ML18022A421

Resubmits Comment on Final Draft Tech Specs.Basis Section of Record 779 Expanded to Provide Addl Info Re Valve Testing & Cycling

ML18022A421
Person / Time
Site:	Harris
Issue date:	09/11/1986
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NLS-86-341, NUDOCS 8609220024
Download: ML18022A421 (152)

Text

fr (g Carolina Paver L Light Company SERIAL: NLS-86-30l SEP 11 Qgg P

Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO. I - DOCKET NO.50-000 COMMENT ON FINAL DRAFT TECHNICAL SPECIFICATIONS

REFERENCE:

Letter dated September 2, 1986 (NLS-86-326) from Mr. S. R. Zimmerman (CPkL) to Mr, Harold R. Denton (NRC).

Dear Mr. Denton:

Carolina Power 2 Light Company resubrnits a comment on the Final Draft Technical Specifications for the Shearon Harris Nuclear Power Plant. The basis section of Record.

~

No. 779 (attached) has been expanded to provide additional information to assist the staff

~

~

in their review of this comment. ~

If you have any questions, please contact Mr. Gregg A. Sinders at (9i9) 836-8l68.

Yours very truly, Orignal Signed By e g. 7lmtTle~n S, R. Zimmerman Manager Nuclear Licensing Section GAS/crs (008 l GAS)

Attachment cc: Mr. R. A. Benedict (NRC)

Mr. B. C. Buckley (NRC)

Mr. G. F. Maxwell (NRC-SHNPP)

Dr. 3. Nelson Grace (NRC-RII) k all t:ayettaville straat ~ p. o. 8ox 15st ~ RIteigh. N. c. artt02

rt g1 l

B"SAL FT ELECTRICAL POWER SYSTEHS ELECTRICAL E UIPHENT PROTECTIVE DEVICES MOTOR-OPERATED VALVES THERHAL OVERLOAD PRQTFCTION LIMITING CONDITION FOR OPERATION 3.8.4.2 The. thermal overload protection of each valve given in Table 3.8-2 shall be bypassed only under accident conditions by an OPERABLE bypass device integral with the motor starter.

APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE.

ACTION:

With the thermal overload protection for one or more of the above required valves not capable of being bypassed under conditions for which it is designed to be bypassed, restore the inoperable device or provide a means to bypass the thermal overload within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or declare the affected valve(s) inoperable and apply the appropriate ACTION Statement(s) of the affected system(s).

SURVEILLANCE RE UIREHENTS 4.8.4.2 The thermal overload protection for the above'required valves shall be verified to be bypassed only under accident conditions by an OPERABLE integral bypass device by the performance of a TRIP ACTUATION DEVICE OPERATIONAL TEST of the bypass circuitry:

]$'ow~~

a. At least once per 9KB~ for those thermal overloads which are normally in force during plant operation and are bypassed only under accident conditions; and

b. Following maintenance on the motor starter.

SHNPP RFvlglAbl AU8 586 SHEARON HARRIS - UNIT 1 3/4 8-39

CPBc L Coxnxnenta SHNPP Proof and Review Technical Specif ication s Record Number: 764 Comment Type: ERROR LCO Number: 3.08.04.02 Page Number: 3/4 8-40 & 42 Section Number: TABLE 3.8-2 Comment:

PAGE 3/4 8-40 VALVE NUMBER 1CS-472 CHANGE FUNCTION TO "RCP SEAL WATER RETURN ISOLATION."

PAGE 3/4 8-42 VALVES 1SW-97, 109$ 98 & 110 CHANGE THE FUNCTION TO "SW FROM FAN CLR Basis ALL OF THESE CHANGES ARE TO CORRECT TYPOGRAPHICAL, ERRORS IN THE ORIGINAL DATA SUPPLIED BY CP&L.

iilt L V,FT SHNPP TABLE 3.8-2 REVISlON MOTOR-OPERATED VALVE5 THERMAL OVERLOAD PROTECTION JUL BYPASS DEVICE VALVE NUMBER FUNCTION ~YES/NO 1CS-341 (2CS" V522) RCP A SEAL ISOL YES 1CS-382 (2CS-V523) RCP B SEAL ISOL YES 1CS-423 (2CS-V524) RCP C SEAL ISOL . YES 1CS-182 (2CS-V600) CSIP A MINIFLOM ISOLATION . YES 1CS-210 (2CS-V601) CSIP B MINIFLOW ISOLATION YES 1CS-196 (2CS-V602) CSIP C MINIFLOM ISOLATION YES 1CS-235 (2CS-V609) CSIP to RCS ISOLATION YES 1CS-166 (2CS- L521) VCT ISOLATION YES 1CS-292 (2CS"L522) RMST ISOLATION YES 1CS-214 (2CS-V585) C5IPS MINIFLOM ISOLATION YES 1CS-165 (2CS-L520) VCT ISOLATION YES 1CS-291 (2CS-L523} iNST ISOLATION YES 1CS-238 (2CS-V610) CSIP TO RCS ISOLATION YE5 1CS-170 (2CS-V587) CSIP SUCTION ISOLATION YES 1CS-169 (2CS-V589) CSIP SUCTION ISOLATION YES 1CS-171 (2CS-V590) CSIP SUCTION ISOLATION YES 1CS-168 (2CS-V588) CSIP SUCTION ISOLATION YE5 ~

1CS-219 (2CS-V603) CSIP DISCHARGE ISOL YES 1CS-217 (2CS-V604) CSIP DISCHARGE ISOL YES 1CS-218 (2CS-V605) CSIP DISCHARGE ISOL YE5 1CS-220 (2CS-V606) CSIP DISCHARGE ISOL YES 1CS-240 (2CS-V611) SEAL WATER INJECTION YE5 1CS-278 (2CS"V586) BORIC ACID TA KPO IP YES 1CS-746 (2CS-V757) CSIP MI YE5 1CS-752 (2t:5"V759) CSIP~ NIFLOM YES 1CS"753 (2CS-V760) C 0 MINIFLOM YES 1CS-745 (2CS-V758) SIP MINIFLOM YES 1CS"472 (2CS-V517) RCPj SEAL MATER R RN I50L YE5 1CS-4?0 (2CS-V516) RCP'EAL WATER ELATION YE5

'RH"25 (2RH"V507) HR TO CSIP ION YES 1RH-63 (2RH-V506) TO SUCTION YES 1RH-31 (2RH-F513) RHR A MINI FLOW YES 1RH-69 (2RH-F512) "4HR B MINI FLOW YE5 1RH-2 (1RH-V503) RHRS INLET ISOLATION YES 1RH-40 (1RH-V501} RHRS INLET ISOLATION YES 1RH-1 (1RH V502) RHRS INLET ISOLATION YES.

1RH-39 (1RH-VQÃ) RHRS INLET ISOLATION YES 15I-1 (25I-V503) BORON INJECTION TANK INLET ISOL YES lSI-4 (2SI-V506) BORON INJECTION TANK OUTLET ISOL YES lSI-2 (25I-V504) BORON INJECTION TANK INLET ISOL YES 1SI-3 {2SI-V505) BORON INJECTION TANK OUTLET ISOL YES 15I-246 (25I-V537) ACCUMULATOR A DISCHARGE ISOLATION YE5 15I-248 (2SI-V535) ACCUMULATOR C DISCHARGE ISOLATION YES 15I-300 (25I-V571) CNMT SUMP TO RHR PUMP A. ISOL- YE5 15I-310 (2SI"V573) CNMT SUMP TO RHR PUMP A ISOL YES 15I-247 (2SI-V536) ACCUM B DISCHARGE ISOLATION YES SHEARON HARRIS - UNIT 1 3/4 8-40

TABLE 3. 8-2 Continued SHXPP MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION REVISION JlJL 85.

BYPASS DEVICE VALVE NUMBER FUNCTION ~YES/NO 1MS-72 (2HS-V9) AFWTD STEAM C ISOLATION AO lSW-39 (3SW-85) NORMAL SW HDR A ISOLATION, YES 1SW-276 (3SW-88) NORMAL SW HDR A RETURN ISOL YES 1SW-270 (3SW-815) SW HDR A TO AUX RSVR ISOL YES 1SW-40 (3SW-86) NORMAL SW HDR 8 ISOL 'W YES 1SW-275 (3SW-813) HDR A RETURN ISOL YES 1SW-274 (3SW-814) SW HDR 8 RETURN ISOL YES 1SW" 271 (3SW-816) SW HDR 8 TO AUX RSVR ISOL YES 1SW-3 (3SW-83) EMER SW PUHP 1A MAIN RSVR INLET YES 1SW-4 (3SW-84) EHER SW PUMP 18 MAIN RSVR INLET YES 1SW-1 (3SW-81) MER SW PUMP~ AUX RSVR INLET YES 1SW-2 (3SW-82) EHER'SW PUMP 18+VX RSVR INLET YES 1SW-92 (2SW-846) SW TO FAN CLR AH3) INLET YES 1SW-97 (2SW-847 / S~W FAN CLR AH3(OUTLET YES 1SW-91 (2SW-845 SW TO FAN CLR AH2, INLET YES 1SW-109 (2SW-84 ) ~~ ~SW & FAN CLR AHg OUTLET YES 1SW-225 (2SW-85')

lSW-98 (2SW-848)

SW TO FAN CLR i INLET Vd FAN CLR A 1 OUTLET YES YES 1SW-227 (2SW-8 1) TO FAN CLR H4 INLET YES 1SW-110 (2SW"8 0) SW FAN CLR AH4 OUTLET YES 1SW-124 (3SW-8 0) SW TO AFWTD UHP YES 1SW-126 (3SW-8)l) SW TO AFWT PUMP YES 1SW-129 (3SW-87 ) SW TO AF PUMP YES 15W-127 (3SW-87 ) SW TO AF D PUMP YES 1SW-123 (3SW-875') SW TO PUMP A SUPPLY YES lSW-121 (3SW-874) SW AFW PUMP A SUPPLY YES 1SW-132 (3SW-877) W TO AFW PUMP 8 SUPPLY YES 1SW-130 (3SW-876) SW TO AFW PUMP 8 SUPPLY YES 1ED-94 ( 2MD-V36) CNHT SUMP ISOLATION YES 1ED-95 ( 2MD-V77) CNHT SUMP ISOLATION YES 3CZ-85 RAB ELEC PROT INLET YES 3CZ-86 RAB ELEC PROT INLET YES 3CZ-87 RAB ELEC PROT EXHAUST YES 3CZ-88 RAB ELEC PROT EXHAUST YES 3CZ-832 RAB ELEC PROT PURGE NAKE-UP YES 3CZ-833 , RAB ELEC PROT PURGE NAKE-UP YES 3CZ-834 RAB ELEC PROT PURGE INLET YES.

3CZ-835 RAB ELEC PROT PURGE INLET YES 3FV-82 FUEL HANDLING EXHAUST INLET NO*

3FV-84 FUEL HANDLING EXHAUST INLET NO*

3CZ-81 CONTROL ROOM NORMAL SUPPLY ISOL NO>>

3CZ-.83 CONTROL ROOM NORMAL EXHAUST ISOL NO*

3CZ-817 CONTROL ROOH PURGE MAKE UP NO*

3CZ-82 CONTROL ROOM NORMAL SUPPLY ISOL NO" 3CZ"84 CONTROL ROOM EXHAUST ISOLATION NO" 3CZ-818 CONTROL ROOM PURGE MAKE UP NO" 3CZ-814 CONTROL ROOM PURGE EXHAUST NO*

f SHEARON HARRIS - UNIT 1 3/4 8-42

jR CP8cL Comments SHNPP Final Dr af t Technical .Specif ication.

R >cur d t<!> rn~l.'e>r 7!.ra Connr~r ~ r'> 'L 'I vf'> i' Ef<<ROI'aue I.CQ flu>rrLrr>r r > ..>8. () 4 ~ 62 NurnLrer: 3/4 8-4'r Sect) <,>! !  !'Iurr>!.,>r. r" 'ABLE =.8-2 Co nrnenl:

!r(!EII: CULL!t'1N "I<<YF"ASS DEVICE" t<ND PU'I A THf;-'.UNC; I;IOtlr'(I !)ESI Rlf"TICIM PQR VAI VES ON Pr'rGL.

+'f TER r',

I l'>I!"'"."), ] r>F - '?,:.. 1Af.-77 ~ ! AF'-137. 1AF-14:,

1AF -14rr'. AND 1tIS-70 i: PAGE 8-4'- ( 1!1S-72. f- V-B'..

r V-B4, 3C7-81, 3C? -B3 r 3CZ-B17, <C7-B", C -B'I . ~ ~

3CZ-816. AND 'CZ-B14): AND PAGE 8-43 <3CZ-B2~

3CZ-.B25. 31.'Z-BI <<, 3CZ-Bl:, 3CZ-B10. 3CZ-B9

<<CZ-811. <<C7-B23. 3C7.-B21, 3CZ-B22, 3<<CZ-L<=4 r 3CZ-B1 ~> . At ID 3CZ-82Cr )

REVISE 1'HE + FOOTNCITE ON PAGE 8-43 TO READ "Included 4:or completeness only and are riot; tested urrder- thi s speci f i cat i on. Ove. I oad bypass i s ac!on>pl i sl'ed in ci rcvi t desi on bv ttre 1 ocati orr oi acti vat i on rel ays. '. hese acti vat i on sl ave ." el ays ar r. tested i rr accor-dance rri th the reouiremer'!ts of

>-'ll'rr fat: le

( 3!q k<as 1

".'H I S C,'HANGE RE'V I SES THIS REC<<UEST IN ACCORDANCE WITH DI SCU !SIGNS ON 8-14-86 WI 1'H f'1R. Q. CHOPRA Of THE NRR STAFF. AT THAT TIt1E IT WAS STATED THAT WHILE IT WAS ACCEPTABLE THAT THESE SPECIAL ITEMS ARE 1'0 I~E T ESl'ED ELSEWHERE ~ HE FEL T THA'! THF BYI"ASS DEVICE CQLUMII SHOULD STILL READ "YES".

SlNCL" 1H!S IJC)LILD INCAN '1HA1 ALL ITEMS IN THE COLUMN WLr\.!I..D BE I E)Ef>I C I CAL CP>1(L FEEI S THAT THE'QLUMf'I ( AN Bl-" I"Cr Mf"LET!. LY DELETED. '7 HE FOOTNOTE HAS }3FEN REVISED TO BE SOI'IEWHAT CLF, << ~ ) MORE SPECIFIC ABC)UI (JFI!:f"E THE OTHER . T REC>!UI RE .-.NTS flAY BE I! OUt!D.

TABLE 3.8-2 MOTOR-OPERATED VALVES THERMA'L OVERLOAD PROTECTION VALVE NUMBER FUNCTION 1CS-341 (2CS-V522) RCP A SEAL ISOL 1CS-382 (2CS-V523) RCP B SEAL ISOL 1CS-423 (2CS-V524) .

RCP C SEAL ISOL 1CS-182 (2CS"V600) CSIP A MINIFLOW ISOLATION 1CS-210 (2CS"V601} CSIP B MINIFLOW ISOLATION 1CS-196 (ZCS-V602) CSIP C MINIFLOW ISOLATION 1CS-235 (2CS-V609) CSIP to RCS ISOLATION 1CS-166 (2CS" L521) VCT - I SOLAT ION 1CS-292 (2CS-L522) RWST ISOLATION 1CS-214 (2CS-V585) CSIPS MINIFLOW ISOLATION 1CS-165 (2CS-L520) VCT ISOLATION 1CS-291 (2CS-L523) RWST ISOLATION 1CS-238 (2CS-V610) CSIP TO RCS ISOLATION 1CS-170 (2CS-V587) CSIP SUCTION ISOLATION 1CS-169 (2CS-V589) CSIP SUCTION ISOLATION 1CS-171 (2CS-V590) CSIP SUCTION ISOLATION 1CS-168 (2CS-V588) CSIP SUCTION ISOLATION 1CS-219 (2CS-V603) CSIP DISCHARGE ISOL 1CS"217 (2CS-V604) CSIP DISCHARGE ISOL 1CS"218 (2CS-V605) CSIP DISCHARGE ISOL 1CS-220 (2CS-V606) CSIP DISCHARGE ISOL 1CS-240 (2CS-V611) SEAL WATER INJECTION 1CS-278 (2CS-V586) BORIC ACID TANK TO CSIP 1CS-746 (2CS-V757) CSIP MINIFLOW 1CS-752 (2CS-V759) CSIP MINIFLOW 1CS"753 (2CS-V760) CSIP MINIFLOW 1CS-745 (2CS-V758) CSIP MINIFLOW 1CS-472 (2CS-V517) RCP( SEAL WATER RETURN ISOL 1CS-470 (2CS-V516} RCP SEAL WATER ISOLATION 1RH"25 (2RH-V507) RHR TO CSIP SUCTION 1RH-63 (2RH-V506) RHR TO CSIP SUCTION 1RH" 31 (2RH" F513) RHR A MINI FLOW lRH-69 (2RH-F512) RHR B MINI FLOW 1RH-2 (1RH-VS03) RHRS INLET ISOLATION 1RH" 40 (1RH-V501) RHRS INLET ISOLATION 1RH-1 (1RH-V502) RHRS INLET ISOLATION 1RH-39 (1RH-V500) RHRS INLET ISOLATION 1SI-1 (25 I-V503) BORON INJECTION TANK INLET ISOL 1SI-4 (2SI-V506) BORON INJECTION TANK OUTLET ISOL 151-2 (2SI-V504) BORON INJECTION TANK INLET ISOL 15 I-3 (2SI" V505) BORON INJECTION TANK OUTLET ISOL 1SI-246 (2SI-V537) ACCUMULATOR A DISCHARGE ISOLATION 15 I-248 (2SI" V535) ACCUMULATOR C DISCHARGE ISOLATION 15 I-300 (25 I-V571) CNMT SUMP TO RHR PUMP A ISOL 1SI-310 (2SI-V573) CNMT SUMP TO RHR PUMP A ISOL 1SI" 247 (2SI-V536) ACCUM B DISCHARGE ISOLATION SHEARON HARRIS - UNIT 1 3/4 8-40

F F3A D FT TABLE 3. 8-2 Continued MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION

-94%86~%

VALVE NUMBER- FUNCTION ~~@=

151-301 (251-V570) CNMT SUMP TO RHR PUMP 8 ISOL 151-311 (251-V572) CNMT SUMP TO RHR PUMP 8 ISOL 151-107 (251-V500) HH SI TO RCS HL 151-52 (251-V502) HH SI TO RCS CL 151-86 (251-V501) HH SI TO RCS HL 151 "326 (251-V577) LH SI TO RCS HL 151-327 (251-V576)- LH 51 TO RCS HL 151"340 (251-V579) LH 51 TO RCS CL 151"341 (251-V578) LH SI TO RCS CL 151-359 (251"V587) LH 51 TO RCS HL 151-322 (251-V575) RWST TO RHR A ISOL 151-323 (251-V574) RWST TO RHR 8 ISOL 1CC-128 (3CC-85) CCS NONESSENTIAL RETURN ISOL 1CC-127 (3CC-86) CCS NONESSENTIAL RETURN ISOL 1CC-99 (3CC-819) CCS NONESSENTIAL RETURN ISOL 1CC"113 (3CC-820) CCS NONESSENTIAL RETURN ISOL 1CC-147 (3CC"V165) RHR COOLING ISOL 1CC-167 (3CC-V167) RHR COOLING ISOL 1CC-176 (2CC-V172) CVCS HX CNMT ISOLATION lCC-202 (2CC-V182) CVCS HX CNMT ISOLATION 1CC-208 (2CC-V170) CCW-RCPS ISOLATION 1CC-299 (2CC-V183) RCPS BEARING HX ISOLATION 1CC-251 (2CC-V190) RCPS THER BARRIER ISOLATION 1CC-207 (2CC-V169) CCW-RCPS ISOLATION 1CC-297 (2CC-V184) RCPS BEARING HX ISOLATION lCC-249 (2CC-V191) RCPS THER BARRIER I50LATION 1CT-105 (2CT-V6) CNMT SPRAY SUMP A RECIRC ISOL 1CT-102 (2CT-V7) CNMT SPRAY SUMP 8 RECIRC ISOL 1CT-26 (2CT"V2) CNMT SPRAY PUMP A INJECT. SUPPLY 1CT-71 (2CT" V3) CNMT SPRAY PUMP 8 INJECT. SUPPLY 1CT-50 (2CT-V21) SPRAY HDR A ISOLATION 1CT-12 (3CT-V85) NAOH ADDITIVE ISOLATION ICT"88 (2CT-V43) SPRAY HDR 8 ISOLATION ICT-11 (3CT" V88) NAOH ADDITIVE ISOLATION 1CT-47 (2CT-V25) CNMT 5PRAY HDR A RECIRC 1CT-24 (2CT-V8) CNMT SPRAY PUMP A EDUCTOR TEST 1CT-95 (2CT"V49) CNMT SPRAY HDR 8 RECIRC 1CT-25 (2CT-V345) CNMT SPRAY PUMP 8 EDUCTOR TEST 1AF" 5 (3AF-V187) AFWP A RECIRC 1AF-24 (3AF-V188) AFWP B RECIRC 1AF-55 (2AF-V10) AFW TO SG A ISOL ~

lAF-93 (2AF"V19) AFW TO SG 8 ISOL W 1AF-74 (2AF"V23) AFW TO SG C ISOL 1AF-137 (2AF-V116) TO SG A ISOL >

+'FWTD 1AF-143 (2AF-V117) AFWTD TO SG 8 ISOL 5 lAF"149 (2AF-V118) AFWTD TO SG C ISOL +

1MS-70 (2MS-V8) AFWTD STEAM 8 ISOLATION <

5HEARON HARRIS - UNIT 1 3'-41

TABLE 3.8-2 Continued MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICE VALVE NUMBER FUNCTION ~YES/KO 1MS-72 (2MS-V9) AFWTD STEAM C ISOLATION W 1SW-39 (3SW-85) NORMAL SW HDR A ISOLATION 1SW-276 (3SW-88) NORMAL SW HDR A RETURN ISOL 1SW-270 (3SW-815) SW HDR A TO AUX RSVR ISOL lSW-40 (3SW" 86) NORMAL SW HDR 8 ISOL 1SW-275 (35W-813) SW HDR A RETURN ISOL 1SW-274 (3SW-814) SW HDR 8 RETURN ISOL 1SW-271 (3SW-816) SW HDR 8 TO AUX RSVR ISOL MbL~S~ .MBt&

lSW-92 (2SW-846) SW TO FAN CLR AH3 INLET 1SW-97 (2SW-847) SW % FAN CLR AH3 OUTLET lSW-91 (2SW-845) W TO FAN CLR AH2 INLET 1SW-109 (2SW-849) SW FAN CLR AH2 OUTLET lSW-225 (2SW-852) W TO FAN CLR AH1 INLET 1SW-98 (2SW-848) SW FAN CLR AH1 OUTLET 1SW-227 (2SW-851) TO FAN CLR AH4 INLET lSW" 110 (2SW-850) SW FAN CLR AH4 OUTLET 1SW-124 (3SW-870) SW TO AFWTD PUMP 1SW-126 (3SW-871) SW TO AFWTD PUMP 1SW-129 (3SW-873) SW TO AFWTD PUMP 1SW-127 (3SW-872) SW TO AFWTD PUMP 1SW-123 (3SW-875) SW TO AFW PUMP A SUPPLY 1SW-121 (3SW-874) SW TO AFW PUMP A SUPPLY 1SW-132 (3SW-877) SW TO AFW PUMP 8 SUPPLY 1SW-130 (3SW-876) SW TO AFW PUMP 8 SUPPLY 1ED-94 (2MD"V36) CNMT SUMP ISOLATION lED-95 (2MD-V77) CNMT SUMP ISOLATION 3CZ-85 RAB ELEC PROT INLET 3CZ-86 RAB ELEC PROT INLET 3CZ-87 RAB ELEC PROT EXHAUST 3CZ"88 RAB ELEC PROT EXHAUST 3CZ-832 RAB ELEC PROT PURGE MAKE"UP 3CZ-833 RAB ELEC PROT PURGE MAKE-UP 3CZ-834 RAB ELEC PROT PURGE INLET 3CZ-835 RAB ELEC PROT PURGE INLET 3FV-82 FUEL HANDLING EXHAUST INLET >

3FV-84 FUEL, HANDLING EXHAUST INLET 3CZ-81 CONTROL ROOM NORMAL SUPPLY ISOLINE 3CZ-83 CONTROL ROOM NORMAL EXHAUST ISOLA 3CZ-817 CONTROL ROOM PURGE MAKE UP+

3CZ-82 CONTROL ROOM NORMAL SUPPLY ISOL 3CZ-84 ROOM EXHAUST ISOLATION+ 8'ONTROL 3CZ-818 CONTROL ROOM PURGE MAKE UP~

3CZ-814 CONTROL ROON PURGE EXHAUST+

SHEARON HARRIS - UNIT 1 3/4 8-42

3.8-2 Continued hi) "L FT TABLE VALVES THERMAL OVERLOAD PROTECTION

'OTOR-OPERATED BYPASS DEVICE VALVE NUMBER FUNCTION ~YES/NO 3CZ-826 CONTROL ROOM NORMAL SUPPLY DISCH +

3CZ-825 CONTROL ROOM SUPPLY DISCHARGE ~

3CZ-813 CONTROL ROOM PURGE EXHAUST ~ ~

3CZ-812 CNTL RM EMER FLTR OUTSIDE AIR INTAKE 3CZ"810 CNTL RM EMER FLTR OUTSIDE AIR INTAKE +

3CZ"89 CNTL RM EMER FLTR OUTSIDE AIR INTAKE W 3CZ"811 CNTL RM EMER FLTR OUTSIDE AIR INTAKE 3CZ-823 ROOM EMER FLTR INLET P" f'ONTROL 3CZ-821 CONTROL ROOM FLTR DISCHARGE +

3CZ-822 CONTROL ROOM EMER FLTR DISCHARGE+

'3CZ-824 CONTROL ROOM EMER FLTR INLET+

3CZ-819 CONTROL ROOM EMER FLTR DISCHARGE W 3CZ-820 CONTROL ROOM EMER FLTR DISCHARGE 8 3AV-81 RAB EMER EXHAUST INLET 3AV-82 RAB EMER EXHAUST OUTLET 3AV-84 RAB EMER EXHAUST INLET 3AV-85 RAB EMER EXHAUST OUTLET 3AV-83 RAB EMER EXHAUST BLEED 3AV-86 RAB EMER EXHAUST BLEED 3AC-82 RAB SMGR .8 EXHAUST 3AC-83 RAB SWGR 8 EXHAUST 3AC-81 RAB SMGR A EXHAUST

+< J p~ ~z'f +eye J uudcs +4~> +pic.iOycA~'~'~

"Included for completeness only Overload bypass is accomplished 4y-circuit Slave Relay> ra a(.. 'les>>ed /u Rcccada~m mc9Ii +~>c Ac~ v i'ac~eu+s ef Qg(g p. 3 "~ ~

SHEARON HARRIS - UNIT 1 3/4 8-43

CP8cL Comment.a BHNPP Final Draft. Technical Specifications Rec or u Nu:rrber-: Cue!r!ent Type: ERROR 70'CQ rluebe.: i.08.04.02 Page Number" ~/4 8-40 Se'i un N rrrrirr: r: TABLE i. 8-2 Comment:

DELETE'OI Ut1tl "BYPASS DEVICE" AND PUT A ~J AFTER THE FUNCTIONAL DESCRIPTION FOR VALVES Of~! PAGE 8-41

' AF-c."..r'. 1A"-'~ . 1AF-74 . 1AF-1 7. 1AF-14 ~ 1 AF, 14c: !AND RB-7<> PAbr= 8-42 (1twB 72 ;cV B2

F V-B4.:<<CZ-B1 ..~CZ-B ~. >CZ-B17, 3CZ-B2 ~CZ-B4.

<<CZ-B18. Ah?D 3C:-814): AhlD PAGE 8-43 <3CZ-B2*

CZ-B25. LZ-81 , >CZ-B12. <<CZ-B10r CZ-BW,

~<<CZ 'B1 crotch 1 NCZ B23 Z<<CZ B21 ~>CZ B22 p NCZ B24

- CZ-B1., At.!D =CZ-B20>

1 p p REVISE THE:: FQOThlQTE ON PAGE 8-4 TQ READ Over i oaci bvpass ~or these val ves i s accoepl I s?red bV ?le aC lvatlOI1 T?'le:e M laye 1'elaVS if1 CXt Cui'tc vat i on ) ave el avs ar e teshec3 as pat'"'t I

o~ the Fngineere 1 Baf ety Features Act.'uati on System 1%tf u>>lerrgatw otl w n accorclance 5!x th the r eoui re!:re!1i:s o>> Tabl e 4. c-2.

Bask s TI.I cB CHANBES REVICFB THIS REDUEST IN ACCORDANCE i?ITH DISCUSSIQNS Oti 8-14-86 At'ID 8-28-86 llITH t'I!i.

Q C. IOPRA OF THE NRR BTA. F AT THAT T I f'1 I T i'JAS 8 ATED THAl LJHILE IT NAS ACCEPTABLE THAT THESE 1

SPECIA'TE?1S APE TO BE TEBTED ELSELJHEREr HE FEl.T THAT THE BYPASS DEV'I CL. CQLUt'?hl SHQLJLD BT ILL READ "YEB". SINCE H'? B I'JQU!LE) f'? AN THAT ALL IT h1B IN 1

1

."'IE CQLUHN liJOULD BE I DEh?T ICAL. CP~(L FEELS THAT THE Cr ILU!'!!J CAf! BE CQI'IPLETELY DE'TED. THE FOOTNQTF rdr-18  ? .."EN REVI'="ED TQ BE AtlD h?ORE SPECIFIC ABOUT

'HER""." TFIt Q? ."'IER TEST REQL! I REh?EflTB t'IAY BE FQUf'JD.

SHNPP Ik pm]et~a~

TABLE 3.8"2 AUG 586 MOTOR-OPERATEO VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTION 1CS-341 (2CS-V522) RCP A SEAL ISOL 1CS-382 (2CS-V523) RCP B SEAL ISOL lCS-423 (2CS-V524) RCP C SEAL ISOL 1CS-182 (2CS-V600) CSIP A MINIFLOM ISOLATION 1C5-210 (2CS-V601) CSIP 8 MINIFLOM ISOLATION 1C5-196 (2C5-V602} CSIP C MINIFLOM ISOLATION 1CS-235 (2CS-V609) CSIP to RCS ISOLATION 1CS-166 (2CS-L521) VCT ISOLATION 1CS-292 (2CS-L522) RWST ISOLATION 1CS"214 (2CS-V585) CSIPS MINIFLOW ISOLATION 1CS-165 (2CS-L520) VCT ISOLATION 1CS-291 (2CS-L523) RWST ISOLATION 1CS-238 (2C5-Y610) CSIP TO RCS ISOLATION 1CS-170 (2CS-V587) CSIP SUCTION ISOLATION 1CS-169 (2CS-V589) CSIP SUCTION ISOLATION 1CS-171 (2CS-V590) CSIP SUCTION ISOLATION 1CS-168 (2CS-V588) CSIP SUCTION ISOLATION 1CS-219 (2CS-V603) CSIP DISCHARGE ISOL 1CS-217 (2CS-V604) CSIP DISCHARGE ISOL 1CS-218 (2CS-V605) CSIP DISCHARGE ISOL 1CS-220 (2CS-V606) CSIP DISCHARGE ISOL 1CS" 240 (2CS-V611) SEAL MATER INJECTION 1CS-278 (2CS-V586) BORIC ACID TANK TO CSIP 1CS-746 (2CS-V757) CSIP MINIFLOW 1CS-752 (2CS-V759) CSIP MINIFLOW 1CS-753 (2CS-V760) CSIP MINIFLOM 1CS"745 (2CS-V758) CSIP MINIFLOW 1CS-472 (2CS-V517) RCPT SEAL MATER RETURN 150L 1CS-4?0 (2CS"Y516) RCP SEAL MATER ISOLATION 1RH-25 (2RH-Y507) RHR TO CSIP SUCTION 1RH-63 (2RH-V506} RHR TO CSIP SUCTION 1RH-31 (ZRH-F513) RHR A MINI FLOW 1RH-69 (2RH-F512) RHR 8 MINI FLOW 1RH-2 (1RH-V503) RHRS INLET ISOLATION 1RH"40 (1RH-V501) RHRS INLET ISOLATION 1RH-1 (1RH-V502) RHRS INLET ISOLATION 1RH-39 (1RH- Y500) RHRS INLET ISOLATION lSI"1 (2SI- Y503} BORON INJECTION TANK INLET ISOL 1SI-4 (25I-V506} BORON INJECTION TANK OUTLET ISOL 15I-2 (25I-V504) BORON INJECTION TANK INLET ISOL 15I-3 (25I-V505) BORON INJECTION TANK OUTLET ISOL 15I-246 (25I" V537) ACCUMULATOR A DISCHARGE ISOLATION 15I-248 (25I-V535) ACCUMULATOR C DISCHARGE ISOLATION 1SI"300 (25I-V571) CNMT SUMP TO RHR PUMP A ISOL 1SI" 310 (25I-V573) CNMT SUMP TO RHR PUMP A ISOL lSI-247 (2SI-V536) ACCUM B DISCHARGE ISOLATION SHEARON HARRIS - UNIT 1 3/4 8-40

SHNPP FML tT p~itptA4< TABLE 3. 8-2 Continued AUG $ 86 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTION lSI-301 (2SI-V570) CNMT SUMP TO RHR PUMP B ISOL 1SI-311 (2SI-V572) CNMT SUMP TO RHR PUMP B ISOL 1SI-107 (2SI-V500) HH SI TO RCS HL 1SI-52 (2SI-V502) HH SI TO RCS CL 1SI-86 (2SI-V501) HH SI TO RCS HL 1SI-326 (2SI"V577) LH SI TO RCS HL 1SI-327 (2SI-V576) LH SI TO RCS HL 15 I-340 (2SI-V579) LH SI TO RCS CL 1SI-341 (2SI-V578) LH SI TO RCS CL lSI-359 (2SI-V587) LH SI TO RCS HL 15I"322 (2SI-V575) RWST TO RHR A ISOL 1SI-323 (2SI"V574) RWST TO RHR B ISOL 1CC-128 (3CC-85) CCS NONESSENTIAL RETURN ISOL 1CC-127 (3CC-86) CCS NONESSENTIAL RETURN ISOL 1CC-99 (3CC"819) CCS NONESSENTIAL RETURN ISOL 1CC-113 (3CC-B20) CCS NONESSENTIAL RETURN ISOL 1CC-147 (3CC- Y165) RHR COOLING ISOL 1CC-167 (3CC-V167) RHR COOLING ISOL 1CC-D6 (2CC-V172) CVCS HX CNMT ISOLATION 1CC-202 (2CC-Y182) CVCS HX CNMT ISOLATION 1CC-208 (2CC-V170) CCW-RCPS ISOLATION 1CC-299 (2CC-V183) RCPS BEARING HX ISOLATION 1CC-251 (2CC-V190) RCPS THER BARRIER ISOLATION 1CC-207 (2CC-V169) CCW-RCPS ISOLATION 1CC-297 (2CC-Y184) RCPS BEARING HX ISOLATION 1CC-249 (2CC-V191) RCPS, THER BARRIER ISOLATION 1CT" 105 (2CT-Y6) CNMT SPRAY SUMP A RECIRC ISOL 1CT-102 (2CT-V?) CNMT SPRAY SUMP B RECIRC ISOL 1CT-26 (2CT-V2) CNMT SPRAY PUMP A INJECT. SUPPLY 1CT-71 (2CT-V3) CNMT SPRAY PUMP B INJECT. SUPPLY 1CT-50 (2CT-V21) SPRAY HDR A ISOLATION 1CT-12 (3CT-V85) NAOH ADDITIVE ISOLATION ICT-88 (2CT-V43) SPRAY HDR B ISOLATION ICT-11 (3CT-V88) NAOH ADDITIVE ISOLATION 1CT-47 (2CT-V25) CNMT SPRAY HDR A RECIRC 1CT-24 (2CT-V8) CNMT SPRAY PUMP A EDUCTOR TEST 1CT-95 (2CT-Y49) CNMT SPRAY HDR B RECIRC 1CT-25 (2CT-Vl45) CNMT SPRAY PUMP B EDUCTOR TEST lAF-5 (3AF"V187) AFWP A RECIRC 1AF-24 (3AF-Y188) AFWP B RECIRC 1AF-55 (2AF-Vlo) AFW TO SG A ISOL W 1AF-93 (2AF-V19) AFW TO SG B ISOL +

1AF-74 (2AF-Y23) AFW TO SG C ISOL +

1AF-137 (2AF-V116) AFWTD TO SG A ISOL ~

1AF-143 (2AF-Y117) AFWTD TO SG B ISOL W 1AF-149 (2AF- Y118) AFWTD TO SG C ISOL +

1MS-70 (2MS-V8) AFWTD STEAM 8 ISOLATION W SHEARON HARRIS - UNIT 1 3/4 8-41

SHNP P REVIS3ON AU6 NS TABLE 3.8-2 Continued MOTOR"OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTION 1MS-72 (2MS-V9) AFWTD STEAM C ISOLATION 8 1SW-39 (3SW-85) NORMAL SW HDR A ISOLATION 1SW-276 (3SW-88) NORMAL SW HDR A RETURN ISOL 1SW-270 (3SW-815) SW HDR A TO AUX RSVR ISOL 1SW-40 (3SW-86) NORMAL SW HDR 8 ISOL lSW-275 (3SW-813) SW HDR A RETURN ISOL 1SW-274 (3SW-814) SW HDR 8 RETURN ISOL 1SW-271 (3SW-816) SW HDR 8 TO AUX RSVR ISOL 1SW-92 (2SW"846) SW TO FAN CLR AH3 INLET 1SW-97 (2SW-847) SW % FAN CLR AH3 OUTLET lSW-91 (2SW-845) W TO FAN CLR AH2 INLET 1SW-109 (2SW-849) SW FAN CLR AH2 OUTLET 1SW-225 (2SW-852) W TO FAN CLR AHl INLET 1SW-98 (2SW-848) SW FAN CLR AH1 OUTLET 1SW-227 (2SW-851) TO FAN CLR AH4 INLET 1SW-110 (2SW-850) SW FAN CLR AH4 OUTLET 1SW-124 (3SW-870) SW TO AFWTD PUMP 1SW-126 (3SW-871) SW TO AFWTD PUMP 1SW-129 (3SW-873) SW TO AFWTD PUMP 1SW-127 (3SW-872) SW TO AFWTD PUMP 1SW-123 (3SW-875) SW TO AFW PUMP A SUPPLY 1SW-121 (3SW-874) SW TO AFW PUMP A SUPPLY 1SW-132 (3SW-877) SW TO AFW PUMP 8 SUPPLY 1SW-130 (3SW-876) SW TO AFW PUMP 8 SUPPLY 1ED-94 (2MD-V36) CNMT SUMP ISOLATION 1ED"95 (2MD-V77) CNMT SUMP ISOLATION 3CZ"85 RAB ELEC PROT INLET 3CZ-86 RAB ELEC PROT INLET 3CZ"87 RAB ELEC PROT EXHAUST 3CZ-88 RAB ELEC PROT EXHAUST 3CZ-832 RAB ELEC PROT PURGE MAKE"UP 3CZ-833 RAB ELEC PROT PURGE MAKE-UP 3CZ-834 RAB ELEC PROT PURGE INLET 3CZ-835 RAB ELEC PROT PURGE INLET 3FV-82 FUEL HANDLING EXHAUST INLET K 3FV-84 FUEL HANDLING EXHAUST INLET W 3CZ-81 CONTROL ROOM NORMAL SUPPLY ISOLA 3CZ-83 CONTROL ROON NORMAL EXHAUST ISOLA 3CZ-817 CONTROL ROOM PURGE MAKE UP+

3CZ-82 .CONTROL ROOM NORMAL SUPPLY ISOL k 3CZ-84 CONTROL ROOM EXHAUST ISOLATION%

3CZ-818 CONTROL ROOM PURGE MAKE UP%

3CZ-814 CONTROL ROON PURGE EXHAUST%

SHEARON HARRIS - UNIT 1 3/4 8-42

SHNPP gm! Ic.lw~~

$ 86 TABLE 3.8-2 Continued

'fi)ALRIF AUG MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTION 3CZ"826 CONTROL ROOM NORMAL SUPPLY DISCH +

3CZ-825 CONTROL ROOM SUPPLY DISCHARGE +

3CZ-813 CONTROL ROOM PURGE EXHAUST+

3CZ-812 CNTL RM EMER FLTR OUTSIDE AIR INTAKE'NTL 3CZ-810 RM EMER FLTR OUTSIDE AIR INTAKE'NTL 3CZ-89 RM EMER FLTR OUTSIDE AIR INTAKEP 3CZ-Bll CNTL -RM EMER FLTR OUTSIDE AIR 3CZ-823 EMER FLTR INLET+

INTAKE'ONTROL ROOM 3CZ-821 CONTROL ROOM FLTR DISCHARGE+

3CZ-822 CONTROL ROOM EMER FLTR DISCHARGE 6 3CZ-824 CONTROL ROOM EMER FLTR INLET ~

3CZ" 819 CONTROL ROOM EMER FLTR DISCHARGEw 3CZ-820 CONTROL ROOM EMER FLTR DISCHARGE W 3AV-81 RAB EMER EXHAUST INLET 3AV-82 RAB EMER EXHAUST OUTLET 3AV-84 RAB EMER EXHAUST INLET 3AV-85 RAB EMER EXHAUST OUTLET 3AV-83 RAB EMER FXHAUST BLEED 3AV-86 RAB EMER EXHAUST BLEED 3AC-82 RAB SWGR 8 EXHAUST 3AC-83 RAB SWGR 8 EXHAUST

'3AC-Bl RAB SWGR A EXHAUST Chloe $

Overload bypass is accomplished by s}Rve Relays <w +e GiRca t to Mes@ Qc~lo'/~<<

~ Ache>>ki; 5(+<< 'RelAys ARe +es+cd hs p ~R+ 4 +bc ~~gIIJeeRcd 5agc+ pe>fu mes

$ $ +4m ~~$ 'Aumcsf f+i toA) gnl ~~gd>>,~ec ~ +t >> +/ e ge ~ ge~ +$

SHEARON HARRIS - UNIT 1 3/4 8-43 o4 7mbte V.3-2

0/j C P8cm C camrnmn<m SHNPP Final Dra+t. Technical Speci+icat ion.

Record". Number: 780 Comment: Tyae: ERROR LCO Number: '3.08.04.D2 Paoe Number: 3/4 8-42 Section t4umber: TABLE 3.8-2 Comment:

DELETE VALVES 1SW-1 . 18M-2, 1Slrj-w AND 1St'-4 FRQtl THF TABLE.

&eel 5 Dl.)E TQ A PLANT tqQDIF I CAT I ON. Tl-IFSF VALVES HAVE BEEN CHANGED TO htANUAL VALVES AND THEREFORE THERE I S t~lQ THERMAL OVERLOAD BYPASS. REHOTE OPERATION OF THFSE VALVES WAS NQT ASSUtCED BY ANY SAFElY ArtALYSIS.

gVi

SHNPP REVtStON HIS Pili:1 AU6 NNi TABLE 3.8-2 Continued MOTOR"OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTION.

1MS-72 (2MS-V9) AFWTD STEAM C ISOLATION 5 1SW-39 (3SW-85) NORMAL SW HDR A ISOLATION 1SW-276 (3SW-88} NORMAL SW HDR A RETURN ISOL 1SW-270 (3SW-815) SW HDR A TO AUX RSVR ISOL 1SW"40 (3SW-B6} NORMAL SW HDR 8 ISOL 1SW-275 (3SW-813) SW HDR A RETURN ISOL 1SW-274 (3SW-814) SW HDR 8 RETURN ISOL 1SW-271 (3SW-816) SW HDR 8 TO AUX R5VR ISOL lSW-92 (2SW-846) SW TO FAN CLR AH3 INLET lSW"97 (2SW-847) SW & FAN CLR AH3 OUTLET 1SW-91 (2SW-845) TO FAN CLR AH2 INLET lSW-109 (2SW-849) SW FAN CLR AH2 OUTLET 15W-225 (2SW-.852) TO FAN CLR AHl INLET 1SW-98 (2SW-B48) SW FAN CLR AHl OUTLET 1SW-227 (2SW-851) TO FAN CLR AH4 INLET 1SW-110 (2SW-850) SW FAN'LR AH4 OUTLET 15W-124 (3SW-870) SW TQ AFWTD PUMP 1SW-126 (3SW-871) SW TO AFWTD PUMP 15W-129 (3SW-873} SW TO AFWTD PUMP ISW-127 (3SW-872) SW TO AFWTD PUMP 1SW-123 (3SW-875) SW TO AFW PUMP A SUPPLY 1SW-121 (3SW-874) SW TO AFW PUMP A SUPPLY 1SW-132 (3SW-877) SW TO AFW PUMP 8 SUPPLY 1SW-130 (3SW-876) SW TO AFW PUMP 8 SUPPLY lED-94 (2MD-V36) CNMT SUMP ISOLATION lED-95 (2MD-V77) CNMT SUMP ISOLATION 3CZ-85 RAB ELEC PROT INLET 3CZ-86 RAB ELEC -PROT INLET 3CZ-87 RAB ELEC PROT EXHAUST 3CZ-88 RAB ELEC PROT EXHAUST 3CZ-832 RAB ELEC PROT PURGE MAKE"UP 3CZ-833 RAB ELEC PROT PURGE MAKE-UP 3CZ-834 RAB ELEC PROT PURGE INLET

'CZ-835 RAB ELEC PROT PURGE INLET 3FV-82 FUEL HANDLING EXHAUST INLET K 3FV"84 FUEL HANDLING EXHAUST INLET W 3CZ-81 CONTROL ROOM NORMAL SUPPLY ISOI +

3CZ-83 CONTROL ROOM NORMAL EXHAUST ISOLA 3CZ-817 'ONTROL ROOM PURGE MAKE UP +

3CZ-82 CONTROL ROOM NORMA SUPPLY ISOL 5 3CZ-84 CONTROL ROOM EXHAUST ISOLATION%

3CZ-818 CONTROL ROOM PURGE MAKE 3CZ-814 ROOM PURGE EXHAUST&

UP+'ONTROL SHEARON HARRIS - UNIT 1 3/4 8"42

>/~

CP RL Comxnenta

~PP Proof and Review Technical Specification8 f

Record Number: 706 Comment Type: ERROR LCO Number: 3.08.04.02 Page Number: 3/4 8-41,42 Section Number: TABLE 3.8-2 Comment:

THE LAST SEVEN ITEMS ON PAGE 3/4 8-41 AND THE FIRST ITEM ON PAGE 3/4 8-42 CHANGE THE BYPASS DEVICE COLUMN FROM "YES" TO HN04" Basis THIS CHANGE IS REQUIRED DUE TO RECENT PLANT MODIFICATIONS. THE RESULT OF THESE MODIFICATIONS IS THAT THE THERMAL OVERLOAD BYPASS FUNCTION IS NOW COVERED BY INHERENT FEATURES DESIGNED INTO THE CIRCUITRY AND THERE IS NO LONGER A BYPASS DEVICE" TO BE TESTED.

qg(u

hN TABLE 3. 8-2 Continued SHXPP p&llsigM MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION JUL 586 BYPASS DEVICE VALVE NUMBER FUNCTION ~YES/NO 1SI-301 (2S I-V570) CNMT SUMP TO RHR PUMP 8 ISOL YES 1SI-311 (25 I-V572) CNMT SUMP TO RHR PUMP 8 ISOL YES 1SI-107 (ZSI-V500) HH SI TO RCS HL YES 1SI-52 (2SI-V502) HH SI TO RCS CL YES 15 I-86 (2SI-V501)

'H SI TO RCS HL YES 1SI-326 (2SI-V577) LH SI TO RCS HL YES 15 I-327 (2SI-V576) LH SI TO RCS HL YES 1SI-340 (2SI-V579) LH SI TO RCS CL YES 1S I-341 (25 I-V578) LH SI TO RCS CL YES 1SI-359 (2SI "V587) LH SI TO RCS HL YES 1SI-322 (2SI-V575) RWST TO RHR A ISOL YES 1SI-323 (2SI-V574) RWST TO RHR 8 ISOL YES 1CC-128 (3CC-85) CCS NONESSENTIAL RETURN ISOL YES 1CC-127 (3CC-86) CCS NONESSENTIAL RETURN ISOL YES 1CC-99 (3CC-819) CCS NONESSENTIAL RETURN ISOL YES 1CC-113 (3CC-820) CCS NONESSENTIAL RETURN ISOL YES 1CC-147 (3CC-V165) RHR COOLING ISOL YES 1CC-167 (3CC-V167) RHR COOLING ISOL YES.

1CC-176 (2CC-V172) CVCS HX CNMT ISOLATION YES 1CC-202 (2CC-V182) CVCS HX CKMT ISOLATION YES 1CC-208 (2CC-V170) CCW-RCPS ISOLATION YES 1CC-299 (2CC-V183) RCPS BEARING HX ISOLATION YES 1CC-251 (2CC-V190) RCPS THER BARRIER ISOLATION YES 1CC-207 (2CC-V169) CCW-RCPS ISOLATION YES 1CC-297 (2CC-V184) RCPS BEARING HX ISOLATION YES 1CC-249 (2CC-V191) RCPS THER BARRIER ISOLATION YES 1CT-105 (2CT" V6) CNMT SPRAY SUMP A RECIRC ISOL YES 1CT-102 (2CT-V7) CNMT SPRAY SUMP 8 RECIRC ISOL YES 1CT-26 (2CT"V2) CNMT SPRAY PUMP A INJECT. SUPPLY YES 1CT-71 (2CT-V3) CNMT SPRAY PUMP 8 INJECT. SUPPLY YES 1CT-50 (2CT-V21) SPRAY HDR A ISOLATION YES 1CT-12 (3CT-V85) NAOH ADDITIVE ISOLATION YES ICT-88 (2CT-V43) SPRAY HDR 8 ISOLATION YES ICT-ll (3CT-V88) NAOH ADDITIVE ISOLATION YES 1CT-47 (2CT-V25) CNMT SPRAY HDR A. RECIRC YES 1CT-24 (2CT-V8) CNMT SPRAY PUMP A EDUCTOR TEST YES 1CT-95 (2CT-V49) CNMT SPRAY HDR 8 RECIRC YES-1CT-25 (2CT-V145) CNMT SPRAY PUMP 8 EDUCTOR TEST YES 1AF" 5 (3AF-V187) AFWP A RECIRC YES

]AF-24 (3AF-V188) AFWP 8 RECIRC YES 1AF-55 (2AF-V10) AFW TO SG A ISOL lAF"93 (2AF-V19) AFW TO SG 8 ISOL 1AF-74 (2AF-V23) AFW TO SG C ISOL 4Q& /VO+

1AF-137 (2AF-V116) AFWTD TO SG A ISOL 44& A/O+

lAF-143 (2AF-V117) AFWTD TO SG 8 ISOL lAF"149 (2AF-V118) 1MS-70 (2MS-Vs)

AFWTD TO SG. C ISOL AFWTD STEAM 8 ISOLATION ~

4QF h/0+

uD~

SHEARON'HARRIS - UNIT 1 3/4 8"41

rN TABLE 3.8-2 Continued MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION oN JUL 6.

BYPASS DEVICE VALVE NUMBER FUNCTION ~YES/NO 1MS-72 (2MS-V9) AFWTD STEAM C ISOLATION hO*

1SW-39 (3SW-85) NORMAL SW HDR A ISOLATION YES 1SW-276 (3SW-88) NORMAL SW HDR A RETURN ISOL YES 1SW-270 (3SW-815)

. SW HDR A TO AUX RSVR ISOL YES 1SW"40 (3SW-86) NORMAL SW HDR 8 ISOL YES 1SW-275 (3SW-813) SW HDR A RETURN ISOL YES 1SW-274 (3SW-814) SW HDR 8 RETURN ISOL YES 1SW-271 (3SW-816) SW HDR 8 TO AUX RSVR ISOL YES lSW-3 (3SW-83) EMER SW PUMP 1A MAIN RSVR INLET YES lSW-4 (3SW-B4) EMER SW PUMP 18 MAIN RSVR INLET YES 1SW"1 (3SW-Bl) EMER SW PUMP lA AUX RSVR INLET YES 1SW-2 (3SW-82) EMER SW PUMP 18 AUX RSVR INLET YES 1SW-92 (2SW-846) SW TO FAN CLR AH3 INLET YES 1SW-97 (2SW-847) S~& FAN CLR AH3 OUTLET YES lSW-91 (2SW-845) SW TO FAN CLR AH2 INLET YES 1SW-109 (2SW-849) ~~ ~SW & FAN CLR AH2 OUTLET YES 1SW-225 (2SW-852) SN TO FAN CLR AN1 INLET YES lSW-98 (2SW"848)  % FAN CLR AHl OUTLET YES 1SW-227 (2SW-851) TO FAN CLR AH4 INLET YES 1SW-110 (2SW-850) SW FAN CLR AH4 OUTLET YES 1SW-124 (3SW-870) SW TO AFWTD PUMP YES 1SW-126 (3SW-871) SW TO AFWTD PUMP YES 1SW-129 (3SW-873) SW TO AFWTD PUMP YES 1SW-127 (3SW-B72) SW TO AFWTD PUMP YES 1SW-123 (3SW-875) SW TO AFW PUMP A SUPPLY YES 1SW-121 (3SW-874) SW TO AFW PUMP A SUPPLY YES 1SW-132 (3SW-877) SW TO AFW PUMP 8 SUPPLY YES 1SW-130 (3SW-876) SW TO AFW PUMP 8 SUPPLY YES 1ED-94 (2MD-V36) CNMT SUMP ISOLATION YES 1ED-95 (2MD-V77) CNMT SUMP ISOLATION YES 3CZ-85 RAB ELEC PROT INLET YES 3CZ-86 RAB ELEC PROT INLET YES 3CZ-87 RAB ELEC PROT EXHAUST YES 3CZ-88 RAB ELEC PROT EXHAUST YES 3CZ-832 RAB ELEC PROT PURGE MAKE-UP YES 3CZ-833 RAB ELEC PROT PURGE MAKE-UP YES 3CZ-834 RAB ELEC PROT PURGE INLET YES.

3CZ-835 RAB ELEC PROT PURGE INLET YES 3FV-82 FUEL HANDLING EXHAUST INLET NO" 3FV-84 FUEL HANDLING EXHAUST INLET NO~

3CZ-81 CONTROL RIM NORMAL SUPPLY ISOL NO*

3CZ-83 CONTROL ROOM NORMAL EXHAUST ISOL NO" 3CZ-817 CONTROL ROOM PURGE MAKE UP NO*

3CZ-82 CONTROL ROOM NORMAL SUPPLY ISOL NO*

3CZ-84 CONTROL ROOM EXHAUST ISOLATION NO" 3CZ-818 CONTROL ROOM PURGE MAKE UP NOsN 3CZ-814 CONTROL ROOM PURGE EXHAUST NO" SHEARON HARRIS - UNIT 1 3/4 8-42

CPBc.L Coxnxnenta HNPP Proof and Rev iew Technical 8 pecif ication 8 Record Number: 734 Comment Type: IMPROVEMENT LCO Number: 3.09.01 Page Number: 3/4 9-2 Section Number:, TABLE 4.9-1 Comment:

REVISE TABLE PER THE ATTACHED MARKUP.

Basis THESE CHANGES ARE PROPOSED FOR CONSISTENCY WITHIN THE TABLE AND TO PROVIDE ADDITIONAL INFORMATION USEFUL TO PLANT PERSONNEL.

TABLE 4.9-1 FI Fl SHNPP ADMINISTRATIVE CONTROLS OW/)Qt&hl TO PREY N LU N U N UELING VALVE POSITION JUL Ie6 VALVE t88~N/ID DURING REFUELING LOCK DESCRIPTION 1CS-149 Closed Yes RN to the CVCS makeup control (cs -bi+'se) system 1CS-510 Closed Yes Boric Acid Batch Tank Outlet Ccs->~a/av) valve.july be opened if the batching tank concentration is > 2000 ppm boron, and valve 1CS-503 (makeup water supply to batch tank) is closed.

1CS-503 Closed Yes Rl% to Batching Tank. Do not (cs-z zs.i) open unless outlet valve 1CS-510 is closed.

~CYL5 uraouA n BTRs.

1CS-570 Closed No Q Place valve in "shut/ at valve

(~-~s-~s.s~) control switch and p'Lance BTRS function selector swia.'h in "off." No lock required.

1CS-670 Closed Yes RN to BTRS loop.

(cs->s99 $ 4) 1CS-649 Closed Yes Resin sluice to BTRS (cs-7198  %) demineralizers.

1CS-93 Closed Yes Resin sluice to CVCS (cs -Ds I srV3 demineralizers 1CS-320 Closed Yes Recycle Evaporation Feed (cs-Doe su) Pump to charging/safety injection pump suction, 1CS"98 Open No BTRS bypass valve. Place g~->we s.) valve control switch in "open" position; e'b g(

SHEARON MARRIS - UNIT 1 3/4 9-2

CP8cL Coxnxnents HNP P Proof and Review Technical S deci%'ication s Record Number: 777 Comment Type: IMPROVEMENT LCO Number: 3.09.06 Page Number: 3/4 9-7 Section Number:, 4.9.6.1 Comment:

CHANGE "when the refueling machine load exceeds" TO "at less than or equal to".

Basis THIS CHANGE IS NECESSARY TO ENSURE THAT THE LOAD CUTOFF IS SET AT OR -BELOW 2700 lbs., NOT WHEN THE LOAD EXCEEDS 2700 lbs.

FN REFUELING OPERATIONS SHNPP 3/4.9.6 REFUELING MACHINE OPERABILITY RFv) p)A~j L

JUL $ 86 LIMITING CONDITION FOR OPERATION 3.9.6 The refueling machine and auxiliary hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:

a. The refueling machine, used for movement of fuel assemblies, having:

1. A minimum capacity of 4000 pounds, and

2. An automatic overload cutoff limit less than or equal to 2700 pounds.

b. The auxiliary hoist, used for latching and unlatching drive rods, having:

1. A minimum capacity of 3000 pounds, and

2. A 1000-pound load indicator that shall be used to monitor loads to prevent lifting more than 600 pounds.

APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel.

ACTION:

With the requirements for the refueling machine and/or auxiliary hoist OPERA-BILITY not satisfied, suspend use of any inoperable refueling machine and/or auxiliary hoist from operations involving the movement of drive rods and fuel assemblies within the reactor vessel.

~

SURVEILLANCE RE UIREMENTS 4.9.6.1 The refueling machine used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE, within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations, by performing a load test of at least 4000 pounds and demonstrating an automatic load cutoff 27 t N LZSJ 7HAr4 dR ~CIA<

4.9.6.2 The auxiliary hoist and associated load indicator used for'ovement of drive rode within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least 900 pounds.

SHEARON HARRIS - UNIT 1 3/4 9-7

0 CP RL Cornxnenta

~

iNPP Proof and Review Tech.nical Specifications t

Record Number: 703 Comment Type: ERROR LCO Number: 1.09.12 Page Number: 3!4 9-14, 15,16 g'> < g-3 Section Number: VARIOUS Comment:

ITEMS 4 . 9. 12.

b . 1, 4 . 9. 12. d. 5, '4 . 9. 12. e, 4 . 9. 12 . f AND BASES CHANGE ANSI N510-1975 TO ANSI N510-1980.

Basis THIS CHANGE IS NECESSARY FOR CONSISTENCY WITH THE FSAR.

REFUELING OPERATIONS FINA Ft'HNPP 3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST REV)S)ON JUL N6 LIMITING CONDITION FOR OPERATION 3.9.12 Two 'independent Fuel Handling Building Emergency Exhaust System Trains shall be OPERABLE.

APPLICABILITY: Whenever irradiated fuel is in a storage pool.

ACTION:

a~ With one Fuel Handling Building Emergency Exhaust System Train inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the OPERABLE Fuel Handling Building Emergency Exhaust System Train is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorber.

b. With no Fuel Handling Building Emergency Exhaust System suspend all operations involving movement of fuel %thin Traf~'PERABLE, the storage pool or crane operation with loads over the storage pool until at least one Fuel Handling Building Emergency Exhaust System Train is restored to OPERABLE status.

C. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.12 The above required Fuel Handling Building Emergency Exhaust System trains shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following significant painting, fire, qr chemical release in any ventilation zoril comunicating with the system by:

l. Verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05K and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the unit flow rate t

is 6600 cfm 10K during system operation when tested in accordance with ANS.I N510-~.

680 SHEARON HARRIS - UNIT 1 3/4 9-14

REFUELING OPERATIONS SHNPP FUEL HANDLING BUILDING EHERGENCY EXHAUST @+I)pr l i JUL $ 86 SURVEILLANCE RE UIREHENTS Continued

4. 9. 12 (Continued)

2. Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, Harch 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, Harch 1978, by showing a methyl iodide ~enetration of less than 1.(C when tested at a temperature of 30 C and at a relative humidity of 70K in accordance with AS'3803.

C. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C. 6.b of Regulatory Guide 1.52, Revision 2, March %78, meets the laboratory testing criteria of Regulatory Position+.6.a of Regulatory Guide 1.52, Revision 2, }hrch 1978, by showing a methyl iodide penetration of less than 1.0X when tested at a temperature of .

30'C and at a relative humidity of 7'n accordance with'ASTM 03803.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA fil-ters and charcoal adsorber bank is not greater than 4.1 inches water gauge while operating the unit at a flow rate of 6600 cfm a lOX,

2. Verifying that, on a High Radiation test signal, the system automatically starts and directs its exhaust flow through the HEPA filters and charcoal adsorber banks,

3. Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/8 inch water gauge, relative to the outside atmosphere, during system operation at a flow rate of 6600 cfm l: 10K,

4. Verifying that the filter cooling bypass valve is locked in the balanced position, and f

5. Verifying that the heaters dissipate 40 a 4 N when tested in accordance with ANSI N510-%%5:

zoo

e. After each complete or partial replacement of a HEPA'filter bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-i8$ 8 for a OOP test aerosol while operating the unit at a flow rate a+6600 cfm f 10K.

~~8o SHEARON HARRIS - UNIT 1 3/4 9-15

FINAL RAFT REFUELING OPERATIONS SHNPP FUEL HANDLING BUILDING EMERGENCY EXHAUST RFWen~

J0l. $ 86 SURVEILLANCE RE UIREHENTS Continued 4.9;12 (Continued)

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510- for a halogenated hydrocarbon refrigerant test gas while operating he unit at a flow rate of 6600 cfe t 10K.

198>

SHEARON HARRIS " UNIT 1 3/4 9-16

REFUELING OPERATIONS

";;., FN LORAFT 586 BASES 3/4.9. 10 AND 3/4.9. 11 WATER LEVEL - REACTOR VESSEL AND NEW AND SPENT FUEL LS The restrictions on minimum water level ensure that sufficient water depth is available to rhmove 99K of the assumed 10K iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consis-tent with the assumptions of the safety analysis'/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST 'SYSTEM The limitations on the Fuel Handling Building Emergency Exhaust System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptjons of the safet anal ses. ANSI N510- will be used as a procedural gui5e-. for i 980 surveillance testing. riteria for laboratory testing of charcoal anchor in-place testing of HEPA filters and charcoal adsorbers is based upon removal efficiencies of 95K for organic and elemental forms of radioiodine and 99K for.

particulate forms. The filter pressure drop was chosen to be half-way between the estimated clean and dirty pressure drops for these components. This assures the full functionality of the filters for a prolonged period, even at the Technical Specification limit.

SHEARON HARRIS - UNIT 1 B 3/4 9"3

i CPScL Comment s

~MMI P Final Draft. Techn j.cal Specifications 7 5"i Cc:>1~(1<c f' Tup&

LCO Hier.:b:.:r: ... 0 . 6 ~,".( i (.)7. 12

~

~ ." ac:e Nunib='r: v/.J /-17 4 j .".

S': C ', I '.st;: l< ill 't / /

~ ~ W: 4 CU(tiiA< I t'L:

Iti I TEt!S /i. 7.7.b. 1:P 7-17> and <. ~. 12 b I ~ ~

C!HAtilGE "O. Ci5i.'" TO "cJ. 0':/. HEPA 1. 0'/.'l

~

II ('~

7. 7, v iP 7-18) ar)d 4. +. 12.

/ II TO II +/ II f  ! P ~ 16) CHAf!CE Jt.

1 r)

? H.= "; ILTERS COV" RED BY THESE TL~!0 SPECIFI ATIGfdS AjE v'.!  !.", F I CIEt~!T ~ ACCORDI!<>C T'0 BENEPIC LETTFP.

Gi>-13 .."!A.-(CH 2. 1~83. A VALUE QF 1. 0/ I S AP"..;QPRIATE FO'"", FILTERS ASSUf1E') TO BE 95'/'.

E." ICIEST. 1 H INCi3P. lECT VALUE f')AS .,ROtdEOUSLY S!. /t~I ITEMS) BY CP~~cL.

REFUELING OPERATIONS 5 I 3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST Au. 586 LIMITING CONDITION FOR OPERATION 3.9. 12 Two independent Fuel Handling Building Emergency Exhaust System Trains shall be OPERABLE.

APPLICABILITY:* Whenever irradiated fuel is in a storage pool.

ACTION:

With one Fuel Handling Building Emergency Exhaust System Train inoperable, fuel movement within the storage pool or crane operation with loads over the storage pool may proceed provided the OPERABLE Fuel Handling Building Emergency Exhaust System Train is capable of being powered from an OPERABLE emergency power source and is in operation and discharging through at least one train of HEPA filters and charcoal adsorber.

b. With no Fuel Handling Building Emergency Exhaust System Trains OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one Fuel Handling Building Emergency Exhaust System Train is restored to OPERABLE status.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVE I L LANCE RE UI REMEN TS 4.9.12 The above required Fuel Handling Building Emergency Exhaust System trains shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters nd charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months or (1} after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2} following significant painting, fire, or chemical release in any ventilation zone communicating with the system by:

HcPA) 1% ~~4~

1. Verifying that the cleanu system satisfies the in-place penetration and bypass eakage testing acceptance criteria of less than 0.05K and uses the test procedure guidance in Regulatory Positions C.5.a, C.5.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the unit flow rate is 6600 cfm a 10K during system operation when tested in accordance with ANSI N510-~.

/98o SHEARON HARRIS - UNIT 1 3/4 9"14

INAL FT REFUELING OPERATIONS 8HNPP RFVIRIAtI FUEL HANDLING BUILDING EMERGENCY EXHAUST SURVEILLANCE RE UIREMENTS Continued

4. 9. 12 (Continued)

After each complete or partial replacement of a charcoal adsorber bank, by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than in accordance with ANSI N510- for a halogenated hydrocarbon efrigerant test gas while operating the unit at a flow rate of 6600 cfm k 10K.

l98o i.o

/'HEARON HARRIS - UNIT 1 3/4 9-16

CPS'.L Comxnenta HNPP Proof and Review Technical Specifications Record Number: 735 Comment Type: ERROR LCO Number: 3.09.12 Page Number: 3/4 9-15 Section Number: 4.9.12.d.2 Comment:

DELETE "(UNLESS ALREADY OPERATING)".

Basis IN ORDER TO PROPERLY CONDUCT THIS TEST, THE FAN MUST BE STOPPED PRIOR TO THE START OF THE TEST.

SHNPP FANS DO NOT REDIRECT FLOW.THEREFORE)IF THE FAN IS ALREADY OPERATING$ NO CONCLUSION COULD BE REACHED REGARDING A SATISFACTORY COMPLETION OF THE TEST.

I

f) L FT REFUELING OPERATIONS SHNPP FUEL HANDLING BUILDING EMERGENCY EXHAUST P+t)Plr Kl JUL $ 86 SURVEILLANCE RE UIREMENTS Continued 4.9.12 (Continued)

Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide [enetration of le'ss than 1.0X when tested at a temperature of 30 C and at a relative humidity of 70K in accordance with ASTM D3803.

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying, within 31-days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 2978, meets the laboratory testing criteria of Regulatory Position ~.6.a of Regulatory Guide 1.52, Revision 2, March 1978, by showing a methyl iodide penetration of less than 1.0% when tested at a temperature of .

30'C and at a relative humidity of 70K in accordance with'STM 03803.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA fil-ters and charcoal adsorber bank is not greater than 4.1 inches water gauge while operating the unit at a flow rate of 6600 cfm a 10K,

2. Verifying that, on a High Radiation test signal, the system automatically starts and, directs its exhaust flow through the HEPA filters and charcoal adsorber banks,

3. Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/8 inch water gauge, relative to the outside atmosphere, during system operation at a flow rate of 66DO cfa a 10K,

4. Verifying that the filter cooling bypass valve is locked in the balanced position, and

5. Verifying that the heaters dissipate accordance with ANSI N510-~

40 i 4 N when tested in rP80

e. After each complete or partial replacement of a HEPA filter bank> by verifying that the unit satisfies the in-place penetration leakage testing acceptance criteria of less than 0.05K in accordance with ANSI N510-%8%8 for a DOP test aerosol while operating the unit at a flow rate o/6600 cfm f 10K.

i~8m SHEARON HARRIS - UNIT 1 3/4 9-15

Shearon Harris A'age:

Technical Specifications Resolution of Staff Comments Ori ginator: FO g, g; +/V ~l 7 Comment Date: q/o/ 7~ vuI,'I Comment:

nt~

4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a

~

representative sample of the tank's contents 7 d tion of radioactive material to the tank.

Oga Resolution Basis (w axM~Q, Resolution Acce ted:

NRC CPSL Date: (I g Date:

CP Bc. L Coxxxxne nt e Proof and Review Technical Specifications Dh'NPP Record Number: 765 Comment Type: IMPROVEMENT LCO Number: 3. 11 02. 01

~ Page Number: 3/4 11-9 Section Number:, TABLE 4.11-2 Comment:

EXTEND THE HORIZONTAL LINE IN THE CENTER OF THE ITEM THREE BLOCK OVER TO A POINT ABOVE THE LETTER Itb II Basis THIS CHANGE IS NEEDED TO PROVIDE GREATER CLARITY TO THE TABLE.

QI1~

p P'~ y I(

TABLE 4. 11-2 RADIOACTIVE GASEOUS WASTE SAHPLING AND ANALYSIS PROGRAH HINIHUH LOWER LIHIT Ogl SAHP LING ANALYSIS DETECTION (LLD)

GASEOUS RELEASE TYPE FREQUENCY FREQUENCY TYPE OF ACTIVITY ANALYSIS 'pCi/ml) aste as tprage Tank Each Tank Each Tank Principal Gama Emitters lxlO-~

Grab Sam le on aInmen urge or Vent Each PURGE Each PURGE Principal Gama Emitters lxlO-i Grab Sample H H-3 oxide lxl0-6

3. a. Plant Vent ~ ~

Principal Gama Ewtters lxlO-i Stack Grab Sample H-3 oxide lx10-6

b. Turbine Bldg H Principal Gama Emitters lxlO-i Vent Stack; Grab Sample Waste Pro- //-3 Qoxr c- 3)~

/p ld cessing Bldg Vent Stacks g 6'na &JtVe~i,S't~~k C l i+~

MSA fhu8 q/jgfjrg J)Z

4. All Release Types Continuous I-131 lx10-ta r~

as listed in l., 2.,

and 3. above Charcoal Sample I-133 1x10 to Continuous W Principal Gama Emitters lxlO->>

Particulate Sa le Continuous H Gross Alpha lxlo->>

Composite Par-ticulate Sam le Continuous 51-89, Sl-90 lxlo->>

Composite Par-ticulate Sample

CPScL. Cummen<m BHNPP Final Draft Technical SPeci+icatians Re=a:-v'juoiber: 7I Ph Comrrlerft Tvpe' t1P ROVE("iE!!T LCG fvuil/her: 8 /4. 01 01 01

~

N

~ Paae Number: Ef q/tt 1 Sec'tion flu.ob r: '9 3/4. 1. 1. 1 Comrrret1 'L ADD A NEI! SEN'I ENCE AFTER THE LJORDS " i nad ver't. en t di 1 utior'r e>>eni.". " AS FOLLOWS:

he un' "Prm" i-.-; used thr aughaut tfiese saeoiiiaaLi arr.= ta can~or-m wi Lh the r cacti vi tv inkac mi~L'i arr f~r-ovided bv Lhe NSSS suppli er,: 1000 Pam i s rou.; Lo 1/ dr'I ta f /I:.

BRs 1 ">>

TH:S C: ANCE IS I!j RESPOt1SE Tg AN NRC COtff1EfdT. IT PROV I DES THE NECEBSARY EQUI VALENCY I NFQRt1AT I QN.

BUT D('!ES NOT (."QNF USE THE ACTUAL SPECIF I CAT XQtd

3/4. 1 REACTIVITY CONTROL SYSTEMS SHNPP Rc I IAL DRAFT AU6 $ 86 BASES 3/4.1.1 BORA ION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A.sufficient SHUTDOWN MARGIN ensures that:, (1) the reactor can be made sub-critical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T . The most restrictive condi-occurs at EOL, with Tav at no, load operating temperature, and is asso-avg'ion ciated with a postulated steam'line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1770 pcm is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T less than 200 F, avg the reactivity transients resulting from a postulated steam line break cooldown are minimal,-but a 2000 pcm SHUTDOWN MARGIN is required to provide adequate protection for postulated inadvertent dilution events.

Analysis of inadvertent boron dilution at cold shutdown is based on:

1. all RCCA's in the core while the RCS, except the reactor vessel, is drained (i.e., not, filled), and

2. all RCCA's, except shutdown banks C and D, are fully inserted in the core while the RCS is filled.

In addition, by assuming the most reactive control rod is stuck out of the core, its worth is effectively added to the 2000 pcm shutdown margin in calculating the necessary soluble boron concentration.

3/4. l. 1.3 MODERATOR "TEMPERATURE COEFFICIENT The limit.ations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant conditions; i.e., the positive limit is based on core conditions for all rods withdrawn, BOL, hot zero THERMAL POWER, and the negative limit is based on core conditions for all rods withdrawn, EOL, RATED THERMAL POWER. Accordingly, veri-fication of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

p~ ~~~ s+ c rfQc 'h v Wy'

) '/~ ag/'~,

+~%7 cY) P~ QQ Qq ~ H+5 5 5 PP lier a '""o Pc.e s ~y. dL SHEARON HARRIS - UNIT 1 8 3/4 l-l

CPRL Comments RNPP Proof and Review Technical Specifications Record Number: 719 Comment Type: ERROR LCO Number: B 3/4.01.02 Page Number: B 3/4 1-2 Section Number:, B 3/4.1.2 Comment:

THE FIRST LINE IN PARAGRAPHS 2 AND 3 CHANGE "200 F" TO "350 F".

Basis THE CHANGE IS NEEDED FOR CONSISTENCY WITH LCO's 3.1.2.1 AND 3.1,2.2 FOR CSIP OPERABILITY'HE TEMPERATURES ON B 3/4 1-3 DO NOT NEED TO CHANGE BASED ON BORATED WATER SOURCE AVAILABILITYIN LCO's 3 '.2.5 AND 3.1.2.6. THIS IS THE SAME AS THE BYRON BASES.

SHNPP REACTIVITY CONTROL SYSTEMS aavisiON FINALD F 586 BASES MODERATOR TEMPERATURE COEFFICIENT Continued The most negative HTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the HDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the HDC associated with a core condition of all rods inserted (most positive HDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the HDC was then transformed into the limiting HTC value -42 pcm/ F. The HTC value of -33 pcm/ F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting HTC value of -42 pcm/4F.

The Surveillance Requirements for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confirm that the HTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4. 1. 1. 4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 5514F. This limitation is required to ensure: (1) the moderator temperature coefficient is within analyzed temperature range, (2) the trip instrumentation is within,its normal it operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RTNDT temperature.

3/4. 1. 2 BORAT ION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging/safety injection pumps, (3) separate flow paths, (4) bor' transfer pumps, and (5) an emergency power supply from OPERABLE el ge ratorp~

pe C r)

With the RCS average temperature abo ~F, minieuh of two boron injection flow paths are requi~ed to ensure si le func onal capability in the event an assumed failure renders one of the fl s inoperable. The boration capa-bility of either flow path is sufficient to provide a SHUTDOWN HARGEN from full ppm power e bor

' 'n expected operating conditions of 1770 pcm after xenon decay and cooldown to 200'F. The maximum expected boration capability requirement occurs at EOL from water be main conditions ed and requires 16800 gallons of 7000 in the boric acid storage tanks or 436,000 gal-lons M~~

tank ( WST). Qzoao -z2oof borated wa r be maintained in the refueling water storage QSd With the 5 tom ure bel w SHY'F olfe oron injection flow path is accept-able without single failure onsid ation on the basis of the stable reactivity SHEARON HARRIS - UNIT 1 B 3/4 1-2

QK CP8cL Comxnenta 9HNPP Proof and Review Technical Specification8 Record Number: 747 Comment Type: IMPROVEMENT LCO Number: 8 3/4.01.02 Page Number: B 3/4 1-2 L 3 Section Number: B 3/4. l. 2 Comment:

IN THE SECOND PARAGRAPH ON PAGE B 3/4 1-2 AND IN THE SECOND FULL'ARAGRAPH ON PAGE B 3/4 1-3 CHANGE "2000 ppm" TO "2000-2200 ppm".

Basis THIS CHANGE IS REQUIRED'OR CONSISTENCY BETWEEN THE BASES AND THE SP CIFICATIONS OF ECTION 3.1,2.

8HNP O~~

l'FVIS!

FINALD F REACTIVITY CONTROL SYSTEMS 586 BASES MODERATOR TEMPERATURE COEFFICIENT Continued The most negative MTC, value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the HDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the HDC associated with a core condition of all rods inserted (most positive HDC) to an all rods withdrawn condition and, a conversion for the rate of, change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the HDC was then transformed into the limiting MTC value -42 pcm/ F. The MTC value of -33 pcm/ F represents a conservative value (with corrections for'urnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting HTC value of -42 pcm/F.

The Surveillance Requirements, for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4. 1. 1. 4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 5514F. This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RTNDT temperature.

3/4. 1. 2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging/safety injection pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power supply from OPERABLE diesel generators.

ggo With the RCS average temperature above 40KF, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capa-bility of ~ ither flow path is sufficient to provide a SHUTDOWN HARGEN from expected operating conditions of 1770 pcm after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 16800 gallons of 7000 ppm lons of ~~

borated water be maintained in the boric acid storage tanks or 436,000 gal-borated water be maintained in the refueling water storage tank (RWST). Q oooo -zaoo~~n Qgd With the RCS temperature below 98YF, one boron injection flow path is accept-able without single failure consideration on the basis of the stable reactivity SHEARON HARRIS - UNIT 1 B 3/4 1-2

S8NP P REACTIVITY CONTROL SYSTEMS PcL/f gf Ph)

+

ltd DIM JUL..

BASES BORATION SYSTEMS (Continued condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single boron injection flow path becomes inoperable.

The limitation for a maximum of one charging/safety injection pump (CSIP) to be OPERABLE and the Surveillance Requirement to verify all CSIPs except the required OPERABLE pump to be inoperable below 335 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boron capability required below 2004F is sufficient a SHUTDOWN MARGIN of 1000 pcm after xenon decay and cooldown 00 , to 140 F. This maintained in the RWST.

of~

condition requires either 4900 gallons of 7000 m orated water be maintained in the boric acid storage tanks or 82,000 gall ns ppm bor ed water be gV The gallons given above are the amounts that ne to b >ntained in W tank in the various circumstances. To get the specific value, each value had added to it an allowance for the unusable volume of water in the tank, allowances for other identified needs, and an allowance for possible instrument error. In addition, for human factors purposes, the percent indicated levels were then raised to either the next whole percent or the next even percent and the gallon figures rounded off. This makes the LCO values conservative to the analyzed values. The specified percent level and gallons differ by less than 0.2X.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The BAT minimum temperature of 65'F ensures that boron solubility is maintained for concentrations of at least the 7750 ppm limit. The RWST minimum temperature is consistent with the STS value and is based upon other considerations since solubility not an issue at the specified concentration levels. 7~< HdS1 i'<<~~"

isSC ~~~~~r DhS A~TZb 'JES v~ yBAFAE-Y&c'Ad. AMEPWtlloysjS Sbk oIAmjWJ4<MP ~AT L8D.

The OPERABILITY oY one Boron Injection System during REFUELING ensures that ~ )m":~

this system is available for reactivity control while in MOOE 6. lJ"

-r Is.-4.

3/4. l. 3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

SHEARON HARRIS - UNIT 1 B 3/4 1-3

CPBc L Coxnxnenta

~HNPP Pr oof and Rev'iew. Tech.nival SWecitiaatione Record Number: 720 Comment Type: IMPROVEMENT LCO Number: B 3/4.01,02 Page Number: B 3/4 1-3 Section Number: B 3/4.1.2 Comment ADD TO THE EN OF THE NEXT TO LAST PARAGRAPH OF SECTION B 3 .).2 THE FOLLOWING SENTENCE:

The RWST temperature was selected to be consistent with analytical assumptions for containment heat load.

Basis THIS CHANGE IS TO PROVIDE ADDITIONAL INFORMATION FOR THE TECH SPEC USERS.

S8NPP REACTIVIT't CONTROL SYSTEMS P~i)g fP}K)

N6 istic fjIIt BASES BORATION SvSTEMS (Continued condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single boron injection flow path becomes inoperable.

The limitation for a maximum of one charging/safety injection pump (CSIP) to be OPERABLE and'the Surveillance Requirement to verify all CSIPs except the required OPERABLE pump to be inoperable below 335 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boron capability required below 200 F is sufficient to provide a SHUTOOWN MARGIN of 1000 pcm after xenon decay and cooldown from 200 F to 140'F. This maintained in the RWST.

of~

condition requires either 4900 gallons of 7000 ppm borated water be maintained in the boric acid storage tanks or 82,000 gallons 3040 Zgog ppm borated water be The gallons given above are the amounts that need to be maintained in in the various circumstances. To get the specified value, each value had added

~ tank to it an allowance for the unusable volume of water in the tank, allowances for other identified needs, and an allowance for possible instrument error. In addition, for human factors purposes, the percent indicated levels were then raised to either the next whole percent or the next even percent and the gallon figures rounded off. This makes the LCO values conservative to the analyzed values. The specified percent level and gallons differ by less than 0. ll.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment .after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The BAT minimum temperature of 65 F ensures that boron solubility is maintained for concentrations of at least the 7750 ppm limit. The RWST minimum temperature is consistent with the STS value and is based upon other considerations since solubility is not an issue at the specified concentration levels.

ACrCrtb m Bt 4>>>>iSmJT +Cnr AaAC.rrtCAC. AJOCrAPtlON5 FbR CkuJTVti4rtMT 7'S7 Mgr LOhD.

<<~~<~~"'pi The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MOOE 6.

3/4. l. 3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTOOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses 'are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

SHEARON HARRIS - UNIT 1 B 3/4 1-3

CPBc.L Comments HNPP Proof and Review Technical Specifications Record Number: 767 Comment, Type: IMPROVEMENT LCO Number: FIRE PROTECTION Page Number: VARIOUS ~i Section Number: fIRE PROTECTION Comment:

DELETE THE FIRE PROTECTION SYSTEM SPECIFICATIONS PER THE ATTACHED MARKUPS.

Basis c.-i PER PREVIOUS CPS(L LETTERS NLS-86-188 DATED JUNE 4, 1986 AND NLS-86-230 DATED JULY 22, 1986.

I'IHAL UIu INSTRUMENTATION SHNPP REVlStCN BASES 586 REMOTE SHUTDOWN SYSTEM Continued This capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR Part 50.

3/4. 3. 3. 6 ACCIDENT HONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that suffi-cient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.9?, Revision 3, "Instrumentation for Light-Mater-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG-0737, "Clarification of THI Action Plan Requirements," November 1980.

3/4. 3. 3. 7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that suffic%nt capa-bility is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to pro- .

tect control room personnel and is consistent with the recoaeendations of Regu-latory Guide 1.95, Revision 1, "Protection of Nuclear Power Plant Control Room, Operators Against an Accidental Chlorine Release," January 1977.

3/4 3 3 8 war 'apability is available for prompt detection of fires and that Fire Suppres Systems, that are actuated by fire detectors, will discharge extin-guishing age in a timely aanner. Preapt detection and suppression of fires will reduce the p tial for damage to safety-related equipment and is an integral element in t erall facility Fire Protection Program.

Fire detectors that are used to uate Fire Suppression Systems represent a more critically important coaponent plant's Fire Protection Program than detectors that are installed solely for fire warning and notification.

Consequently, the minimum number of OPERABLE detectors must be greater.

loss of detection capability for Fire Suppression ms, actuated by fire

'he detectors, represents a significant degradation of firn pro ion for any area.

As a result, She establisheent of a fire watch patrol must be in ted at an earlier stage than would be warranted for the loss of detectors that ide only early fire warning. The establishment of frequent fire patrols in 3/4.3.3.9 HETAL IMPACT MONITORING SYSTEM The OPERABILITY of the Metal Impact Honitoring System ensures that sufficient capability is available to detect loose metallic parts in the Reactor System SHEARON HARRIS - UNIT 1 B 3/4 3-5

CP RL Coxnxnenta RHNPP Proof and Review Technical Specifications Record Number: 778 Comment Type: ERROR LCO Number: NRC TYPOs Page Number: SEE LIST Section Number:,

Comment:

CHANGES HAVE BEEN MADE TO THE FOLLOWING PAGES TO CORRECT TYPOGRAPHICAL ERRORS MADE IN THE TYPING OF THE FINAL DRAFT TECH SPECS.

~ z-7 ~W

~2-9 ~

6<<

l

~ 3/4 3-22K O~

(~,

~

QH 3/4 6-3 > OPS 3/4 6-20 & 21 6-25 ~

.8 v'p'/4

& 26 6 II, ~j4+K &W ~L) 3/4 8-2 u OP

. 3/4 8-5

/4 3-3q q /OH~~

Basis TYPOGRPHICAL ERRORS

SHNPr pmillinN INSTRUMENTATION duL 586 MAf1 BASES ez78c i~PAcY AouimAiAb dt's ~

ontinued an vo d or mitigate damage to Reactor System components. The allowable out-of-service times and surveillance requirements are consistent with the recom-mendations of Regulatory Guide 1,133, "Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," Hay 1981.

3/4.3.3.10 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and con-trol, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Set-points for these instruments shall be calculated and adjusted.in accordance with the methodology and parameters in the ODCH to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of. General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3/4. 3. 3. 11 RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUHENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumenta-tion also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the GASEOUS RADWASTE TREATHENT SYSTEH.

The OPERABILITY and use of this instrumentation is consistent with the require-ments of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10-e pCi/ml are measurable.

3/4. 3. 4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will pro-tect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related com-ponents, equipment or structures.

SHEARON HARRIS - UNIT 1 B 3/4 3-6

CPRI Comments

'SHNPP

/

Proof and Review'echnical Specif ication s Record Number': 707 Comment Type: ERROR LCO Number: B 3/4.04.05 Page Number: B 3/4 4-3 Section Number:, B 3/4.4.5 Comment:

IN THE LAST PARAGRAPH OF THE SECTION> CHANGE "SPECIFICATION 6.9.2" TO "SPECIFICATION 4.4.5.5.c".

Basis THIS CHANGE IS TO PROVIDE CONSISTENCY WITH THE BODY OF THE SPECIFICATIONS.

Oi~

0 SHNPP REV!S!ON REACTOR COOLANT SYSTEM JUL 886 BASES STEAM GENERATORS (Continued)

The plant is expected to be operated in a manner such that the secondary cool-ant wi 11 be maintained within those chemistry limits found to result in negli-gible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant oper-ation would-be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-seconda~y leakage = 500 gallons per day per steam generator). Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage"type defects are unlikely with proper chemistry treatment of t%e second-ary coolant. However, even if a defect should develop in service, it found during scheduled inservice steam generator tube examinations. Plugging will be will be required for all tubes with imperfections exceeding the plugging limit of 40K of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degra-dation that has penetrated 20K of the original tube wall thickness.

Whenever the results of any stea generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission in a Special Report pursuant to Specification within 30 days and prior to resumption of plant operation, Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examina-tions, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

3/4. 4. 6 REACTOR COOLANT SYSTEM LEAKAGE 3/4. 4. 6. 1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coo]ant pressure boundary.

These Detection Systems are consistent'with the recoaeendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"

May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

SHEARON HARRIS - UNIT 1 8 3/4 4"3

C P8c.L Comxnent'a

""HNPP Proof and Review Technical Specification8 Record Number: 754 Comment Type: IMPROVEMENT LCO Number: B 3/4 4-6 Page Number: B 3/4 4-6 &

P7 ll Section Number: B 3/4.4.9 Comment:

IN THREE PLACES, CHANGE "Figures 3.4-2 and 3.4-3" TO "Figures 3,4-3 and 3.4-2 and Table 4.4-6".

Basis THESE CHANGES PROVIDE A MORE COMPLETE REFERENCE TO ALL OF THE PLACES WHICH PROVIDE HEATUP AND COOLDOWN LIMITATION DATA AND MAKE THE REFERENCES TO THE HEATUP AND COOLDOWN CURVES GRAMMATICALLY CORRECT.

~4 P g(R,6

+~IfPIgh)

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY Continued) distinction between the radionuclides above and below a half-life of 15 minutes.

For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNQARY under any accident condition.

Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to per-form the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties. The counter should be reset to a reproducible efficiency versus energy. It is not necessary to identify specific nuclides. The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about 1 week, and about 1 month.

Reducing T to less than 500'F prevents the release of activity shoul'd a steam generator tube rupture occur, since the saturation pressure of the reactor cool-ant is below the lift pressure of the atmospheric steam relief valves. The ~

Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G, and 10 CFR 50 Appendix G. 10'CFR 50, Appendix G also addresses the metal temperature of the closure head flange and vessel flange regions. The minimum metal temperature of the closure flange region should be at least 1204F higher than the limiting RT NDT for these regions when the pressure exceeds 20K (621 psig for Westinghouse plants) of ~the reservice hydrostatic test pressure. For Shearon Hats&nit-1; the m>nimum temperature of the closure flange and vess~~ge regions is 1204F because the lim'T NOT is 0 F (see Tab -BW/4 4-1). The .Shearon Harris Unit eatu and ol down

)ski 0&

sh 'igures 3.4- and 3.4- a o impact b mit.

kvD Tae~g g

1. he reactor coolant tern eratui t heatu and with the exception of the pressurizer) be ed in accordance with Figures 3.4-R and 3.4-$ for hewervice-eriod specified thereon: ~P' ~o i~ac KW-/

aa able combinations of pressure and tern specific temperature c ange ra es are e ow and to the right of the limit lines shown. Limit. lines for cooldown rates between those pre-sented may be obtained by interpolation; and SHEARON HARRIS - UNIT 1 B 3/4 4-6 dl( ~

g

REAC/OR COOLANT SYSTEM BASES PRESSURE!TEMPERATURE LIMITS (Continue

b. Figures 3.4 % and 3.4-X define limits to assure prevention of non-'ductile failure only. For normal operation, other inherent /"

plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods pro-vided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70oF

4. The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200 F/h, respectively. The spray shall not be used if the tem-perature difference between the pressurizer and the spray fguid is greater than 625'F, and

5. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness testing of the ferritic materials in the reactor vessel was performed in accordance with the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code. These properties are then evaluated in accordance with the NRC Standard Review Plan.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTN>T, at the end of-4 effective full power years (EFPY} of service life. The 4 EFPY service life period is chosen such that the limiting RTN>T at the 1/4T location in the core region is greater than the RTNDT of the limiting unirradiated material. The selection of such a limiting RTN>T assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNOT, the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 HeV) irradiation can cause an increase in the RTNOT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of LRTNOT computed by either Regulatory Guide 1.99, Revision 1, "Effects of Residual Elements on Predicted Radiation Oamage to Reactor Vessel Materials," or the Westinghouse SHEARON HARRIS " UNIT 1 B 3/4 4-7

REACTOR COOLANT SYSTEM FIN L DRAFT, SHNPP 0%/IO I +hl BASES PRESSURE/TE!'PERATURE LIMITS (Continued Copper Trend Curves shown in Figure 8 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-Z~and 3.4- ~include predicted adjustments for this shift in RTNpT at the end of 4 EFPY as well as adjustments for possible errors in piacg t'-c the pressure and temperature sensing instruments.

Values of ORTNpT determined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed and evaluated in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50; Appendix H. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. The lead factor repre-sents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and thy with-drawal time of the capsule. The heatup and cooldown curves must be regilculated when the DRTNOT determined from the surveillance capsule exceeds the c8culated GRTNp T for the equi va1 ent caps ul e radi ati on exposure, Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor, operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference tempera-ture, RTNpT, is used and this includes the radiation-induced shift, hRTNpT, correspondingto the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the SHEARON HARRIS - UNIT 1 8 3/4 4-11

Shearon Harris Technical Specifications Resolution of Staff Comments Originator: EP >

~<I~<<~ by E~lt<+ Page: 8 /t' -0 Comment Date: 8//

Comment:

Values of hRTNDT determined in this manner may be used until the results from the material surveillance pr grXm, evaluated according to ASTH E185, are available. Capsules will removed and evaluated in accordance with the requirements of ASTH E18 73 and 10 CFR Part 50, Appendix H. The surveillance specimen withdrawal scheduue is shown in Table 4.4-5. The lead factor repre-sents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the with-drawal time of the capsule. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure.

Resolution Basis Resolution Acce ted:

NRC CPBL Date: Date:

Shearon Harris Technical Specifications Resolution of Staff Comments Originator'P Page: 8 /8 'f /~

Comment Date: g /i/O'P Comment:

3/4.4. 10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASNE Code Classreadiness 1, 2, and 3 of components ensure that the structural integrity and operational these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASHE Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commis-sion pursuant to 10 CFR 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections~ in accordance with Section XI of the ASIDE Boiler and Pressure Vessel Code, 1977 Edition and Addenda through Summer 1978.

Resolution Basis q,'lo I

~

Resolution Acce ted:

NRC CPSL Date: Date:

CPBc.L Commenta proof an 1 Review Technical Specifications Record Number: 751 Comment Type: IMPROVEMENT LCO Number: B 3/4.06.01.04 Page Number: B 3/4 6-1 Section Number: I B 3/4.6.,1.4 Comment:

REWORD THE BEGINING OF THE SECOND PARAGRAPH AS FOLLOWS:

line break event is 40.9 psig using a value of 1.9 psig for initial positive containment pressure. However, since the instrument.....

Basis THIS CHANGE IS MADE TO MAKE THIS DISCUSSION MORE ACCURATE AND TO PROVIDE THE EXACT RESULTS OF THE LIMITING CALCULATION.

iiii~L 3/4. 6 CONTAINMENT SYSTEMS SHNPP P ~ I )Q I ~h)

BASES 3/4.6. 1 PRIMARY CONTAINMENT

'/4. 6. 1. 1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNOARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions 3/4.6. 1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P . As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to OB5. L ,

performance of the periodic test, to account for possible degradation of a'uring the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50.

3/4.6. 1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6. 1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the contain-ment structure is prevented from exceeding its design negative pressure dif-ferential with respect to the outside atmosphere of -2 psig, and (2) the con-tainment peak pressure does not exceed the design pressure of 45 psig.

uPA& <

The maximum peak pressure expecte to be obtained from a postulated main steam 5 f(;

line break event is psigg value of 1.9 psig woo wed for initial posi" ~?Q',

tive containment pressure.

~ 4y. . However, since the instrenent tolerance for containment pressure is 1.32 psig and the high-one setpoint is 3.0 psig, the pressure limit was reduce from the high-one setpoint by slightly more than the tolerance and was set at 1.6 psig. This value will prevent spurious safety injection signals caused by instrument drift during normal operation. 7~x -/" > ~~s erosru vn sx comisre~r aATR wE iQITIRc hsdv&PN44V oF Thug Ac +I DLeJ7 44gcyggs bfi SHEARON HARRIS - UNIT 1 B 3/4 6-1

CP LL Commenta Jjt HNPP Proof and Reviewer Technical Specif icationa Record Number: 721 Comment Type: IMPROVEMENT LCO Number: B 3l4. 06'. 01. 04 Page Number: B 3/4 6-1 Section Number: B 3/4.6.1.4 Comment:

ADD TO THE END OF THE SECOND PARAGRAPH THE FOLLOWING SENTENCE:

The -1" wg was chosen to be consistent with the initial assumptions of accident analyses.

Basis THIS CHANGE IS TO PROVIDE ADDITIONAL INFORMATION FOR TECH SPEC USERS.

IiiI~L L.t I 3/4.6 CONTAINMENT SYSTEMS SHNPP g&,')Q!A<)

BASES 3/4.6. 1 PRIMARY CONTAINMENT 3/4. 6. 1. 1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNOARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4. 6. 1. 2 CONTAINMENT LEAKAGE The limitations on containment leakage. rates ensure that the total containment leakage volume will not exceed the ~alue assumed in the safety analyses at the peak accident pressure, P . As an added conservatism, the measured overall leakage rate is further limited to less than or equal to OB5. L, a'ntegrated performance of the periodic test, to account for possible degradation of 'a'uring the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50.

3/4. 6. 1. 3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6. 1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the contain-ment structure is prevented from exceeding its design negative pressure dif-ferential with respect to the outside atmosphere of -2 psig, and (2) the con-tainment peak pressure does not exceed the design pressure of 45 psig.

uPA6 4 The maximua peak pressure expecte to be obtained from a postulated main steam line break event is psigg value of 1.9 psig wed for initial posi-tive containment pressure.

However, since the instreaent tolerance for containment pressure is 1.32 psig and the high-one setpoint is 3.0 psig, the pressure limit was reduced from the high-one setpoint by slightly more than the tolerance, and was set at 1.6 psig. This value will prevent spurious safety injection signals caused by .

instrument drift during normal operation. lax -/ ~g uA cpofEN 75 Bl ~o~ive~r IrATPhC h$ 8u~P7l~ OF @AC AC+ID~ 44hlyg+5 SHEARON HARRIS - UNIT 1 8 3/4 6-1

Shearon Harris Technical Specifications Resolution of Staff Comments Originator: fg 5 Page: < /S Comment Date: (+/g C, Comment:

Section B 3/4.6.2, Item B 3/4.6.2.3 Containment Cooling System. Item (I)

Page B 3/4 6-3: should be deleted from the bases since operability of the containment fan coolers does not ensure the containment air temperature will be maintained within limits during normal operation. The non-nuclear safety fan coil units are required for normal operation.

I C~~

Resolution Basis

~F'g

~~8A c

Q4. qi) E;5%4 ( Port C~~

~P JgD )

Resolution Acce ted:

NRC cpaL Date: Date:

fiiNL UISI I CONTAINMENT SYSTEMS BASES CONTAINMENT VENTILATION SYSTEM Continued g/c4 gross leakage failures could develop. The 0.60 L leakage limit of Specifica- ~

a tion 3.6.1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type 8 and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM s/c <

The OPERABILITY of the Containment Spray System ensures that containment de-pressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Fan Coolers are redundant to each othe~ in providing post-accident cooling of the containment atmosphere.

However, the Containment Spray System also provides a mechanism for removing.

iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2

. /

SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consis-tent with the iodine removal efficiency assumed in the safety analyses. With 100,000 gallons of water in the iNST, sufficient head pressure, approximately 70 feet of water, is available at the eductor.

3/4.6.2. 3 CONTAINMENT COOLING SYSTEM . 4/(i l The OPERABILITY of the Containment Fan Coolers ensures that: the containment air temperature will be maintained within limits during no a operation, and (2) adequate heat removal capacity is available when opera ed in conjunction with the Containment Spray Systems during post-LOCA condi ions.

The Containment Fan Coolers and the Containment'Spra ystem are redundant to each other in providing post-accident cooling o containment atmosphere.

SHEARON HARRIS - UNIT 1

CONTAINMENT SYSTEMS HiQL PIN BASES CONTAINMEN, VENTILATION SYSTEM Continued gross leakage failures could develop. The 0. 60 L leakage limit of Specifica-a tion 3.6. 1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4. 6. 2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4. 6. 2. I CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures 'that containment de-pressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Fan Coolers are redundant to each other in providing post-accident cooling of the containment atmosphere.

However, the Containment Spr'ay System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are consis-tent with the iodine removal efficiency assumed in the safety analyses. With 100,000 gallons of water in the RWST, sufficient head pressure, approximately 70 feet of water, is available at the eductor.

3/4. 6.2. 3 CONTAINMENT COOLING SYSTEM The OPERABILITY of the Containment Fan Coolers ensures that)(

~ adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post-LOCA conditions.

The Containment Fan Coolers and the Containment Spray System are redundant to each other in providing post-accident cooling of the containment atmosphere.

SHEARON HARRIS " UNIT 1 8 3/4 6-3

C:PScL, Dnmmen<m SHNPP Final Draft Technical Speci+ ications Re..a. d t!u:;iber". 7'~i 0='amment Tyae: FRRCR L CO t lurrrber": 8 ~/4. r)6. 0"='. Pao~= Number: 8 3/4 Bea t i ar'i Nurr.ber': 8 .>/4. 6 ..'2 Caiiiiiiefi t:

D. '-'TE THE LAST BENTEtilCE QF THF BASES PARAGRAPH 2/ t ~ o ~ ."; ~ 2 Qt'J THE SPRAY ADDITIVE BYBT t'! Al'JD REPLACE IT l~JI Tr!:

"".he RtJBT 1 eval o4 r!i6.000 aal lans pr-avides

~adequate test aandi tians ta demonstrate thai the

<? ar~ i" ='.t e i s w ii.!iin the max imum arid minimuiA assuoratians a-. the analyses."

S~si s

. HIS CHAtlHE IS NECESSr"-rRY TO BE COr!BISTENT KrITH Tf!E CURREt!T VJORDIr'JB QF THE SPECIFICATION. THE SPEC IF I CAT I ON l JAB- CHAt ISED IN JULY r-.ND THE CHAt~BE i:AB BEE/! AGREED TQ ErY THE NRR STAFF.

gti

/

HiNL lll5 CONTAINMENT SYSTEMS BASES CONTAINMENT VENTILATION SYSTEM (Continued gross leakage failures could develop. The 0. 60 L leakage limit of Specifica-a tion 3.6. 1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined

'otal for all valves and penetrations subject to Type B and C tests.

3/4.6.2 OEPRESSURIZATION ANO COOLING SYSTEMS 3/4. 6. 2. I CQNTAINMEHT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment de-pressurization and cooling capability wi ll be available in the event of a LOCA or steam line break. The pressure reduction and resultant lour containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Fan Coolers are redundant to each other in providing post-accident cooling of the containment atmosphere.

However,'. the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4. &. 2. 2 SPRAY AOOITIVE SYSTEM The, OPERABILITY of the Spray Additive System ensures that sufficient HaOH is added to the containment spray in the event of a LOCA. The limits on HaOH volume and concentration ensure a pH value of between 8.5 and 1l.O for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allo~ance for solution not usable because ~f tank discharge line location or other physical char acter istics. These assumptions are consis-tent with the iodine removal efficiency assumed in the safety analyses.

Yhc Ras 7 I eve I

'Pico@>dt's ~g~>> I c ++sf coral>k>>>wg ~ Vcms>>shostc wet +gg 3/4.6.2.3 COHTAIHMEHT COOLIHG SYSTEM ' " ~~ss p~"' +4 >yscs.

The OPERABILITY of the Containment Fan Coolers ensures that: (1) the containment air temperature will be maintained within limits during normal operation, and (2) adequate heat removal capacity is available when operated in conjunction with the Containment Spray Systems during post"LOCA conditions.

The Containment Fan Coolers and the Containment Spray System are redundant to each other in providing post-accident cooling of the containment atmosphere.

SHEARON HARRIS -. UNIT 1 B 3/4 6-3

CP &,L Comxnenta BHNPP Proof and Review. Technical Specifications I

Record Number: 766 Comment Type: IMPROVEMENT LCO Number: B 3/4.06.05 Page Number: B 3/4 6 4 Section Number:

P B 2/4.6.5 Comment:

CHANGE THE TITLE OF THE SECTION TO "VACUUM RELIEF SYSTEM".

Basis THIS CHANGE IS MADE TO PROVIDE CONSISTENCY WITH THE BODY OF THE'PECIFICATIONS.

FLi~AL DRAFF I CONTAINMENT SYSTEMS SHNF P

~Ri!S)~N BASES JUL 8%

As a result of this redundancy in cooling capability, the allowable out-of-service time requirements for the Containment Fan Coolers have been appropri-ately adjusted. However, the allowable out-of-service time requirements for the Containment. Spray System have been maintained consistent with that assigned other inoperable ESF equipment since the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere.

3/4. 6. 3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50. Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environ-ment will be consistent with the assumptions used in the analyses for a LOCA.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to main-ment Following a LOCA,'e VACUUI4 RELIE SP'/4.6.6

'l tain the hydrogen concentration within containment below its flaaeable limit during post-LOCA conditions."'ither recombiner unit is capable of controlling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within con-tainment, This hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7,

. 2, of Combustible Gas Concentrations in Contain-ovember 1978.

The OPERABILITY of the primary containment to atmosphere vacuum relief valves ensures that the containment internal pressure does not become more negative than -1.93 psig. This condition is necessary to prevent exceeding the con-tainment design limit for internal vacuum of -2 psig.

SHEARON HARRIS - UNIT 1 B 3/4 6-4

Shearon Harris Technical Specifications Resolution of Staff Comments Originator: EP Su/li"a>> >g Elllon Q.harm.ej) ~

page: $ -a/q 7- f Comment Date: g/jP t, Comment:

The visual inspection frequency is based upon maintaining a constant level of snubber protection to each safety-related system during an earthquake or severe transient. Therefore, the required ins ection interval varies inversely with the observed snubber failures n a sven and is determined by the number of inoperable snubbers found during an snspec son eac s s e In order to establish the-inspection frequency for each type of snubber n a safety-re a e e it was assumed that the frequency of snubber failures and >n> ia sng f b could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has SHEARON HARRIS - VNIT 1 B 3/4 7"4 Resolution Basis Resolution Acce ted:

iigc R.M CPSL Date: Date:

0 CP Bc.L Coxnxnenta gK HNPP Proof and Bevierv'ech nical S Pecif ication 8 Record Number: 767 Comment Type: IMPROVEMENT LCO Number: FIRE PROTECTION Page Number: VAR IOUS ,

Section Number: FIRE PROTECTION Comment: yv'/q 7-~7-e~ .~'!

DELETE-THE FIRE PROTECTION SYSTEM SPECIFICATIONS PER THE ATTACHED MARKUPS. '/q Q-'13 fl>i~ 7 ~

): i~

Basis C-/

PER PREVIOUS CP&L LETTERS NLS-86-188 DATED JUNE 4>

1986 AND NLS-86-230 DATED JULY 22, 1986.

SHNPP PLANT SYSTEHS

'EVISIO<

rII<II,LO F gg 186 BASES SEALEO SOURCE CQNTAHINATION Continued limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Haterial sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Surveillance, Requirements commensurate with the probability of damage to a source in that group. Those sources that are frequently handled are required to be tested more often than those that are not. Sealed sources that are con-tinuously enclosed within a shielded mechanism (i.e., sealed-sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.1O sup ssion capability is available to confine and extinguish fires o~arring .

in any tion of the facility where safety-related equipment is locate, The Fire Suppr ion System consists of the fire protection water supply ariiK dis-tribution sys preaction and multicycle sprinkler systems, fire hose stations, and yard fire hy ts. The collective capability of the Fire Suppression Sys-tems is adequate to m 'mize potential damage to safety-related equipment and is a major element in th cility Fire Protection Program.

In the event that portions of t ire Suppression Systems are inoperable, alternate backup fire-fighting equi t is required to be made available in the affected areas until the inoperable uipment is restored to service. When the inoperable fire-fighting equipment is nded for use as a backup means of fire suppression, a longer period of time is a d to provide an alternate means of fire fighting than if the inoperable equ nt is the primary means of fire suppression.

The Surveillance Requirements provide assurance that the a)n OPERABILITY requirements of the Fire Suppression Systems are met.

In the event the Fire Suppression Water System becomes inoperable, imme 'e 3/4. 7. 11 ensures t a e confined

' or adequately retarded from spreading to adjacent po~tions of the design features minimize the possi-bility of a single fire rapidly involving of the, facility prior to detection and extinguishing of the fire. The fire a tions are SHEARON HARRIS - UNIT 1 B 3/4 7-6

PLANT SYSTEMS BASES dampe considered functional when the visually observed condition is the same as the a ned condition. For those fire barrier penetrations that are not in the as"des ondition, an evaluation shall be performed to show that the modification has no , ed the fire rating of the fire barrier penetration.

Ourfng periods of time when a barrier is not functio ther: (1) a contin-uous fire watch is required to be maintained in the vicinity affected barrier, or (2} the fire detectors on at least one side of the affe rier 3/4. 7. 12 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipmen~ll not be subjected to temperatures in excess of their environmental qualifiC&ion temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits do not include an allowance for instrument errors.

3.4.7.13 ESSENTIAL SERVICES CHILLEO WATER SYSTEM The OPERABILITY of the Emergency Service Chilled Water System ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.

SHEARON HARRIS - UNIT 1 B 3!4 7-7

CPS',L Coxnmenta S~pp px-oof and Review Technical Specifications I

1 Record Number: 771 Comment Type: ERROR LCO Number: B 3/4.08.01 Page Number: B 3/4 8-1 Section Number: B 3/4.8.1 Comment:

IN THE SECOND LINE OF THE SECOND PARAGRAPH> CHANGE "five" TO "six".

Basis ANOTHER TRANSMISSION LINE HAS RECENTLY BEEN PLACED INTO SERVICE.

3/4.8 ELECTRICAL POWER SYSTEMS AR~lS)C "~

i(iNL BASES 3/4.8.1, 3/4.8.2 AND 3/4.8.3 A.C. SOURCES D.C. SOURCES AND ONSITE POWER ISTRIBU ION The OPERABILITY of the A.C. and D.C power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. Th'e minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

~ 5/)(

The currently has ~

switchyard is)designed using a breaker-and-a-half scheme. The switchyard connections with the CPAL transmission network; each of these transmission lines is physically independent. The switchyard has one connection with each of the two Startup Auxiliary Transformers and each SAT can be fed directly from an associated offsite transmission line. The Startup Auxiliary Transformers are the preferred power source for the Class lE ESF buseg, . The minimum alignment of offsite power sources will be maintained such thA ..at least two physically independent offsite circuits are available. The cally independent circuits may consist of any two of the incoming transmission

~ physi-lines to the SATs (either through the switchyard or directly) and into the Class 1E system. As long as there are at least two transmission lines in ser-vice and two circuits through the SATs to the Class lE buses, the LCO is met.

During MODES 5 and 6, the Class 1E buses can be energized from the offsite transmission net work via a combination of the main transformers, and unit auxiliary transformers. This arrangement may be used to satisfy the require-ment of one physically independent circuit.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coin-cident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times .

are based on Regulatory Guide 1.93, "Availability of Electrical Power Sources,"

December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE dieseT generato~

as a source of emergency power, are also OPERABLE, This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term, verify, as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.

SHEARON HARRIS - UNIT 1 B 3/4 8-1

oui

.." ".".'CPRL-: Coxnxne nt s

'NPP Proof and. Review Technical Syecif ication s Record Number: '740 Comment Type: ERROR LCO Number: ,B 3/4.08;Ol. '.Page Number: B 3/4 8-1

. Section Number: ~ B 3/4 8.1 Comment:

IN THE IAST'PARAGRAPH OF THE PAGE) DELETE THE PHRASE "AND THAT THE STEAM DRI'VEN AUXILIARY

':FEEDWATER'PUMP IS OZPRABLE."

Basis THIS CHANGE IS REQUIRED F R CONSIS CY WITH THE ACTION SThTEMENT OF '3.8.1 l. IRECTED BY MR.

J.T. BEARD OF THE NRC) THE REQUIREMENT THAT THE STEAM DRIVEN AUXIIIARY FZEDWATER. PUMP BE OPERABLE WAS CHANGED TO PROVIDE DIRECTION ONLY IF ALL THREE FEEDWATER PUMPS ARE INOPERABLE.

s~ iu ~ ~ t'((NL 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2 AND 3/4.8.3 A.C. SOURCES D.C. SOURCES AND ONSITE POWER Dl 5 RI 0 10N The OPERABILITY of the A.C. and D.C power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.CD power sources and distribution systems satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

currently has ~ 5/P'he switchyard is)designed using a breaker-and-a-half scheme. The switchyard connections with the CPEL transmission network; each of these transmission lines is physically independent, The switchyard has one connection with each of the two Startup Auxiliary Transformers and each SAT can be fed directly from an associated offsite transmission line. The Startup Auxiliary Transformers are the preferred power source for the Class lE ESF buseg.. The minimum alignment of offsite power sources will be maintained such thR,.at least two physically independent offsite circuits are available. The @go physi-cally independent circuits may consist of any two of the incoming transmission lines to the SATs (either through the switchyard or directly) and into the Class 1E system. As long as there are at least two transmission lines in ser-vice and two circuits through the SATs to the Class 1E buses, the LCO is met.

During MODES 5 and 6, the Class lE buses can be energized from the offsite transmission net work via a combination of the main transformers, and unit auxiliary transformers. This arrangement may be used to satisfy the require-ment of one physically independent circuit.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the po~er sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coin-cident with an assumed loss-of-offsite power and single failure of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times .

are based on Regulatory Guide 1.93, "Availability of Electrical Power Sources,"

December 1974. When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systemssystems. ~rains, components and devices, that depend on the rem 'i.ag OPERABLE dieseT generato as a source of emergency power, are also OPE BLE This req provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term, verify, as used in this context means to administratively check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.

SHEARON HARRIS - UNIT 1 B 3/4 8-1

CI )

CP Bc.L Comment a

~HNPP Proof and Review Technical Specifications Record Number: 736 Comment Type: ERROR LCO Number: B 3/4.08.01.01 Page Number: B 3/4 8-2 Section Number: B 3/4.8.1.1 Comment:

IN THE SECOND PARAGRAPH OF THE PAGE, CHANGE "IN ACCORDANCE WITH" TO BASED UPON".

Basis THE LATEST NRC STAFF GUIDANCE WAS PROVIDED FOR THE SHNPP DIESEL SPECIFICATION. THIS GUIDANCE DIFFERS IN SOME DETAILS FROM THAT PROVIDED IN REG GUIDE 1,108.

gIi

( ql

ELECTRICAL POWER SYSTEMS p~g~~iWh'I 586 IM BASES A.C. SOURCES. D.C. SOURCES AMD ONSITE POWER DISTRIBUTION Continued The OPERABILITY of the minimum specified A.C. and D.C. power sources and asso-ciated distribution systems during shutdown and refueling ensures that: (1) the

'he facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient in entation and control capability is available for moni ~g~~usmc& aining tte unit s4Wus.

~~~ F( C

)

The Surveill nce Requirements for demonstr ting the OP RABILI Y f the diesel generators a e r ommendations of Regulatory Guides l. 9, "Selection o iesel Generator pacity for Standby Power Supplies,"

December 1979; .

" >c Testing. of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977 as modified in accordance with the guidance of IE Notice 85-32, April 22, 1985; and 1. 137, "Fuel-Oil Systems for Standby Diesel Generators," Revision 1, October 1979.

The Surveillance Requirement for demonstrating the OPERABILITY of the Station batteries are based on the recommendations of Regulatory Guide 1. 129, Mainte-nance Testing,and Replacement of Large Lead Storage Batteries for Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."

Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares the battery capacity at that time with the rated capacity.

Table 4.8-2 specifies the normal limits for each designated'pilot cell and each connected cell for electrolyte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2. 13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2. 13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 belo~ the manufacturer's ful'1 charge specific gravity, ensures the OPERABILITY and 'capability of the battery.

Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days.

During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than 0.020 below the manufacturer's recommended full charge specific gravity, ensures that the decrease in rating will be less than the safety SHEARON HARRIS - UNIT 1 B 3/4 8-2

4 I

lh

CP8c L Coxnxnenta HNPP Proof and Review Technical Specifications Record Number: 741 Comment Type: IMPROVEMENT LCO Number: B 3/4.08.04 Page Number: B 3/4 8-3 Section Number: B 3/4 .8.4 Comment:

IN THE SECOND PARAGRAPH) DELETE ALL REFERENCES TO FUSES PER THE ATTACHED MARKUP.

Basis THIS CHANGE IS NECESSARY DUE TO THE CHANGE PREVIOUSLY APPROVED BY THE NRC WHICH DELETED SURVEILLANCE TESTING OF FUSES. WHEN THE CHANGE WAS MADE TO THE SURVEILLANCESI THE BASES CHANGES WERE INADVERTANTLY MISSED.

ELECTRICAL POWER SYSTEMS

","-.',:FINAL D FT 85" .

BASES A.C. SOURCES, D.C. SOURCES AND ONSITE POWER DISTRIBUTION Continued margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an accept-able limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.

3/4.8.4 ELECTRICAL E UIPMENT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are protected by either deenergizing circuits not required during reactor operation or by demon-strating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.

The Surveillance Requirements applicable to lower voltage circuit breakers 4wea provide assurance of breaker ea4akeee reliability by testing at gast one representative sample of each manufacturer's brand of circuit breaker aa44er

~

feee Each manufacturer's molded case and metal case circuit breakers end~

%wee are grouped into representative samples which are then tested on a rotat-and treat purposes.

each group as a separate type of breaker ~~ for surveil'lance The bypassing of the motor-operated valves thermal overload protection during accident conditions by integral bypass devices ensures that safety-related valves will not be prevented from performing their function. The Surveillance Require-ments for demonstrating'the bypassing of the thermal overload protection during accident conditions are in accordance with Regulatory Guide 1.106, "Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1, March 1977.

SHEARON HARRIS - UNIT 1 B 3/4 8-3

CF'ScL Cnmmmnt a F'in',l I

RHNF'F" De-aa+4 Teec=Ani.c=ml R~eemk+ic=eat'.inn Re~or d WumL>er: 737 Comment Type- ERROR LCO Number: 5. 07. 01 Page Number: 5-8 Sec t. i on Number: TABLE 5. 7- l Comment:

IN THE DESIGN CYCLE OR TRANSIENT COLUMN FOR THE REACTOR COOLANT SYSTEM 10 AUXILIARY SPRAY ACTUATION CYCLES, CHANGE ",625 F" TO "Greater than 320 F but 1ess than 625 F."

Bdsl s THIS CHANGE IS REQUIRED TO MAKE THE SPECS MORE A(:CURATE. THE CYCLE IS FOR THE TEMPERATURE RANGE

>320 F TO <625 F. ACTUATION BELOW 320 F DOES HOT APPLY TO TH1S CYCLIC LIMIT.

6Q p/1 7

CP8c,L Comxnents RHNPP Proof and Review Technical Specifications Record Number: 737 Comment Type: ERROR LCO Number: 5.07.01 Page Number: 5-8 Section Number: TABLE 5.7-1 Comment:

IN THE DESIGN CYCLE OR TRANSIENT COLUMN FOR THE REACTOR COOLANT SYSTEM 10 AUXILIARY SPRAY ACTUATION CYCLES, CHANGE ",625 F" TO "Greater than 320 F but less than 625 F."

Basis THIS CHANGE 1S REQUIRED TO MAKE THE SPECS MORE CCURATE ~ THE CYCLE IS FOR THE TEMPERATURE RANGE 20 F T ) 0 F.

~ ACTUATION BELOW 320 F DOES NOT APPLY TO IS CYCLIC LIMIT.

~

lg yN(~

TABLE 5.7-1 tll m COMPONENT CYCLIC OR TRANSIENT LIMITS CD CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT Reactor Coolant System 200 heatup cycles at < 100'F/h Heatup cycle - "T from < 200"f and 200 cooldown cycles at to > 550'F.

< 100 F/h. Cooldown cycle - T from

> 550'F to < 200 F 200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200 F/h. temperatures from > 650'F to

< 200'F.

200 loss of load cycles, without > 15K of RATED THERMAL POWER to iaeediate Turbine or Reactor trip. OX of RATED THERMAL POWER.

c 3) Q) 40 cycles of loss-of-offsite Loss-of-of fs i te A.C. el ectri ca A.C. electrical power. ESF Electrical System.

1 r gX

~n~.

80 cycles of loss of flow in reactor coolant loop.

400 Reactor trip cycles.

one Loss coolant 10'o of only pump.

OX of one reactor RATED THERMAL POWER.

I 5D

~~

10 auxiliary spray Spray water temperature differential actuation cycles. ~ 4egjQQQ. Cea<g nable ~F e~r CZARS ma~ C'Z+g 200 leak tests. Pressurized to > 2485 psig.

d~

10 hydrostatic pressure tests. Pressurized to > 3107 psig.

Secondary Coolant System 1 steam line break. Break in a > 6-inch steam line.

10 hydrostatic pressure tests. "i '" Pressurized to > 1481 psig.

CPScL Comments SHNPP Final Dra ft Technical Specif ication R~~cor d Number: 737 Comment Type- EPROR LCO Number: 5.07.01 Page Number: 5-8 Section Number: TABLE 5.7-1 Comment:

IN THE DESIGN CYCLE OR TRANSIENT COLUMN FOR THE RFACTOR COOLANT SYSTEM 10 AUXILIARY SPRAY ACTUATION CYCLES, CHANGE ".625 F" TO "Greater than

20 F but less than 625 F."

Basl s THIS CHANGE IS REQUIRED TO MAKE THE SPECS NORE ACCURATE. THE CYCLE IS FOR THE TEMPERATURE PANGE

-~20 F TO <625 F. ACTUATION BELOlj 20 F DOES NOT APPLY TO THIS CYCLIC LIMIT~

~M TABLE 5.7-1 0)+

COMPONENT CYCLIC OR TRANSIENT LIMITS CYCLIC OR DESIGN CYCLE COMPONENT TRANSIENT LIMIT OR TRANSIENT Reactor Coolant Sp'stem 200 heatup cycles at < 1004F/h Heatup cycle -T from . 200"T and 200 conldown cycles at to > 550'F.

100"F/h Cooldown cycle - T from

> 550 F to < 200 F. g 200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200'F/h. temperatures from > 650 F to

< 200'F.

200 loss of load cycles, without > 15K of RATED THERMAL POWER to immediate Turbine or Reactor trip. OX of RATED lHERMAL POWER.

40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical A. C. electrical power. ESF Electrical System.

80 cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump.

400 Reactor trip cycles. 100'o (C of RATED THERMAL POWER.

10 auxiliary spray actuation cycles. ~~

Spray water temperature differential pKc<4cn +h*~ 32o4F'~t less +>< 4~~

200 leak tests. Pressurized to > 2485 psig.

lO hydrostatic pressure tests. Pressurized to > 3107 psig.

Secondary Coolant System 1 steam line break. Break in a > 6-inch steam line.

10 hydrostatic pressure tests. Pressurized to > 1481 psig.

C',P8c.L Comxnents 8 NPP Proof and Review Technical Specifications Record Number:

LCO Number:

767 FIRE PROTECTION Comment Page Type:

Number: VARIOUS ~

IMPROVEMENT v~

Section Number: FIRE PROTECTION Comment:

DELETE THE FIRE PROTECTION SYSTEM SPECIFICATIONS PER THE ATTACHED MARKUPS.

Basis PER PREVIOUS CPE L LETTERS NLS-86-188 DATED JUNE 4, 1986 AND NLS-86-230 DATED JULY 22, 1986.

SHNPP REV)SlON I'Il&L UN i

6. 0 AOMINISTRAT I VE CONTROLS

6. 1 RESPONSIBILITY 6.1.1 The Plant General Manager shall be responsible for overall unit opera-tion and shall delegate in writing the succession to this responsibility dur-ing his absence.

6. 1.2 The Shift Foreman (or, during his absence from the control room, a

'designated individual) shall be responsible for the control room command func-tion. A management directive to this effect, signed by the Vice President-Harris Nuclear Project shall be reissued to all station personnel on an annual basis.

6. 2 ORGANIZATION OFF SITE 6.2. 1 The offsite organization for'nit management and technical support shall be as shown in Figure 6.2-1.

UNIT STAFF 6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1;

b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MOOE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room; C. An individual qualified as a Radiation Control Technician" shall be ~

on site when fuel is in the reactor;

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation; 4yld T

• the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

SHEARON HARRIS - UNIT 1 6-1

CP Bc.L Coxnmenta HNPP Px-oog and Review Technical S pecif ication Record Number: 772 Comment Type: ERROR LCO Number: 6.02.01 Page Number: 6-3 Section Number: FIGURE 6.2-1 Comment:

DELETE THE BLOCK FOR "MANAGER ENGINEERING AND CONSTRUCTION SERVICES". ALSO, REMOVE THE "s" FOR THE TITLE OF THE MANAGER FUEL"S" DEPARTMENT, Basis THESE CHANGES ARE TO CORRECT PHICAL ERR AND TO DELETE A POSITION WHICH NO LONGER STS WITHIN THE ORGANIZATION i~'~"'

(~, r.<W J

CORPORATE ORGANIZATION CIIAWRIAWPR(5 tel ND IMFf IKOIIIVC<If KfR f(IROP (IKCIIIIrf VKfPRf SRKNI SCCA YK( IR( SlaNf S(IAal VKfPRI SKCNf ISANAaR Car(SIAI(

I%%RAIIOIISSIP@M f IRK(CAR It(ICRAIIOI QIAII I r ASSIAt AtKC IIAIIAIÃRIRK( (AR SAI firI IIANAaRIICI/PIPARII5511 I ~ IIANAaROA SIRVKt S fIhMOPKWfAL %RYKC5 W(t PR(SKI NI CP(RAf CXIS IIANAaa Iultaw aIAlnr IRAROKt IIAIIAQR IRK(CAR t

~ I ANI CONS IRUC IIDW IIANAC(RIACl I AR IIANAaR CPCR A IKPIIOAt0(

IRAINPfS SI( IKPI VICC PRCSICCNT IAKI(AR tl404(RPKt AIKILKIHSRKt IIANArCR ICX(fAR 5 IAII SaP(51 f OFFSIT E

~O~O~O~O~O ~ OO ~ O ~ O ~ O ~ OOO ~ ~ O ~ O ~ O ~ ~ O ~ OO ~ OO ~ OO ~ O ~ OO ~ O ~ OOO ~ O ~ O ~ O ~ ~ OO ~ O ~ O OHQTE OO((IS OA4C RPICC I(% VKC PR(SCKN1 aolf 5 ICKLCAR 5Alffr 1 IWSIIS IPAPINILPRf NMRIS IRK(CAR PROXCf IIARRIS ILANf PL ANI aIK5 % ISAIIAaR alKR AL ISAIIAaR CIKt0% (R IWIAaR A(POOSIRA 1 KPI alKPAL Ituslaa IO t 5laK cow(f IKw f Mle

~ " ~~ ~ "- LRCSa CCPPCPNCAIKPI IIANAaRPLAN%+ 5 AaPNrf IPAIIV/ aH'tAIR(AIPW CIWIRIR

~

tt

,c $

f 6.2"1 'IGURE OFF SITE ORGANIZATION

CPScL Camment.a RHNP'P' fnN3. g}t ~44 VREHRX.&&X 5pBG+4+gQQ+g p~

Recur d Number: 791 Cr}m%~e>>t. Tyoe: ERRQR LCQ hlu>>~L er: 6. 02. 0 i Paae Number: 6-3 Ser l i c>>'i hfu(Aber: i F GUPE 6. 2-1 Cei>>r>>eu L:

Cl<<:ANB'HE F:BUR~ PER THE ATTACHED.

&a.,6l. 5

Hi S f1AR!'UP REFLLCTS RECENTLY AhlNQUflCED CHA! lBES i h!

THE CORPORATE SfRUCTUfiE QF CP~L.

Jg, t

"~P '" ~..', dh'4 3C VP 9(

ir ~k

CORPORATE ORGAN1 ZAT ION OQ OIIAIStttfCCIII Ate ISICS I VICLIIIVCOIKIA

'.KIOCO lICCltfIVC VICf ttCSCKIII Itt 5tlC% VKf SCS III SIIACO VKI AAISCXNf ~ IAVACIA CIOIIOAIt IVCNAIKWS SIPVVIAI ~ AtttAAUIOAAICN OIAAIIIASIAtiA~I IMIIACCAILKLCAA SAtttVS IIAIIA4C~ OA Sl AVKI 5 fIIVNOttldAL SISVICC 5 IWQN5 CIIIS VKC ttt ftllIIMSAICWS IIAAIACtatulfaimOAAIIIV

)

IAAMN'Af CWW A IIAIIAQ0 IAKLCAO

~ I AllI COCII AIKI CW IIAIIACCOIS tlI M IMAIACI1 OCTA I ttlfOAAX IAAMICitlC IIOI IIIAiiAfSA Q&ALLAIL VKC ttfSCC IIIIAKIIAO

- L Se~ft>> CMAECAVC'iAICI ItINltti IIAIIAI4%

ISKLCAO SI Aft SOVOII CFF SIN

~ 11 ~ 10 ~ tttt ~ Ott ~ 0 ~ ~ 0000 ~ 011 ~ 0 ~ ~ 0 ~ 000140 ~ ttttt ~ 04 ~ 10 ~ 1 ~~

VICC ttlSKKIIf IeflfAASAffff WC'ttlff IIAIAIS IAAIQQLSOf ILAKSISISKLt AA IOOKC I SLAHI QIKSALIIAIIAStt SCION AL tIUCACCO C INC tt t%

IIAILASI0 Aft%0% IO 5 IIOI rtmALIINIaua IILI 5IOK COSTI IKII

~ "- -

~ LOOSOI CO+LAOCAIIOI I IAIIASC0tLANCKig AOttSIAAIIVCCOOAISIAISVI CIVIIAlt FIGURE 6.2-1 OFF SITE ORGANI EATIOH

Shearon Harris Technical Specifications Re<olution of Staff Comments crt gt astor: tcL$ &c IIc m Comment Cate: g/g/tti Comment:

1g'e find the Qh-related material in Section 6.0 acceptable except, that.

the organizational positions for the positions of the blanager gA Services and Manager Quality Check in Technical Specifications figure 6.2-1 are not. shown in FSAR Figures 17. 2, 1-1 or 17. 2. 1-2. Also, these positions are not described in FSAR Section 17.2, "Quality Assurance During The Operations Phase." Consequently, in order to assure compatability between the technical specifications and FSAR Section D.2, the posit,ion descriptions should be indicated in the appropriate figures and accurately described in t,he FSAR.

. Resolution Figure G.2-1 of Tcc'h Spec should be revised to delete the 1.

Basis The Manager ttnatecy Check does noc perform any functions or have re-sponsibilities requfrsd by the QA Manager'uality Check, Program.

2. The Manager QA Services is shown dn PSAR Figures 17.2.1-1 .and 2. The Tech Spec Figure 6,2-1 shows the 17.2.1-2 and is described on page QA organfaatfon at license,fssue 17.2.1-6, first paragraph, Me, (Tach Spec affective date) which assume your comment mean't.to depicts the shift of the Manager address the Manager Material QA/QC Harris Plant to Yunager Quality. Chapter 17.2 wi11 be re- Materfel Quality. Chaprer 1?.2 vised after lfcense issue'o describes the QA organfaatfon at reflect the Manager Material present with the Manager QA/QC Barris Quality. (Note: The Hanager Material Quality is currently functioning as the Manager .QP,/QC contfnue until license i.ssue). The'er Plant functions (these functions will Manager Material Qual"'ty's functions Harris Plant as described .in FSAR wQ.1 start Bt licanse issue and Chapter Chapter 17.2.) 17.2 will be revised at that time to reflect these new functions.

/

Resolution cce e:

CP!t Date: Gate:

SHNPP CPBc.L Proof and Review Technical Coxnxnenta S

Ol( ~

pecifications Record Number: 759 Comment Type: ERROR LCO Number: 6.02.02 Page Number: 6-4 Section Number: FIGURE 6.2-2 Comment:

DELETE THE BLOCK FOR THE ADMINISTRATIVE SUPERVISOR WHICH REPORTS TO THE DIRECTOR PROGRAMS PROCEDURES.

Basis THIS POSITION NO LONGER EXISTS WITHIN THE SHNPP ORGANIZATION.

PLANT OAGANlZATION It'll

' h AMI t AR

~SutlfreSSIIIC

~ SR4SAAIO A MOCfHJtf S ASSIS IAHI ftAttt Cl IC 1 AL ttAIIAfAR Of SIR A I Otf EEHA t~f IMIIMil%

ETOf%00%liflt A ttttlCCR ISARACCR NAWAIIOIGWNN 0%R ATIOIS TEOSRCAI. QATAR f SIVS EIOIOO0%IIIAL4 ttlWIE~ QKRAIIOIS ClCOftRRO CKIRSNT SLW RYl%% %PCS VISOR QPCRYIXR SISCRvISOI Et EC litKAL ENACT KCCIALIST

$ &fCMttlltl EIIVOORCIIIAL L OCISSNV

~ %0%Cf &CCIALIST RAOANW COIN%

RAOATCHICOIN0.

SIPERVISOI AJE IAAYOfCRAIOIS LEGS'OMWS NATIVEOIGAIRIAIIOI

~ """" LOKSOf COSSMCATIOI

~ SEIROI % AE t Ot EICRAT OIS UCEIISE W. SEAEIOIOCRAIOtSLICE~

FIGURE 6.2-2 UNIT ORGANIEATIOH

C:PSci Cummen<m SHNPP Final Dr a f0 Technical Speci ficat j.one I.:r;r'r <f t lumber: 8'.> 1 Comrnr-; n t, 'T yf r o: EtiROf':

LL:O Nurrrhvr: . b. <) '. r.r2 Paqr'lubber". 6- 0 Sec, t. ] err r tdurrrbr r: F I GURE 6-CQrlmN 'l r l lI ON THE F 1 BURE FOR THE PLAllT QRBAt ZAT I ON - "DD Tl-lE TI1LF "f-LANS AND f'RQBfiANS" TO THE TABLE.

Bc.ES1 5 THIS CHANGE IS PURELY AN ADt f I NI STRATI VE CHANGF. IN THAl'ftE 1 I TLE OF THE POSITION OF "PLAN" AND PRUBRAl'fS" I S A NEW POSI TION CREATED WITHIN THE PLANT Of"BANIZAT ION. THE f. SAR IS CURRENTLY BEING REV I SEl;; ANi) WlLL SllOlf THI S NEW POSITION.

~Vj c s>t~

ggp

~ 0

~ ~ ~ '

~

~ ~ Sl

~ 1' ~ 1 '

~

I ~ ~I '

O1 1~

~ I 1

~ '

"~

' ~ 5~

~ I .~

~ 1' C

I ~ ~ '

~0 ' ~ '1' ~

~ ~I ~

~

'I' ~

C C I ~~

~ 'it ~

~

~ I 1

~

~

SHNPP FShR Physics and Nuclear Safety policiesi He is responsible forthe personal revisv of the training <<nd qualification requirements of the following managers who report directly to himp'anager " Operations, Hinager-Haintsnanca, Hanager - Environmental and Radiation Control, and Hansgar 27 Technical Support. ri ns, thr A~ent-

~ responsrbi'li~s". The hssiatant Plant Ginaril Hanager reports directly to the Plant General Hanager.

13,1,2.2,3 Plant Programs and Proceduraa Unit The Plant Programs ind Procaduzea Unit provides support functions such as security, procedure control, and emergency preparednessi The Director Plant Programs and Procedurea provides direct support to the Plant Cenera Hanagar in the ireas of security, emergency preparedness, procedure development and control~ personnel administration and plant administrative coordination) directs plant security planning and activities) directs emergency praparadnasa planning and activt,tiaa at the plant staff 1avdl aupervi pea the preparation, review, approvil and dist ribut ion of plan't procadurea and directives. He is assisted in thpae ggjep u~~~Wecurity Buperviaor~4%8 a HRn or pat(i%1st ~ ~~s

--~Pn Emergency Preparedness, The Director~plant Programs and Procedures reports to the Plant General Hanagar - Harris Fiant The

~M<r ~Ad - Ffa.~ W P~ ru,m5 ~'+>

the administrative functions of the plant including incoming correspondence screening and action assignmant; action item/response development and fol o up) outgoing ~op ogd nce preparation, screening and coordination) pzocadure preparation, review, and approve+

io

~7 The Security Supervisor develops> implements, and maintains a sacuri.ty program which ensures that the security of the plant i ~ maintained in accordance with NRC requirements< He maintains a close working relationship with Local law enforcement agencies to ensure coaplianca with MRC regulitions. He provides input to the Training Unit so that employees requiring access to tha plant are proparly trained and hedged. He ansuraa that equipment and guards are availsbla and in a state of readinesa. The Senior Specialist - Security is assisted by Technical Aidaa and a contract security 'guard forca. The Security Suparvisor reports to the Director Plant Programs and Procedures The Senior Specialist - Emergency Preparedness ia responsible for the continuing refine>ant of tha plant Emergency Pieparedness Program which ensures that a "state of raadineas" ia maintained at the plant to copa with any classification of emergancy, He incorporates the provisions of the plant Emergency Plan in ths program and revises the program and related procedures as chsngas are made in the plant Emergency Plan. Be coordinates the training of Technical Support Center participants and the annual Emergency Drilli The Sanior Specialist - Emergency Preparedness reports to tha Director Plant 27 Programs and Procedures.

CP RL Coxnments

~ RNPP Proof and Review Technical Specifications

(

Record Number: 744 Comment Type: ERROR LCO Number: 6. 02'. 03. 01 Page Number: 6-6 Section Number: 6.2.3.1 Comment:

INSERT IN THE SECOND LINE AFTER "industry advisories" THE FOLLOWING'WORDING ~'(including information forwarded from INFO from their ev'aluation of all industry LER's),

Basis SEE ITEM 743 THIS CHANGE IS NEEDED TO ACCURATELY REFLECT THE EXACT ORGANIZATION THAT PERFORMS THE VARIOUS REVIEWS'LL ITEMS MENTIONED IN THE FINAL DRAFT ARE STILL COVERED, BUT HAVE BEEN MOVED TO THEIR PROPER PLACE.

",;., FINAL DRAI AOMINISTPATIVE CONTROLS

6. 2. 3 ONSITE NUCLEAR SAFETY ONS UNIT FUNCTION (lduubi4D W~~g~41'DRuaRDE'D iW~ ~~ rRa~ P'u4 SNRuar y oS AC DSRfnay gag'C 6.2.3. 1 The ONS Unit shall function to examine unit operating characteristics, NRC issuances, industry advisories, and other sources of unit design and operat-ing experience information, including units of similar design, which may indi-cate areas for improving unit safety. The OHS Unit shall make detailed recom-mendations for revised procedures, equipment modifications, maintenance activ-ities, operations activities, or other means of improving unit safety, to appro-priate levels of management, up to and including the Senior Vice President-Operations Support, if necessary.

COMPOSITION 6.2.3.2 The ONS Unit shall be composed'of at least five, dedicated, full-time engineers located on site. Each shall have a baccalaureate degree in engineer-ing or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ONS Unit shall be responsible for maintaining surveillance"of unit activities to provide independent verification" that these activities are performed correctly and that human errors are reduced as much as practical.

RECORDS 6.2.3.4 Records of activities performed by the ONS Unit shall be prepared, maintained, and forwarded each calendar month to the Hanager-Nuclear Safety and Environmental Services, 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4. 1 The Shift Technical Advisor shall provide advisory technical support to the Shift Foreman in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. The Shift Technical Advisor shall have a baccalaureate degree or equivalent in a scien-tific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and,layout, including the capabilities of instrumentation and controls in the control room.

6. 3 UNIT STAFF UALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of the September 1979 draft of ANS 3.1, with the exceptions and alter-natives noted on FSAR pages 1.8-8 (Am.20), 1.8-9 (Am.17), 1.8-10 (Am.22),

"Not responsible for sign-off function.

SHEARON HARRIS - UN'IT 1 6-6

CP8cL Cummen<~

SHNPP Final Draft Technical Bpeci+icatians

t. r ~ ~ r V f 'l (5 i ~ ~ 1 A) Cornrnerrl: Typr".

L Q t !r. rrrh r.rr  : P=rqe t lumber: 6-E 8r ;vii

""r: "5 i csr! tl~~<<tL" -'r: Zc I!'JDE X i '.>nrlnQ'r DE'T". TH" SEC: IQtl a..~. I AND t'!ARK THE SECTIQt! AB "DE!. ETED" .

ALBQ

'nUAr QW PAGE .",

IFIC'ATIQtlS" TV "D=LE vi i. i CH~Nt IBE "UNIT STAFF

'E IHFQR?!ATION I tl THI S PARABRA."H IS COVERED I tl DETAIL IN 'r+ . FSAR. AS t'1!,rC', t'1QRE t;!JS ii!.-r'z r:EgLrIRt .-

'r'- ".t'JBrE ) FQR Pt 'lPrrt v ADt r a EQLEtlr TECH S=rEC pg r ~H VE RE SQ

~

.-ICATIQ!i IS HEI!!B DELETED Itd TH" FQR:HCQt'!..hB

~

""..F &>.QUK TECHNICAL BF'E" IFI CA"r'IQ!JB. THIS CHAI'!BE HAB

-RE'i>IQJSLY DISCUSSED !~JITH tlRR STAFF.

~tt t The applicant proposes to delete Specification 6.3, Staff gualification. The staff finds this proposal acceptable because the staff's Safety Evaluation includes finding of acceptable criteria to be used by the applicant and because changes to these criteria~under the provisions of 10 CFR 50.59>will afford an adequate opportunity for review by the staff.

ADNINJSTPATIVE CONTROLS 6.2.3 ONSI NUCLEAR SAFETY (ON5 UNIT FUNCTION (IAICLUEs/AJ4 /Al~gjf)rp~+ pgg~pgE9E ES pg ~pg >CIR Fyitcug+g r>RDucmr ZeW' Og AS.

6.2.3.1 The ONS Unit shall function to examine unit operating characteristics, NRC issuances, industry advisories, and other sources of unit design and operat-ing experience information, including units of similar design, which may indi-cate areas for improving unit safety. The ONS Unit shall make detailed recon-mendations for revised procedures, equipment modifications, maintenance activ-ities, operations activities, or other means of improving unit safety, to appro-priate levels of management, up to and including the 5enior Vice President,-

Operations 5upport, if necessary.

COHPOS I 7 ION

6. 2. 3. 2 The ONS Unit shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a baccalaureate degree in engineer-ing or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ONS Unit shall be responsible for maintaining surveillance of unit activities to provide independent verification" that .these activities are performed correctly and that human errors are reduced as much as practical.

RECORDS

6. 2. 3. 4 Records of activities performed by the ONS Unit shall be prepared, maintained, and forwarded each calendar month to the Hanager-Nuclear Safety and Environmental Services.

6. 2.4 SHIFT TECHNICAL ADVISOR

'.2.4.1 The Shift Technical Advisor shall provide advisory technical 'support to the Shift Foreman in the areas of thermal hydraulics, re-ctor engineering, and plant analysis with regard to the safe operation of the unit. The Shift Technical Advisor shall have a baccalaureate degree or equivalent in a scien-tific or engineering discipline and shall have received specific training in in the response and analysis of the unit for transients and accidents, and unit design and layout, including the capabilities of instrumentation and controls in the control room.

DELGTEp 6.3 6..1 E member f themhi staff s+11 et or exceed the minimal qual~i<<

t'ons o the'ep aher f979 astro Rhs .1, with the gaeeptsoris Rrio astgr-ative oted o SAR pages .8 4 (Am.2Q , $ ..8"9 (Am.lQ -1;8-10 (A'm.22).

"Not respons ib1 e for s i gn-of f function.

SHEARON HARRIS - UN'IT 1 6-6

SHNPP RFViS~OM A06 t986 FINAL IjRi ADMINISTRATIVE CONTROLS UNIT/STAFF '." IFICATIONS ICnnninn U

1. 11 (Am.20), 1.8-12 (Am.17), a'nd 1.8-13 ( m.17), for mparable p sitio s, except fo tne Manager-Environme'ntal and Radiation Contr 1 who shal meet or e ceed t e qualifications of Regulatory Gujde 1. 8, Sep ember 1975. g The censed perato s and Senior Operators shall also/meet or exc ed the miniybm qu ifica-tions of the supplemental requirements specified in ections A a d C o Enclo-

, sure 1 of the March 28, 1980, NRC letter to all licpsees.

6. 4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Director"Harris Training Unit and shall meet or exceed the requirements and recommendations of the September 1979 draft of ANS 3. 1, with the exceptions and alternatives noted on FSAR pages l. 8-8 (Am.20), 1.8-9 (Am.17), 1.8-10 (Am.22), 1.8-11 (Am.20), 1.8-12 (Am.17), and

1. 8-13 (Am. 17), and Appendix A of 10 CFR Part 55 and the supplemental require-ments speci fied in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

6. 5 REVIEW AND AUDIT

6. 5.1 SAFETY AND TECHNICAL REVIEWS 6.5. 1. 1 General Pro ram Control 6.5. 1. 1. 1 A safety and a technical evaluation shall be. prepared for each of the following:

a. All procedures and programs required by Specification 6. 8, other procedures that affect nuclear safety, and changes thereto;

b. All proposed tests and experiments that are not described in the Final Safety Analysis Report; and

c. All proposed changes or modifications to plant systems or equipment that affect nuclear safety.

6.5. 1.2 Technical Evaluations 6.5. 1.2. 1 Technical evaluations will be performed by personnel qualified in the subject matter and will determine the technical adequacy and accuracy of the proposed activity. If interdiscipl'inary evaluations are required to cover the technical-scope of an activity, they will be'erformed.

6. 5 1. 2. 2 Technical review personnel will be identified by the responsible

~

Manager or his designee for a specific activity when the review process begins.

6.5. 1.3 uglified Safet Reviewers

6. 5. 1, 3. 1 The Plant General Manager shall designate those individuals who will be responsible for performing safety reviews described in Specification 6.5. 1.4.

SHEARON HARRIS - UNIT 1 6-7

C.Pal C:nmrne.num SHNPP Final Da-a+%. Taahnic=al Sp c=,i+irat Reco( d Numbe(': 806 Comment Tvoe: ERROR LCO Nu(nber: '.4 Paoe Number: 6-3 5~7 Section Number: 6. 4 5 FIG 6-2. 1 Co(ament:

IN &0TH THE FIGUPE AND IN SECTION 6.4 CHANGE THE TITLE "DIRECTOR HARRIS TRAINING UNIT" TO "MANAGER HARRIS TRAINING UNIT" ON PAGES 6-6 AND 6-7 CHANGE THE REFERENCE FSAR AMENDMENTS TO THE FOLLOWING:

PAGE 1.8-9 (AM. 26)

PAGES 1.8-lo. 1 1, 12. AND 1: (AM. 27)

E(gal s CORPORATE MANAGEMENT HAS CHANGED THE TITLE OF THIS POSITION NEITHER THE'ERSON HOLDING THE POSITION

~

OR THE DUTIES OF THE POSITION HAVE CHANGED.

THE CHANGES TO THE FSAR AMENDMENT NUME(ERS IS TO MAKE THE TECH SPECS CONSISTENT WITH THE LATEST FSAR CHANGES.

CORPORATE ORGANI 2AT ION OIAIRtIAH/PAESICEMAt4 CHIEF G(ECLITIYE Of FICER SEMOR EXECUTIVE ViCE PRESIOENT SENIOR VICE PRESIDENT SEHICR YICE PAESIOENT t1ANAGER CORPORATE OPERATIONS SUPPORT tAlCLEAR GENERATION QUALITY ASSLAAIiCE G

~

CA I1ANAGER NUCLEAR SAFETV L t1AHAGER HJCLEAR OA SERYiCES VICE PRESIDENT OPERATIONS EHYI~HTALSER YICES TRAIHIWT TECH SUPPOR't PLANt CONST RUCT ION 0

~~e t1AHAGER NUCLEAR tlANAGER GAIA lERI AL OJALI'IV FLED SECTION VICE PRESICENT NUCLEAR 00 EMISEERIt4i ANDLICENSIHG tTT C l ~/AF88Zg'1ANAGER t1AHAGER M/LEAR TRAIHII4IlSECTIM tlAHAGER t1AHAGER CPERATKt6 O<1OC hl NUCLEAR STAFF SUPPORT OFF SITE

~ ~ 1 ~ ~~~

ONSITE 11111 ~~1 ~ ~ 11 ~ 1 ~ 0 ~ 111 ~1~ ~ 1~ ~ ~~ 11 ~ 111 ~ 111111 ' ' ~ 111 ~ 1~ 11 ~ ~ ~ ~ 11 ~ ~

O OIRECTOR WRE8%0t. VICE PRESIOENT DIRECTOR OAlOC-ONSITE HUCLEAR SAFETY HARRIS TRAINIHG ellr HARRIS QKLEAR PROJECT HARRIS PLANT

~ A~

PLANT GEHERK tlANAGER GEHERALtlAHAGER

~~~~~ ~ ~A~~ ~S ~ ~ ~ AO ENGINEER@6 tlAHAGER ADI1IHISTRATION GEHERKt1AHAGER t1lLESTCfK COtPLETIOH LEGEND

-. ~ - ~ ~ - ~ - ~ ~ LlhKS OF Ef5t1PICAT IOH tIANAGER PLANNING 4 AGIIIIIITTAATTIT OIGAIIITATATI CONT ACL

I'

~ ' I' ~

~ ~ ~

7 ~

~ ~ I ~)

~ ')' ~ iI ~

I ~ I ' I I 1'

I '

C'

~ ~ I I '

0 l.INAL Utgt ADtlINISTPATIVE CONTROLS 6.2.3 ONS17E NUCLEAR SAFETY (OHS UNIT FOIICTIOR 6.2.3. 1 The ONS Unit shall function to examine unit operating characteristic, HRC issuances, industry advisories, and other sources of unit design and operat-ing experience information, including units of similar design, which may indi-cate areas for improving unit safety. The ONS Unit shall make detailed recom-mendations for revised procedures, equipment modifications, maintenance activ ities, operations activities, or other means of improving unit safety, to atIpro priate levels of management, up to and including the Senior Vice President..

Operations Support, if necessary.

CONPOSITIOH 6.2.3.2 The OHS Unit shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a baccalaureate degree in engineer" ing or related science and at least 2 years professioaal level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPOHSIBI LITIES 6.2.3.3 The ONS Unit shall be responsible for maintaining surveillance nf unit.

activities to provide independent verification" that these activities are performed correctly and that human errors are reduced as much as practical.

RfCOROS 6.2.3.4 Records of activities performed by the ONS Unit shall be prepared, maintained, and forwarded each calendar month to the Hanager"Nuclear Safety and Environmental Services.

6. 2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Foreman in the areas of thermal hydraulics, reactor engine<<ing and plant analysis with regard to the safe operation of the unit. The Sh>ft Technical Advisor shall have a baccalaureate degree or equivalent in a scien" tific or engineering discipline and shall have received specific training in the response and analysis of the unit for. transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6. 3 UNIT STAFF UALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qu>>ifica tions of the September 1979 draft of ANS 3.1, with the exceptions and alter natives noted on FSAR pages 1.8-8 (Am.20), 1.8-9 (Am.H), 1.8-10 (Am.+)R 4/

"Hot responsible for s i gn-of f function.

SHEARON HARRIS UNIT 1 6-6

ADLAI NI STRATI VE CONTROLS UNIT STAFF UALIFICATIONS (Continued

-'7 47 p7 1.8-11 {Am.N), 1.8-12 (Am.X), and 1.8-13 (Am.M), for comparable positions, except for the Hanager-Environmental and Radiation Control who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifica-tions of the supplemental requirements specified in Sections A and C of Enclo.

sure 1 of the March 28, 1980, NRC letter to all licensees.

6. 4 TRAINING rrgggcE'4 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the -Harris Training Unit and shall meet or exceed the requirements and recommendations of the September 1979 draft of ANS 3.1, with the yxceptions and alternatives noted on FSAR pages 1.8-8 (Am.20), 1.8-9 (AmQ@, 1.8-10 (Am.~, 1.8-11 (Am ), 1.8-12 (Am.X(, and 1.8-13 (Am4$ ), and Appendix A of 10 CFR Part 55 and the supplemental require-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall includ familiarization with relevant industry operational experience.

a7 6.5 REYIEM AND AUDIT 6.5.1 SAFETY AND TECHNICAL REYIEMS 6.5.1.1 General Pro ram Control 6.5.1.1.1 A safety and a technical evaluation shall be prepared for each of the fol lo~ing:

a. All procedures and programs required by Specification 6.8, other procedures that affect nuclear safety, and changes thereto;

b. All proposed tests and experiments that are not described in the Final Safety Analysis Report; and

c. All proposed changes or modifications to plant systems or equipment that affect nuclear safety.

6.5. 1.2 Technical Evaluations 6.5. 1.2. 1 Technical evaluations will be performed by personnel qualified in the subject matter and will determine the technical adequacy and accuracy of the proposed activity. If interdiscipl'inary evaluations are required to cove~

the technical 'scope of an activity, they will be performed.

6.5. 1.2.2 Technical review personnel will be identified by the responsible Hanager or his designee for a specific activity when the review process begins.

6.5.1.3 uglified Safet Reviewers 6.5.1.3.1 The Plant General Hanager shall designate those individuals who will be responsible for performing safety reviews described in Specification 6.5.1.4 SHEARON HARRIS

- UNIT 1 6-7

SHhPP FSAR Regulatory Guide 1.8 PERSONNEL SELECTION AND TRAINING (REVISION 2, FEBRUARY 1979 DRAFT)

SHNPP will comply with the requirements of ANSI/ANS 3.1; September 1979 Draft, with the alternatives listed herein. It is understood that the NRC has not endorsed this Standard, but when the SHNPP applied for its operating license, the September 1979 Draft was current. Because this standard was the existing 20 guidance at the time of our operating license application, CP6L believes it is (

acceptable to use the draft Standard as the basis for selecting and training SHNPP personnel. The Company has received approval from NRC to follow the September 1979 Draft without further revisions. 20 a) Paragraph 2 defines the terms of the Standard. As stated in SHNPP FSAR Section 1.8, paragraph 1.74, CP&L has combined'he definitions given in various ANSI standards, in order to provide an available reference source.

The definitions in Section 1.8, paragraph 1.74 agree with ANSI/ANS 3.1, September 1979 Draft with the following exception:

When the phrase "Bachelor's Degree or Equivalent" is used, the qualifications considered as minimal acceptable substitutes for a Bachelor's Degree are a high school diploma or its equivalent and one of the following:

1) Four years of formal schooling in science or engineering;

2) Four years applied experience at a nuclear facility in the area for which qualification is sought;

3) Four years of operational or technical experience or education or training in nuclear power; or li

4) Any combination of the above totaling four years.

b) Table 1.8-1 cross references the "Functional Level and Assignment of Responsibility" definitions found in Section 3 of the Standard with the positions/titles of the SHNPP organisation and the "Qualifications" found in Section 4 of the Standard. The numbers enclosed in parentheses denote the specific exceptions or proposed alternatives to the Standard's requirements which are described in paragraph (c) below.')

Exceptions or proposed alternatives:

1) Paragraph 4.3.1 describes the qualifications for supervisors requiring NRC licenses. This paragraph requires that one year of nuclear power plant experience shall be at the plant where the supervisor is licensed, unless such experience is acquired on a similar (same NSSS) unit. CPM shall alternatively provide the qualifications prescribed by 10CFR55 and the NRC letter dated March 28, 1980, which is titled "Qualifications of Reactor Operators". The qualifications cited in these two references shall be applicable to individuals employed as Operating Supervisor and Shift Foreman.

1.8-8 Amendment No. 20

SHNPP FSAR 2)) P aragraph 4.3.2 describes the qualifications for supervisors who are not required to hold an NRC license, but who are associated with "systems, equipment, or procedures involved in meeting the Limiting Conditions for Operation, which are identified in Technical Specifications". CP&L does not feel plant safety will be enhanced b y r equiring these supervisors to perform their duties under direct on-site supervision for a minimum of six months. Instead CP&L propose s t 0 s 1 ect qualified individuals for these positions based upon past se performance and experience.

3) Paragraph 4.5.1.1 describes the requirements for non-licensed operators. CP&L does not feel plant safety will be enhanced b requiring non-licensed operators to have one year power plant experience. CP&L shall alternatively provide a training/qualification program commensurate to the functions and responsibilities these employees will perform.

4)) Paragraph 4.5.1.2 describes the requirements for licensed operators. CP&L takes exception to these requirements. Prior to operating the facility, licensed operators shall be qualified in accordance to IOCFR55 and the NRC letter dated March 28, 1980, "Qualification of Reactor Operators".

5) Paragraphs 4.5.2 and 4.5.3 describe the qualifications for technicians and maintenance personnel. CP&L considers these technicians and maintenance employees to be "in training or apprentice positions",

as described in paragraph 3.2.4. Therefore, CP&L shall comply with the requirements as stated in paragraph 3.2.4.

6) Members of the QA staff wi ll be trained and qualified in accordance with Regulatory Guide 1.58, which endorses ANSI 45.2.6. The SHNPP position on Regulatory Guide 1.58 addresses the SHNPP position.: relative 26 to ANSI N45.2.6.

7) Various CP&L positions are not addressed in the Standard.

Therefoxe, CP&L lists these, positions in Table 1.8-1 for reference, and CP&L will prescribe the training, responsibilities, and qualifications commensurate to the job requirements.

8) The ALARA Specialist shall have a BS Degree or the equivalent and two years experience, one of which shall be nuclear power plant experience, or the employee shall have an advanced degree and one year nuclear power plant experience.

9) The Project Engineer - On-Site Nuclear Safety shal.l have a BS Degree in Engineering or the equivalent and shall have a minimum of four years experience. These qualifications are required prior to preoperational testing or at position appointment, whichever is later.

n n

S! L'HAPP FSAR

10) The positions specified in Table 1.8-1. shall have a BS Degree in Engineering or the equivalent and two years experience, one of which shall be nuclear power plant experience, or the employees shall have an advanced degree and one year nuclear power plant experience. These qualifications are required at initial core loading or at position appointment, whichever is later.

11) The Training Specialist shall have at least four years power plant experience, two of which shall be nuclear power plant experience.

Individuals in this position shall demonstrate their competence by having held an SRO license or by having trained at the SRO level prior to teaching NSSS, integrated response, transient analysis, or simulator courses. These qualifications are required at initial core loading or at position appointment, whichever is later.

12) The Director On-Site Nuclear Safety and the Principal Engineer-On-Site Nuclear Safety shall have a BS degree in Engineering or the equivalent and shall have a minimum of six years experience. These 27 qualifications are required prior to preoperational testing or at position appointment, whichever is later.

d) Paragraphs 4.7.1 and 4.7.2 describe the qualifications for independent review personnel. Standard Technical Specifications also address the personnel requirements for individuals functioning in this capacity, and alternatively, CP&L shall comply with STS requirements for independent review personnel.

e) Paragraph 5.2 outlines an acceptable training program for personnel to be licensed by the NRC. However, CP&L feels this portion of the Standard is unnecessarily prescriptive. CP&L will provide a training program as described in FSAR Section 13.2 for licensed operators and senior operators, which will comply with the intent of the standard, requirements in 10CFR55, and the NRC letter dated March 28, 1980, "Qualifications of Reactor Operators".

Paragraph 5.5.1 outlines the retraining program for licensed personnel.

10CFR55 requires a requalification program to be submitted and approved to meet Appendix A, 10CFR55. CP&L proposes to requalify licensed personnel in accordance to the NRC approved requalification program outlined in Appendix A, 10CFR55. In addition, CP&L will comply to the NRC letter dated March 28, 1980, "Qualifications of Reactor Operators" and the intent of paragraph 5.5.1.

f) Paragraph 5.5.2.3 describes requirements to maintain certain documents. In order to provide consistency in the Document Control program, CP&L shall retain and maintain documents as required by ANSI N45.2.9-1974.

g) Paragraph 1, Scope, states in part, "this standard is further limited to personnel within the owner organization." However, paragraph 5.4 refers to temporary maintenance and service personnel. CP&L will apply the requirements of ANS 3.1, September 1979 to only those personnel directly employed by CP&L>

and only the training of paragraph 5.4 will be required to be given to temporary maintenance and service personnel.

h) Positions shown on the SHNPP organization chart that have not been described herein shall be filled by individuals, who by virtue of training and experience, have been deemed qualified to fill these positions.

1.8-10 Amendment No. 27

SlL'HAPP FSAR III TABLE 1.8-1 FUNCTIONAL LEVEL,'ASSIGNMENT OF RESPONSIBILITY~

AND QUALIFICATIONS CROSS REFERENCE FOR SHNPP ANS 3el SHNPP Title Seccian

~Mene ece 4.2.1 Plant General Manager 1 4.2,1 Assistant Plant General Hanager 27 4.3.2 Director Plant Programs and Procedures, 4.2.4 Manager Technical Support 4.2.4 Manager - Start Up 4.2.3 Manager Maintenance e 4.2.2 Hanager Operations 4.4.4 Hanager Environmental and Radiation Control 4.3.2 Director Regulatory Compliance Technical Su ort 4.6.1 Manager - Harris Plant Engineering Section 27 (Refer to FSAR Section 13.1.1.2) 4.6.2 (10) Shift Technical Advisor 4.6.2 (8) ALARA Specialist 4.6.2 (10) Engineer Supervisor - Nuclear 4.6.2 (10) Operations Support Supexvisor 4.6.2 (10) Principal Engineer (Support) 4.6.2 (10) Project Engineer NSSS 4.6.2 (10) Project Engineer Equipment Evaluation 4.6.2 (10) Project Engineer BOP 4.6.2 (10) Project Engineer Engr. Specs.

4 '.2 (10) Project Engineer - ISI 4.6.2 (10) Project Engineer Performance/Reliability 4.6.2 (10) Project Engineer " Maintenance 4.6.2 (10) Pxoject Specialist << RadMaste 4.6.2 (10) Project Specialist Radiation Control 4.6.2 (10) Project Specialist Environmental and Chemistry 4.6.2 (9) Project Engineer On-Site Nuclear Safety 4.6.2 Engineering Subunit 4 '.2 Specialist Subunit 4 '.2 (12) Principal Engineer - On-Site Nuclear Safety Professional Technical 4.4.1 Senior Engineer - Reactor 4.4.4 Radiation Contxol Supervisor 4.4.3 Chemistry and Environmental Supervisor 4.4.2 Maintenance Supervisor - Electrical 4.4.6 Start-Up Supervisor

( ) denotes. number of exceptions or alternatives proposed in paragraph c above.

1.8"11 Amendment No. 27

SlL'HAPP FSAR TABLE 1.8-1 (cont'd)

Professional 4.4.6 Start-Up Engineers

'.4.7 Director - Training 4.4.5 Director QA/QC 4.6.2 (12) Director - On-Site Nuclear Safety Foremen 4.3.1 (1) Operations Supervisor 4.3.1 (1) Shift Foreman 4.3.2 4.3.2 Administrative Supervisor 4.3.2 Security Supervisor 4.3.2 (2) Senior Specialist Fire Protection 4.3.2 (2) Maintenance Supervisor - Mechanical 4.3.2 (2) 16C Foreman 4.3.2 (2) Electrical Foreman 4.3.2 (2) Mechanical Foreman 4.3.2 (2) Painter and Pipe Coverer Foreman 4.3.2 (2) Radwaste Supervisor 4.3.2 (2) Radvaste Shift Foreman 4.3.2 (2) Environmental and Chemistry Foreman 4.3.2 (2) Radiation Control Foreman 4.3.2 (2) Traveling Radiation Control Foreman 4.3.2 Project Engineer Computer 4.3.2 Senior Specialist - Emergency Preparedness 4.3.2 (6) Specialist - QA 4.3.2 (11) Specialist - Training Operators Technicians-Maintenance Personnel 4.5.2 Technician I - Engineering 4.5.2 Technician I - Radiation Control 4.5.2 (5) Technician II-- Radiation Control 4.5.2 Technician I Environmental and Chemistry 4.5.2 (5) Technician II-- Environmental and Chemistry 4.5.2 Technician I Traveling Radiation 4.5.2 (5) Technician,II - Traveling Radiation 4.5.2 Technician I - Regulatory Compliance 4.5 ' (6) Technician - gA 4.5.2 (>) Technical Aide - Security 4.5.2 (1) Technical Aide " Fire Protection

( ) denotes number of exceptions or alternatives proposed in paragraph c above.

1.8-12

SlQIPP FSAR TABLE 1.8-1 (cont'd)

Operators, Technicians Maintenance 4 '.2 (7) Technical Aide Training 4.5.2 Technician I - Maintenance 4.5.2 Technician I - I&C 4.5.2 (5) Technician II - I&C 4.5.3 Electrician I 4.5.3 Planner Analyst 4.5.3 Senior Mechanic 4.5.3 Mechanic I 4.5.3 (5) Mechanic II 4.5.3 Painter and Pipe Coverer 4.5 '.2 (4) Senior Control Operator 4.F 1.2 (4) Control Operator 4.5.1.1 (4) Auxiliary Operator 4.5.1 ~ 1 (3) Control Operator Radvaste 4.5.1.1 (3) Auxiliary Operator - Radwaste 4.5.2 (7) Draftsmen

( ) denotes number of exceptions or alternatives proposed in paragraph c above.

1.8-13

CPBc.L Coxnxnents 8NPP Proof and Review Technical 8 pecif ication s Record Number: 743 Comment Type: ERROR LCO Number: 6.05.03.01 Page Number: 6-11 Section Number: 6.5.3.1 Comment:

CHANGE THE LAST SENTENCE OF THE PARAGRAPH TO READ AS FOLLOWS:

They shall also evaluate all CP&L LER's for their potential applicability to other CP&L units.

Basis SEE ITEM 744 THIS CHANGE IS NEEDED TO ACCURATELY REFLECT THE EXACT ORGANIZATION THAT PERFORMS THE VARIOUS REVIEWS. ALL ITEMS MENTIONED IN THE FINAL DRAFT ARE STILL COVERED% BUT HAVE BEEN MOVED TO, THEIR PROPER PLACE.

SHNP ADMINI STRATI VE CONTROLS Continued t'ESPONSIBILITIES

b. Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President-Harris Nuclear Project and the Manager-Nuclear Safety and Environmental Services of disagreement between the PNSC and the Plant General Manager.

However, the Plant General Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6. 1.1.

RECORDS 6.5.2.8 The PNSC shall maintain written minutes of each PNSC meeting that, at a minimum, document the results of all PNSC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President-Harris Nuclear Project and the Manager-Nuclear Safety and Environmental Services.

6.5.3 CORPORATE NUCLEAR SAFETY SECTION FUNCTION 6.5.3.1 The Corporate Nuclear Safety Section (CNSS} of the Nu'clear Suety and Environmental Services Oepartment shall function to provide independen~eview of plant changes, tests, and procedures; verify that REPORTABLE EVENTS ice in-vestigated in a timely manner and corrected in a manner that reduces the proba-bility of recurrence of such events; and detect trends that may not be apparent 1 I I I Pl II AR +htcw pafcarf(gil Appllaabfll@ ~ ++OIL CP~ Quefg, ORGANIZATION 6.5.3.2 The individuals assigned respons'ibility for independent reviews shall be technically qualified in a specified technical discipline or disciplines.

These individuals shall collectively have the experience and competence required to review activities in the following areas:

ah Nuclear power plant operations,

b. Nuclear engineering, C. Chemistry and radiochemistry,

d. Metallurgy, e, Instrumentation and control, Radiological safety,

g. Hechanical and electrical engineering,

h. Adiinistrative controls,

l. guaHty assurance practices, Jo Nondestructive testing, and

k. Other appropriate fields associated with the unique characteristics.

SHEARON HARRIS - UNIT 1 6-11

c(~

CPRL Coznxnents

>catstone

.'NPP Proof and Review Technical 8 pecif Re<. ord Number: 745 Comment Type.'MPROVEMENT LCO Number: 6. 05. 03. 09 Page Number: 6-13 Sect ion Number: 6. 5. 3. 9. e Comment:

IN THE SECOND LINF. DELETE THE WORD "AND".

REWORD THE LAST LINE TO THE FOLLOWING:

...plant safety-related structures, systems, or components which require written notification to the commission.

Basis THE DELETION OF THE WORD "AND" IS A GRAMMATICAL CORRECTION. THE ADDITION OF THE WORDS "SAFETY"RELATED" IS TO PROVIDE GREATER SPECIFICITY TO THE REQUIREMENT. AND, THE CHANGE TO THE END OF THE SENTENCE IS FOR CONSISTENCY WITH THE WORDING OF ANSI N18."I AND WITH THE WORDING OF BOTH THE ROBINSON AND BRUNSWICK TECH SPECS.

CPS L HAS A CORPORATE PROGRAM IN THIS AREA AND IT 1S NECESSARY THAT THERE BE CONSISTENCY BETWEEN THE REQUIREMENTS FOR THE VARIOUS PLANTS. THIS CHANGE PROVIDES THAT INTERNAL CONSISTENCY AS WELL AS BEING IN CONFORMANCE TO THE APPLICABLE STANDARD.

ADMINI STRATI VE CONTROLS r

REVIEW (Continued)

e. Violations> of applicable codes, regula fications, license requirements, ~ 's, orders, Technical Speci-inte al procedures or instruc-tions having nuclear safety significance, s'ificant operating abnormalities or deviations from normal and xpected performance of plant structures, systems, or components, ttlCJ/ XEOOIAZ. ukl~hl Bldg/lt47l04 7 0 TVC eo Mrg /gag~~

All REPORTABLE EVENTS;

g. All proposed modifications that constitute an unreviewed safety ques-tion as defined in Paragraph 50.59 of 10 CFR Part 50 or involve a change to the Technical Specifications; Any other matter involving safe operation of the nuclear power plant that the Hanager-Corporate Nuclear Safety Section deems appropriate for consideration or which is referred to the Manager-Corporate Nuclear Safety Section by the onsite operating organization or other functional organizational units within Carolina Power 4 Light Company; A rec nize indi tion of a 'una ic ate de ie sp o esig or erat n of tru ure sy ems, r m~e t t ld feet ucl r sa ty; nd lj'. Reports and minutes of the PNSC.

6.5.3. 10 Review of items considered under Specification 6.5.3.9.e, h and g<

above shall include the results of any investigations made and the recommenda-tions resulting from these investigations to prevent or reduce the probability of recurrence of the event.

RECORDS 6.5.3. 11 Records of Corporate Nuclear Safety Section reviews, including recommendations and concerns, shall be prepared and distributed as indicated below:

a. Copies of documented reviews shall be retained in the CNSS files.

Recommendations and concerns shall be submitted to the Plant General Manager and Vice President-Harris Nuclear Project within 14 days of completion of the review. A report summarizing the reviews encom-passed by Specification 6.5.3;9 shall be provided to the Plant General Manager and the Vice President-Harris Nuclear Project every other month.

C. A summation of recommendations and concerns of the Corporate Nuclear Safety Section shall be submitted to the Chairman/President and Chief Executive Officer and other appropriate senior management personnel at least every other month.

SHEARON HARRIS - UNIT 1 6-13

CP8c.L Comments NPP Proof and Review Technical Specifications Record Number: 746 Comment Type: IMPROVEMENT LCO Number: 6.05.03.09 Page Number: 6-13 Section Number: 6.5.3.9.i Comment:

DELETE ITEM i AND RELETTER j TO i. ALSO IN ITEM 6.5.3.10 CHANGE ITEM j TO i.

BaSiS CPE L HAS A CORPORATE PROGRAM FOR 3 REACTOR SITES AND IT IS HIGHLY DESIRABLE TO KEEP THE REQUIREMENTS FOR ALL UNITS COMPATABLE. THIS ITEM DOES NOT APPEAR IN THE BRUNSWICK NOR ROBINSON SPECIFICATIONS.(A IN ADDITION, THE REQUIREMENT IS SO BROAD AND VAGUELY WORDED THAT IT APPEARS ALMOST IMPOSSIBLE TO SET UP AN AUDITABLE PROGRAM TO COVER THE ITEM. IT APPEARS, IN GENERAL> TO COVER THE SAME GROUND AS 10CFR21 AND AS SUCH IS ALREADY COVERED BY ITEM e ABOVE.

J=

ADMINI STRATI VE CONTROLS REVIEW (Continued e.

fications, license requirements, ~

Violations of applicable codes, regulations, orders, Technical Speci-internal procedures or instruc-tions having nuclear safety significance, significant operating

$ )f/+1 abnormalities or deviations from normal and expected performance of RE~

plant structures, systems, or components udice'EpviAZ ~Ri~4 yq~<,~~>~p

>0 TYC Ce nnagg~~,

f. All REPORTABLE EVENTS;

g. All proposed modifications that constitute an unreviewed safety ques-tion as defined in Paragraph 50.59 of 10 CFR Part 50 or involve a change to the Technical Specifications; Any other matter involving safe operation of the nuclear power plant that the Manager-Corporate Nuclear Safety Section deems appropriate for consideration or which is referred to the Manager-Corporate Nuclear Safety Section by the onsite operating organization or other functional organizational units within Carolina Power 8 Light Company; e A rec nize indi tion of a una ic ate de ie Q'Lo o esig or erat n of tru ure sy ems, r m~e t sp t c ld feet ucl r sa ty; nd lj'. Reports and minutes of the PNSC.

6.5.3. 10 Review of items considered under Specification 6.5.3.9.e, h and g <

above shall include the results of any investigations made and the recommenda-tions resulting from these investigations to prevent or reduce the probability of recurrence of the event.

RECORDS 6.5.3. 11 Records of Corporate Nuclear Safety Section reviews, including recommendations and concerns, shall be prepared and distributed as indicated below:

a. Copies of documented reviews shall be retained in the CNSS files'.

Recommendations and concerns shall be submitted to the Plant General Manager and Vice President-Harris Nuclear Project within 14 days of completion of the review. A report summarizing the reviews encom-passed by Specification 6.5.3;9 shall be provided to the Plant General Manage and the Vice President-Harris Nuclear Project every other month.

C. A summation of recoaeendations and concerns of the Corporate Nuclear Safety Section shall be submitted to the Chairman/President and Chief Executive Officer and other appropriate senior management personnel at 1east every other month.

SHEARON HARRIS'- UNIT 1 6-13

CP Bc L Coxnxnenta gy ~ Pgsr, ~ /

SHNPP Pr oof and Review Technical Specifications Record Number: 746 Comment Type: IMPROVEMENT LCO Number: 6.05.03.09 Page Number: 6-13 Section Number: 6.5.3.9.i Comment:

DELETE ITEM i AND RELETTER j TO i. ALSO IN ITEM 6,5.3.10 CHANGE ITEM J TO i.

Basis CPS L HAS A CORPORATE PROGRAM FOR 3 REACTOR SITES AND IT IS HIGHLY DESIRABLE TO KEEP THE REQUIREMENTS FOR ALL UNITS COMPATABLE. THIS ITEM DOES NOT APPEAR IN THE BRUNSWICK NOR ROBINSON SPECIFICATIONS. IN ADDITION) THE REQUIREMENT IS SO BROAD AND VAQUELY WORDED THAT IT APPEARS ALMOST IMPOSSIBLE TO SET UP AN AUDITABLE PROGRAM TO COVER THE ITEM. IT APPEARSi IN GENERAL) TO COVER THE SAME GROUND AS 10CFR21 AND AS SUCH IS ALREADY COVERED BY ITEM e ABOVE.

(

ADMINI STRATI VE CONTROLS REVIEM (Continued}

e.

fications, license requirements, ~

Violations of applicable codes, regulations, orders, Technical Speci-internal procedures or instruc-tions having nuclear safety significance, significant operating abnormalities or deviations from normal and expected performance of plant structures, systems, or components ugnrEN ~~~id ~~>~

All REPORTABLE EVENTS; j

dJHIC,H AEOulAZ rO rh eo~~aSxaa.

g. All proposed modifications that constitute an unreviewed safety ques-tion as defined in Paragraph 50.59 of 10 CFR Part 50 or involve a change to the Technical Specifications;

h. Any other matter involving safe operation of the nuclear power plant that the Manager-Corporate Nuclear Safety Section deems appropriate for consideration or which is referred to the Manager-Corporate Nuclear Safety Section by the onsite operating organization or other functional organizational units within Carolina Power 8 Light Company; p' rec nize indi tion of a una ic ate de ie Q~iad sp t t o

ld esig or feet erat n of tru ure sy ems, r m~e+

ucl r sa ty; nd lj Reports and minutes of the PNSC.

6.5.3. 10 Review of items considered under Specification 6.5.3.9.e, h and gi above shall include the results of any investigations made and the recommenda-tions resulting from these investigations to prevent or reduce the probability of recurrence of the event.

RECORDS 6.5.3. 11 Records of Corporate Nuclear Safety Section reviews, including recommendations and concerns, shall be prepared and distributed as indicated below:

a. Copies of documented reviews shall be retained in the CNSS files.

b. Recoaeendations and concerns shall be submitted to the Plant General Manager and Vice President-Harris Nuclear Project within 14 days of completion of the review. A report summarizing the reviews encom-passed by Specification 6.5.3:9 shall be provided to the Plant General Manager and the Vice President-Harris Nuclear Project every other month.

c. A summation of recommendations and concerns of the Corporate Nuclear Safety Section shall be submitted to the Chairman/President and Chief Executive Officer and other appropriate senior management personnel at least every other month.

SHEARON HARRIS - UNIT 1 6-13

CPLL Coxnxnenta RNP P Proof an d Review Tech nical 8 pecif ication s Record Number: 774 Comment Type: ERROR LCO Number: F 05.04.03 Page Number: 6-15 Section Number: 6.5.4.3 Comment:

CHANGE THE TITLE TO SENIOR EXECUTIVE VICE PRESIDENT-POWER SUPPLY AND ENGINEERING AND CONSTRUCTION.

Basis TYPOGRAPHICAL

FiNAL DR%I ADMINI STRAT I VE CONTROLS RECORDS 6.5.4.3 Records of audits shall be prepared and retained.

6.5.4.4 Audit reports encompassed by Specification 6.5.4. 1 shall be prepared, approved by the Manager-Quality Assurance Services, and forwarded, within 30 days after completion of the audit, to the xecutive Vice President-Power Supply and Engineering and Construction, Senior ice President"Nuclear Generation, Vice President-Harris Nuclear Project, anager-Nuclear Safety and Environmental Services, Plant General Manager, and the management positions responsible for the areas audited.

AUTHORITY SEJM!dR 6.5.4.5 The Manager-Quality Assurance. Service Section, under the Manager-Corporate Quality Assurance Department, shall be responsible for the following:

a. Administering the Corporate Quality Assurance Audit Program.

b. Approval of the individuals selected to conduct quality assurance audits.

6.5.4.6 Audit personnel shall be independent of the area audited, 6.5.4.7 Selection of personnel for auditing assignments shall be based on experience or training that establishes that their qualifications are commen-surate with the complexity or special nature of the activities to be audited.

In selecting audit personnel, consideration shall be given to special abilities, specialized technical training, prior pertinent experience, personal character-istics, and education.

6.5.4.8 Qualified outside consultants, or other individuals independent from those personnel directly involved in plant operation, shall be used to augment the audit teams when necessary.

6.5.5 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM 6.5.5. 1 An independent fire protection and loss prevention inspection and audit shall be performed at least once per 12 months using either qualified offsite licensee personnel or an outside fire protection firm.

6.5.5.2 An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at inter-vals no greater than 36 months.

6.5.5.3 Copies of the audit reports and responses to them shall be forwarded to the Vice President-Harris Nuclear Project and the Manager-Corporate Quality Assurance.

6.6 REPORTABLE EVENT ACTION 6.6. 1 The following actions shall be taken for REPORTABLE EVENTS:

SHEARON HARRIS - UNIT 1 6-15

Shearon Harris Technical Specifications Resolution of Staff Comments Originator: go 6 ( Page: Cp -/7 Comant Date: g/ry(gb 75: le~a9a Comment:

We/e'fe. /he. %econ J Seo7ertce oF t'tg,/,a., /,

Resolution Basis Pe/efe se<eg 5e~7e~cC, f'gr S7 5, A/ever incle Jed Far atng o+er p/+~V, ge o. gene irerel.eirewenf'f47 Qc,S POj'een e,gtouC 3 4C r IJTC /LL5/0 1h )7t +e Xs nog gr il5 + for gh'e 7g 5 dr~

became e .g i can LC(9) htt5 ~O ACI Iek cr...t a cL e'u-rvei(la.~ce, EaZ been reeeepfe< ~erroner afg

+rem 4e I'narFr, 6v'e'~

+) ii'. Z>e-peed uF s/iv/sb s -,

(See enc.leaeck CtgPQ dF Cg& Vf'IQ I >a( PQVhPf?g QC~aateg Resolution Acce ted:

r~ I<'g,

/ a.

~

NRC CPSL il 9

Date:

Shearon Harris Technical Specifications Resolution of Staff Comments Originator: l)e~Wy 5 Page: t Conment Date: 3/f7/gf; gS,',F,V,a.

The leakage of primary coolant from ESP system elements (Pump seals, valves, etcl out ide of containment is an important variable in the evaluation'of the radiological consequences of a LOCA. The SER V-evaluations of the radiological consequences (Section 15.6.5.2) assumed ~

that such leakage is less than 1 gpm. tn order to assure the validity of this assumption to Shearon Harris, a maximum allowable integrated leak rate outside of containment of 1 gpm should be specified and confirmed at each refueling outage. A justification is needed if this requirement is not specified in the technical specifications.

For your reference, SRP 15.6.5, Appendix B states "The leakage for calculating the radiological consequences should be the maximum operational leakage and should be taken as two times the sum of the simultaneous leakaqe from all components ir. the re"ircu'ie .ion st",,:":-

above wQi ch tjsss technical pre. i-.i".:"+ion.' wo" id r= ngirs; ."ef ">'n.-,;!,;

c'v" 1;assis tn 'ip out 0 se vi c-"...

Resolution Basis ch~~pg 7S d,t'~~ +

,'nc I&a. ~ gp~ Ie~4~g ~

~eve'~v~ for IWf'- m cf g c. G Q c g c fe ~g, Se.e m+/mc/imcL 7.fn.m.r/fed

=/e f/-,

Resolution Acce ted:

NRC CP&L Date: Date:

0

- - Pn00FAt,'PEB,B(t;Opy ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS Continued

/~RA%

g. guality Assurance>for effluent and environmental monitoring; and

h. Fire protection program implementation.

6. 8. 2 Each procedure of Specification 6.8. 1, and changes thereto, shall be reviewed and approved in accordance with Specification 6.5.1 prior to imple-mentation and reviewed periodically as set forth in administrative procedures.

6. 8. 3 Temporary changes to procedures of Specification 6. 8. 1 may be made provided:

a. The intent of the original procedure is not altered;

b. The change is appro'ved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and

c. The change is documented, reviewed in accordance with Specifica-tion 6.5. 1, and approved within 14 days of implementation by the Plant General Manager or by the Manager'of the functional area affected by the procedure.

6.8.4 The following programs shall be established, implemented, and maintained:

a. Prima Coolant Sources Outside Containment A program to reduce leakage, to as low as practical levels, from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or*accident. The systems include:

1.

2.

. Residual Heat .Removal System ~

Safety Injection System, except boron injection recirculation subsystem and accumulator

3. Portions of the Chemical and Volume Control System:

a. letdown subsystem, including demineralizers

b. boron re-cycle holdup tanks

c. charging pumps Containment Spray Syst m, except spray additive subsystem and INST .

l/P. Post-Accident Sample System SHEARON HARRIS," UNIT 1 6-17

gf<

Shearon Harris Technical Specifications Resolution of Staff Comments Originator: Pam Page:

Comment Date: 8 4/8&

Comment:

Review Guidelines: The licensee shall affirm that each of the numerical values specified in the Final Draft of the Technical Specifications is in accordance with, or conservative with respect to, the Analyses of Record, making appropriate allowances for instrumentation error.

pS Ia

~,'(( mof be

'yees'f~~+~'asis Pity Resolution

)~)ejretl 0F gg o.pp/joe~/ js zp7+g >

in ~ r g I epe>red +a Sge.c.s are coasts&

~, Fager+. Xak ~.d 6e.

0 s- k ui J7 F~ilifJ de P~

rei.ue4,boj,PS B goes spa+/'s bjsV orjc repmire~en+

cergi Fg ccrc.g lort Resolution Acce ted:

NRC CPSL Date: Date: