RHR) System and to the Reactor Water Clean-up (RWCU) System.

The shock suppressors added to the RHR system were part of the installation of ficw control valves in two of the RF3 loops.

These valves were installed for use during shutdown cooling.

The shock suppressor added to the RWCU system was installed to reduce piping stresses associated with the Design Basis Earthquake.

Therefore, pursuant to Technical Specification 3.11.D.5, which allows snubbere to be added to safety related systems without prior License Amendment provided that a revision to Table 3.11.D.1 is included with the next License Amendment request, it is requested that Table 3.11.D.1 of Unit No. 2 be revised to reflect the addition of the above referenced shock suppressors.

Also included in the changes to the snubber table are several corrections of typographical errors.

Change 11 The Peach Bottom High Pressure Service Water (HPSW) System is designed to maintain the service water pressure on the discharge of the Residual Heat Removal (RHR) heat exchanger higher than the pressure on the reactor coolant system side to preclude leakage of radioactive material into the cooling water.

To assure that the HPSW system meets this design criteria Technical Specification 4.5.B.1 (b) requires quarterly testing of the HPSW pumps to verify their capability to deliver 4500 gpm at a discharge pressure of 280 psig (existing Technical Specifications).

The required discharge pressure was established by an analysis, conducted by the Peach Bottom Architect Ennineer, of HPSW system operation in conjunction with all modes of RHR system operation.

Subsequent to this analysis, modifications have been completed which substantially reduce the pressure on the RHR side (reactor coolant side) of the RER heat exchanger during all operating modes.

This pressure reduction is the result of the installation of flow restricting orifices and drag valves upstream of the heat exchangers.

An analysis of RHR and HPSW pres sures (as was conducted for the existing Technical Specification HPSW pump requirements issued on February 12, 1975 as Amendment Nos. 5 and 3 to Facility Operating Licenses DPR-44 and DPR-56 respectively), considering all modes of operation, indicates that a HPSW pump discharge pressure of 233 psig is sufficient to preclude leakage of radioactive material into the cooling water.

This revised pressure limit includes the same margins for instrument accuracy and river water level variation included in the previous limit.

In.accordance with the above discussion, Section 4.5.B.

1 (b) should be changed to set the minimum pressure of the HPSW pump discharge at 233 psig.

Change III section 6.5.2.8 of the Technical Specification addresses audits of facility activities, however, the language of 6.5.2.8.a, 6.5.2.8.c. and 6.5.2.8.d. suggests complete review rather than audit.

The proposed change would revise the language by removing the word "all" in sections pertaining to audits, so that it is descriptive of an audit function rather than a complete review and would bring the language of the affected sections into conformance with that in the Standard Technical Specifications as published in NUREG 0123 Rev.

1, April 1, 1978.

Since none of the changes pertaining to Unit No. 2 involves a safety consideration and some are editorial or administrative in nature, pursuant to 10 CFR 170.22, Philadelphia Electric Company proposes that the Application for Amendment for Unit No. 2 be considered a Class II Amendment.

Since the changes for Unit No. 3 are a duplicate of changes requested for Unit No.

2, pursuant to 10 CFR 170.22,, Philadelphia Electric Company proposes that the Application for Amendment for Unit No. 3 be considered a Class I Amendment.

-4_

The Plant Operation Review Committee and the operation and Safety Review Committee have reviewed these proposed changes to the Technical Specifications and have concluded that they do not involve an unreviewed safety question, or a significant hazards consideration and will not endanger the health and safety of the public.

Respectfully submitted, PHILADELPHIA ELECTRIC COMPANY 3

m

,c Q

'^

Vi'ce Prssident'

• COMMONWEALTH OF PENNSYLVANIA ss.

COUNTY OF PHILADELPHIA S.

L. Daltroff, being first duly sworn, deposes and says:

That he is Vice President of Philadelphia Electric company, the Applicant herein; that he has read the foregoing Application for Amendment of Facility Operating Licenses and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

,,m l

//,LttCN~j~

subscribed and sworn to before me this N day

,\\f af C

[

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N a y Public ELIZA ETH H DOYER Nota.

,;talic, F%'a Ph:ta Co.

My Commizion Expires Jan. 30,1GE

CERTIFICATE OF SERVICE I certify that service of the foregoing Application was made upon the Board of Supervisors, Peach Bottom Township, York County, Pennsylvania, by mailing a copy thereof, via first-class mail, to Albert R. Steele, Chairman of the Board of Supervisors, R.

D.

No.1, Delta, Pennsylvania 17314 ; upon the Board of Supervisors, Fulton Township, Lancaster County, Pennsylvania, by mailing a copy thereof, via first-class mail, to George K.

Brinton, Chairman of the Board of Supervisors, Peach Buttom, Pennsylvania 17563; and upon *.he Board of Supervisors, Drumore Township, Lancaster County, Pennsylvania, by mailing a copy thereof, via first-class mail, to Wilmer P. Bolton, Chairman of the Board of Supervisors, R.

D.

No.

1, Holtwood, Pennsylvania 17532; all this 12th day of January,

1979.

i uged'e J.

Bradley

/

Atto ney for Philadelphia Electric Company

PBAPS Unit LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.A Core Spray _and LPCI 4.5.A Core _ Spray and LPCI Subsystem (cont'd)

S_u_b_ system (cont 'd )

6. All recirculation pump discharge

6. All recirculation pump discharge valves shall be operable prior to valves shall be tested for reactor startup (or closed if operabilty during any period permitted elsewhere in these of reactor cold shutdown exceeding specif ica tions).

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have not been performed during the preceeding 31 days.

7. If the requirements of 1.5.A cannot be met, an orderly shutdown of the reactor shall be initiated and the reactor shall be in the Ccid Shutdown Condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

B. Containment Coeling B. Containment Ccoli_ng Subsystem (HPSW)

Subsystem (HPSW)

1. Except as specified in 3.5.B.2,

1. Containment Cooling Subsystem 3.5.B.3, 3.5.B.4, and 3.5.F.3 Testing shall be as follows:

below, all containment cooling subsystem loops shall be operable item Frequency whenever irradiated f uel is in the reactor vessel and reactor coolant (a)

Pump Once/ month temperature is greater than 212 F, Operability and prior to reactor startup f rom a Cold Shutdown Condition.

(b)

Motor operated Once/nonth valve operability (c)

Pump Capacity After pump Test. Each IIPSW maintenance pump shall and every deliver 4500 3 months.

gpm at 233 psig.

(d)

Air test on once/5 years drywell and torus headers and nozzles.

2. From and after the date that any two

2. When it is determined that any two llPSW pumps are made or f ound to be llPSW pumps are inoperable, the inoperable for any reason, continued remaining components of the reactor operation is permissible only containment cooling subsystems shall during the succeeding thirty days, be demonstrated to be operable unless such pump is sooner made immediately and weekly thereafter.

operable, provided that during such thirty days all active components of the containment cooling subsystem are operable.

Amendment No. 23, 31, 47

-127-

M PflAPS Unit I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5. A Cor_e_Sgray and LPCI 4.5. A Core Sgray and, LPC,1 Subsy_stm (cont 'd)

S_ubsystem (cont'd)

6. All recirculation pump discharge

6. All recirculation pump discharge valves and bypass valve (s)[*] shall valves and bypass valve (s)[*] shall be operable prior to reactor startup be tested for operability during any (or closed if permitted elsewhere period of reactor cold shutdown in these specifications),

exceeding 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have not been perf ormed during the preceeding 31 days.

7. If the requirements of 3.5.A cannot be met, an orderly shutdown of tha reactor shall be initiated and the reactor shall be in the Cold Shutdown Condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

B. Containment Cooling B. Containment Cooling Subsystem (llPSW)

Subsy_ stem illPSW)

1. Except as specified in 3.5.8.2,

1. Containment Cooling Subsystem 3.5.B.3, 3.5.B.4, and 3.5.F.3 Testing shall be as to2 lows:

below, all containment cooling subsystem loops shall be operable item Fre3uency whenever irradiated fuel is in the reactor vessel and reactor coolant (a)

Punp once/ month temperature is greater than 212 F, Opera bility and prior to reactor startup f rom a Cold Shutdown Condition.

(b)

Hotor operated once/ month valve operability (c)

Punp Capacity After pump Test. Each IIPSW maintenance pump shall and every deliver 4500 3 months.

gpm at 233 psig.

l (d)

Air test on once/5 years drywell and torus headers and nozzles.

2. From and af ter the dat. that any two

2. When it is determined that any two llPSW pumps are made or f ound to be HPSW pumps are inoperable, the inoperable for any reasoa, continued remaining components of the reactor operation is permissible only containment cooling subsystems shall during the succeeding thirty days, be demonstrated to be operable unless such pump is sooner made immediately and weekly thereafter.

operable, provided th.at during such thirty days all active components of

• Upnn the removal of both recirculation the containment cooling subsystem are pump discharge valve bypass valves, operable.

operability and surveillance of only the recirculation pump discharge valves is req uired.

Anendment No. 27, 31, 47

-12/-

PBAPS Unit 2 TABLE 3.11.D.1 (Cont'd)

Safety _Related Shock Su g essor,s_(Snubb_er_s)

SNUBBERS SNUBBERS SNUhBERS SNUBBER IN HIGH(1)

ESPECIALLY INACCESSIBLE ACCESSIBLE SNUBBER RADIATION AREA DIFFICULT DURING NORMAL DURING NORMAL NUMBER LOCATION ELEVATION DU.".ING SHUTDOWN TO REMOVE OPERATION OPERATION 10-C B-S-12 RHR 98 See 4.ll.D.4.b

'C'RHR RM.

10-CB-S-43-1 RiiR 130 k

TORUS RM.

10-GB-S-43-2 RHR 130 X

TORUS RM.

~

10-GB-S-44 RHR 128 X

TURUS RM.

10-GB-S-48 RilR 124

'B'RHR RM.

N 10-GB-S-49 RHR 124 c.

7 10-CB-S-50 RiiR 98

~

10-C B-S-51 RiiR 98

'C'RHK RM.

10-GB-S-52 RHR 124 10-GB-S-53 RiiR 124 10-GB-S-54 RHR 130 X

TORUS RH.

10-CB-S-55 RHR 130 X

10-GB-S-58 RHR 98

'B'RHR RM.

10-DCN-S-73 RHR 180 X

Drywell 10-DCN-S-74 RilR 180 X

Drywell Ame ndment No. 33

i 9

?

PBAPS Unit 2 i

TABLE 3.ll D.1 (Cont'd) l Safety Related Shock Suppressors (Snubbers)

I SNUBBERS SNUBBERS SNUBBERS SNUBBER IN HIGH(1)

ESPECIALLY INACCESSIBLE ACCESSISLE SNUBBER RADIATION AREA DIFFICULT DURING NORMAL DURING NORMAL g

NUMBER LOCATION.

ELEVATION DURING SHUTDOWN TO REMOVE OPERATION OPERATION 10-CB-S-77 RHR 102 See 4.ll.D.4.b

'A'RHR RM.

10-GB-S-76 RHR 102

~

10-CB-S-75 RHR 93

~

10-GB-S-80 RHR 102

~

'D'RHR RM.

10-GB-S-79 RHR 102

~

10-CB-S-78 RHR 93

~

~

s Ny 12-DCN-S-2 RWCU 173.5 X

RWCU ISOLATION r-VALVE RM. 165 12-DCN-S-5 RWCU 165 X

Drywell 12-DCN-S-7 RWCU 165 X

Drywell 12-DCN-S-8A RWCU 162 Drywell l

~

14-DLN-S-23 CORE SPRAY 168 X

Drywell 14-DCN-S-24 CORE SPRAY 168 X

Drywell

~

14-DCN-S-26 CORE SPRAY 168 X

Drywell 14-DCN-S-27 CORE SPRAY 168 X

Drywell Amendment No. 33

PBAPS Unit 2 TABLE 3.ll.D.1 (Cont'd)

Safety Related Shock Suppressors (Snubbers)

SNUBBERS SNUBBERS SNUBBERS SNUBBER IN HICH(1)

ESPECIALLY INACCESSIBLE ACCESSIBLE SNUBBER RADIATION AREA DIFFICULT DURING NORMAL DURING NORMAL NUMBER LOCATION ELEVATION DURING SHUTDOWN TO REMOVE OPERATION OPERATION 13-HC-S-1 RCIC 107 See 4.ll.D.4.b X

RCIC ROOM 13-DDN-S-13 RCIC 96 13-lib-S-14 RCIC 102 X

~

13-DBN-S-15 RCIC 107 X

13-DBN-S-16 RCIC 140 X

TORUS ROOM 2 3-DBN-S-1 HPCI 112 X

llPCI ROOM

~

2 3-D BN-S-2 HPCI 112 X

~

23-DBN-S-3 HPCI 97 23-DBN-S-4 HPCI 97 23-DDN-S-9 HPCI 105 X

2 3-H B-S-16 HPCI 103

~

23-HB-S-19 HPCI 103 X

23-DBN-S-22 HPCI 155 X

Drywell 23-DBN-S-23 IIPCI 155 X

DrywsIl 23-DDN-S-25 IIPCI 105 X

llPCI HOO:1 23-DBN-S-27 HPCI 112 X

~

~

i Amendment No. 33

t k

i e

PBAPS Unit 2 TABLE 3 II.D.1 (Cont'd)

Safety Related Shock Suppressors (Snubbers)

SNUBBERS SNUBBERS SNUBBERS 3NUBB2R IN HIGH(1)

ESPECIALLY INACCESSIBLE ACCESSILLE SNUBBER RADIATION AREA DIFFIQJLT DURING NORMAL DURING NORMAL NUMBER LOCATION ELEVATION OURING SHUTDOWN TO REMOVE OPERATION OPERATION 23-DBN-S-28 HPCI I17 X

TORUS ROOM 23-DBN-S-29 HPCI 117 X

TORUS ROOM

~

2 3-IIB-S-30 HPCI 93 HPCI RO0tt

~

2 3-IIB-S-36 HPCI 103 23-HB-S-37 HPCI 103

~

23-HB-S-38 HPCI 126 X

TORUS ROOM E

s i

Notes for Table 3. II.D.1 (1)

Modifications to this Table due to changes in high radiation areas should be submitted to the NRC as part of the next license amendment.

Amendnent No. 33

PBAPS 6.5.2.7 Continued d.

Proposed changes in Technical Specifications or Licenses.

e.

Violations of applicable statutues, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance, f.

Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

g.

All events which are required by regulations or Technical Specifications to be reported to the NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

h.

Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

i.

Reports and meeting minutes of the Plant Operation Review Committee.

Audits 6.5.2.8 Audits of facility activitios shall be performed under the cognizance of the OSR Committee.

These audits shall encompass:

a.

The conformance of facility operation to provisions l

contained within the Technical Specifications and applicable license conditions at least once per year.

b.

The performance, training and qualifications of the entire facility staff at least once per year.

c.

The results of actions taken to correct l

6eficiencies occuring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per six months.

d.

The performance of activities required by the l

Quality Assurance Program to meet the criteria of 10 CFR 50, Appendix B, at least once per two years, r

Amendment No. 37

-251-p

.