ML18019A818

Forwards Addl Info Re Input Into Final Draft Tech Specs for Facility.Info & Justification Provided for Each Item. Marked-up Tech Specs Encl

ML18019A818
Person / Time
Site:	Harris
Issue date:	05/14/1986
From:	ZIMMERMAN S R CAROLINA POWER & LIGHT CO.
To:	DENTON H R Office of Nuclear Reactor Regulation
References
NLS-86-145, NUDOCS 8605200312
Download: ML18019A818 (289)

Text

/REQULAT XNFORNATION DISTRIBUTION BTEN (RIBS)ACCESSION NBR: 8605200312 DOC.DATE: 86/05/14 NOTARIZED:

NQ DOCVET 0 FACIL: 59-400 Bheav on Harl'35 Nucleal'DUJer Planti Un3t ii Cav olina 05000400 AUTH.NAME AUTHOR AFFILIATION ZINNERNANi S.R.Cav olina Poeer 5 Light Co.RECIP.NANE RECIPIENT AFFILIATION DENTONi H.R.Office of Nuclear Reactor Regulation.

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SUBJECT:

Forwards addi info re input into f inal dv af t Tech Specs for facility.Info 0 justification provided for each item.Marked-up Tech Specs encl.DXSTRIBUTION CODE: B001D COPXES RECEIVED LTR ENCL SIZE TITLE: Licensing Submittal:

PSAR/FBAR Amdts 8c elated Cov'v espondence NOTES: App l ication f ov pev mi t v enewal f i l ed.05000400 RECIPIENT XD CODE/NANE PWR-A ADTS PWR-A EICSB PWR-A PD2 LA BUCNLEY.B 01 PWR-A RSB INTERNAL.ADN/LFNB, XE FILE XE/DGAVT/OAB 21 NRR PWR-A'DTB NRR ROE>l).L NRR/DHFT/rtTB RCN2 EXTERNAL: 24X DNB/DSS (Al'IDTS)NRC PDR 02 PNL CRUELER COPIES LTTR ENCL 1 1 2 2 1 1 2 2 1 1 1 0 1 1 1 1 0 1 1 1 1 3 3 1 1 1 1 1 RECXPXENT ID CODE/NAl'fE PWR-A EB PWR-A FOB PWR-A PD2 PD PWR-A PBB ELD/HDB1 IE/DEPER/EPB 36 NRR BWR ADTS NRR PWR-B ADTS T/HFIB 04 RN/DDANI/NXB BNL(ANDTS ONLY)LPDR 03 NBIC 05 COP XES LTTR ENCL i.1 1 0 1 0 0 1 1 0 1 1 1 1 1 TOTAL NUl'1BER OF COPXES REQUIRED LTTR 34 ENCL 28

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$ÃQP Carolina Power&Light Company SERIAL: NLS-86-105 MAY 14 tgsg 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.1-DOCKET NO.50-000 TECHNICAL SPECIFICATIONS

Dear Mr.Denton:

Carolina Power k Light Company (CPRL)submits additional information for input into the Final Draft Technical Specifications (TS)for the Shearon Harris Nuclear Power Plant.Attachment 1 provides the information as well as justification for each item.Attachment 2 provides a marked-up copy of the affected TS pages.The attachments include information for the lengthy tables in the TS that CPRL had previously proposed to control procedurally and administratively.

CPRL will continue to participate in industry TS improvement programs with the objective of removing these lengthy tables.If you have any questions, please contact Mr.Gregg A.Sinders at (919)836-8168.Yours very t y, GAS/pgp (3780GAS)Attachments cc: Mr.B.C.Buckley (NRC)Mr.R.A.Benedict (NRC)Mr.G.F.Maxwell (NRC-SHNPP)

Dr.3.Nelson Grace (NRC-RII)Mr.Travis Payne (KUDZU)Mr.Daniel F.Read (CHANGE/ELP)

Wake County Public Library S.R.Ztmmerman Manager Nuclear Licensing Section Mr.Wells Eddleman Mr.3ohn D.Runkle Dr.Richard D.Wilson Mr.G.O.Bright (ASLB)Dr.3.H.Carpenter (ASLB)Mr.3.L.Kelley (ASLB)8b08200 OC5 080 PQR>g2 8boppqpp 8~F90 411 Fayettevilte Street o P.O.Box 1551~Raleigh.N.C.27602

/

SLAM'T X 860540031 2 ClP'Bc,X Cloxnxnenta Px-os and Reviewer Techxucal Sgecif ications.Record Number: 508 LCO Number: 3.01.01.03 Section Number: ACTION A.1 Comment Type: ERROR Page Number: 3/4 1-4 Comment;CHANGE"0 pcm/F" TO"restore the MTC to within the above 1imits"'.Basis THIS CHANGE WAS INADVERTENTLY NOT INCLUDBD WHEN THE OTHER CHANGES WERE MADE TO THIS SPECIFICATION IN AN EARLIER REVISION.

GHNT P'x nnX a.net Ram9.ew Tea&n9.mal S~a mid'xca.t,xnan a Record Number: 500 LCO Number: 3.06.03 Section Number: TABLE 3.6-1 Comment Type: IMPROVEMENT Page Number: 3/4 6-14,15,16 Comment: INSERT TABLE 3.6-1 Basis THIS CHANGE IS MADE TO INCORPORATE THE CONTAINMENT ISOLATION VALVE LIST INTO THE TECH SPECS.THIS TABLE DIFFERS FROM FSAR TABLE 6.2.4-1 IN THAT IT LISTS ONLY COMTAXNMENT ISOLATION VALVES AS INDICATED IN THE SECOND TO LAST COLUMN IN THE FSAR TABLE.SOME VALVE CLOSURE TXMES HAVE BEEN CHANGED FROM THE CURRENT FSAR VALUES.THXS XS BASED ON THE FACT THAT THE VALVES WERE NOT ASSUMED TO BE CLOSED FOR 60 SECONDS (AS PERMITTED BY THE SRP).FORTY-FIVE (45)SECONDS WAS CHOSEN AS A CONSERVATIVE LIMIT TO ESTABLISH WITHXN THE ANALYSIS ASSUMPTIONS.

A CHANGE TO THE FSAR IS IN PREPARATION TO REFLECT THE ANALYSIS VALUES.IN ADDITION, SOME VALVES, PARTICULARY IN THE SI AND CONTAINMENT SPRAY SYSTEMS, WHICH SHOW A RESPONSE TIME IN THE FSAR TABLE, ARE SHOWN WITH ISOLATION TIMES OF"NA" IN THE TECHNICAL SPECIFICATION TABLE.THIS IS DUE TO THE.FACT THAT THESE VALVE DO NOT"ISOLATE", BUT OPEN UPON RECEIPT OF AN ACCIDENT SIGNAL.THE REMAINING DESCREPANCIES ARE THE RESULT OF CHANGES TO THE FSAR TABLE WHICH WILL BE FORMALLY DOCUMENTED IN AN UPCOMING LETTER.

s CP&.L Comments HNPP Proof and Review Technical Specifications Ri:.<..ovd Number: 501 LCO Numbo>~:.'>.0 c~10.02 S<.ol ion Number': TABLE 3.7-3 Comme:: l.Type: IMPROVE!'IFX'f Pa>>i.Number: 3 il 7"30, 31, 3" 1 C vmm>'.u.t INSFRT'1'ABLE 3.7"3 Bas'HIS CHANGE IS MADE TO THE PREACT10N AND i~1ULT ICYCLE SPRINKLER TABLE INTO THE TECH SPECS.THE ADDITION OF THE FOOTNOTE TO THE TABLE IS NECESSITATED BY THE FACT THAT THESE SPRINKLERS ARE ACTUATED BY THE INSTRUi4ENTATION OF TABLE 3.3-.1 I.SINCE THE INSTRUMENTATION IS NOT REQUIRED TO BF.OPERABLE DURING THE TYPE A TESTS, THE SPRINKLERS SHOULD NOT BE.

CP S.L Coxnxnents HNPP Proof and Review Technical Specifications 1(<<..oi<3 Number: 502 L(:0 Xumb<::v: 3.0 (.'.0.03 S<<.t i onXumh<": TABLE 3.1-4 Commcu1 Comm<..ut.

Tyu<': I:.'41F'RO'CEMENT Pa p'<=X umb<..<': 3/4 7"33,:?-), 3;i I.<SEHT TA,BI E 3.7-4 BGS 1S THIS CHANGE IS TO INCORPORATE TABLE 3.7-4 ON FIRE HOSE STATIONS INTO THE TECH SPECS.

CPRL Coxaxnenta HNPP Proof and Review Technical Specifications f{eeovd Numbe;: 503 L('0 Xumi)~<':,'(.0!.04.01 Section Numbe.': TABLE 3.8-I Commen t Type.IMPROVEMENT Page Number: 3'-I A-17 k i'.>Commen t INSERT TABLE 3.8-1 Bas:s THIS CHANGE IS MADF.TO INCORPORATE TABl.E 3.8-1 ON THE CONTAINMENT PENETRATION CONDUCTOR OVERCURREXT PROTF(:T IVF, DFVTCES INTO THF.'l'ECH SPECS.

CPRL Coxnmenta HNP P P r oof an d Reviewer Tec h nical Specif ications Heooi d Numbe: ')04 LCO Xumj>ec: 3.08.04.0" Sl!oi'on iXumbet': TABLE 3~8 Comment: Comment Type:[MPROVEMEXT Pago Number':.'3i'8-"0 F.'I'ASEHT TABLF.3.8-2 Basis T f1 IS CH KbIGE IS MADE TO INCORPORATE TABLE 3.8-2 O.'J MOTOR OPERATED VALVE TERMAL OVERLOAD PROTECTION

[770 THE TECH SPECS.

I k CP8c.L Coraments HNPP Proof and Review Technical Specifications Record Number: "05!.CO Number: 3.OU.03.Ol Sec t~on Number: 3.8.3.I.g S: h Comment Type: I~!PROVEMENT Pa j~e Number: 3/l 8-I'Commen t ITE~!g-INSERT T!IE WORDS'and chargers l A-SA or LB-SA" AFTER TffE WORDS"Emerpency Bat'tery IA-SA".ITEM h-INSERT THE WORDS"and chargers IA-SB ocf IB-SB" AT THE END OF THE ITEN.Basis Tf!IS CHANGE IS MADE FOR CONSISTENCY WITH SPEC I!'ICATION 3.8~3.'-'

a CPKL Cornxnents HAPP Pr oof and Review Technical Specifications Record Number: 506 LCO Number: 6.02.02 Section Number: 6'.2'Comment Type: IMPROVEMENT Page Number:*-1 Comment: CHANGE THE SECOND SENTENCE TO READ A S FOLLOWS: The Fire Brigade shall not include any members o$the minimum shiest crew necessary%or the safe shutdown o$the unit as specified in Table 6.2-1 nor any personnel required%or any other essential$unctions during a fire emergency; and Basis THIS CHANGE IS MADE TO CLARIFY THE NECESSARY PERSONNEL REQUIRED FOR SAFE SHUTDOWN AS DEFINED ELSEWHERE IN THE TECH SPECS.

1 g r CPBcL Comxaents SHNPP Proof and Review Technical Specifications Record Number: 509 LCO Number: 5.03.01 Section Number: 5.3.1 Comment Type: IMPROVEMENT Page Number: 5-6 Comment: DELETE THB FOLLOWING WORDING FROM THB SECTION: "...and contains a maximum total weight of 1776 grams uranium.The initial core loading shall have a maximum enrichment of 3.5 weight percent U 235." ALSO DELETE"...and shall have a maximum enrichment of 3.9 weight percent U 235." Basis THE SPECIFIC VALUES IN THIS PARAGRAPH ARE NOT NBEDBD AND COULD RESULT IN UNECESSARY ADMINISTRATIVE WORKLOADS FOR BOTH CP&L AND NRC PERSONNEL TO PROCESS MINOR CHANGES.THE VALUE OF THB TOTAL GRAMS OF URANXUM PER FUEL ROD IS NOT USED IN ANY SAFETY ANALYSIS AND CAN BB AFFECTED BY MINOR CHANGES IN THE MANUFACTURXNG PROCESS.THE INCLUSION OF THE MAXIMUM ENRICHMENTS IS ALSO NOT NECESSARY SINCE THB SPECIFIC ENRICHMENTS TO BE PRESENT XN ANY FUEL CYCLE ARE EVALUATED FOR THAT CYCLE.THIS EVALUATION MUST ENSURE THAT ALL SAFETY CRITERIA ARE MET REGARDLESS OF SPECIFIC ENRICHMENT VALUES.DROPPING THESE VALUES FROM SECTION 5 WILL HAVE NO ADVERSE IMPACT ON PLANT SAFBTY OR REGULATORY CONCERNS AND WILL ELIMINATE UNNECESSARY TECHNICAL SPECIFICATION CHANGES' C'P8r L Coxnxnents HNPP Proof and Review Technical Specifications Record Number: 507 LCO Number: 5.06.03 Secti an Number: 5.6.3 Comment Type: IMPROVEMENT Page Number: 5-7 Camment: LINE, 2-DELETE THE WORDS"IN FIXED RACKS".BBsi 8 THIS CHANGE IS PROPOSED TO CLARIFY THE DESCRIPTION OF THE TYPE OF RACKS USED AT SHNPP.THE RACKS USED ARE FREE STANDING AND ARE NOT"FIXED" TO THE POOL FLOOR OR WALLS (SEE FSAR SECTION 9.i)

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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIEHT SHNPP RFVlS!OX i@1'86 PRQGI'NQ Ill.RB'I t,'QP'(LIMI TING COHO ITION FOR OPERATION2.3.The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored'to within its limit for the all rods withdrawn condition; and A Special Report is prepared and submitted to the Commission, pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.Less negative than-"42 pcm/'F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICASILITY:

Specification 3.1.1.3a.'-MODES 1 and 2" only"".Specification 3.1.1.3b.-MODES 1, 2, and 3 only"".ACTEQH: a.With the MTC more positive than the limit of Specification 3.1.1.3a.above, operation in MODES 1 and 2 may proceed provided: 3..Control rod withdrawal limits are established and maintained sufficient to restore the MTC to Hs~pos+tHv~Ma&~~

~,~,~<naos uni~s within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDSY within%he next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;THls Gf+fgg cP)L~~R or V-23-8(her~g, QS-gg-pig b.With the MTC more negative than the limit of Specification 3.1.1.3b.above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.l<vie mp%o 707ii PATED i HER LAL POu i=/.t a i 8 c limni~rc~ii%re+4 i-p~i..~++~+p./imp i c~l 00 4 RA~+i~HE~q g p~~~p"With k greater than or equal to 1.eff""See Special Test Exceptions Specification 3.10.3.SHEARON HARRIS-UNIT,l 3/4 1-4

I<00~~'m~atIEIIt t;~>~CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION P&itglC)hJ mv 586 3.6.3 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION: With one or more of the containment isolation valve(s)specified in Table 3.6-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is opeq~and:.'"'

a.Restore the inoperable valve(s}to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or b.Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve secured in the isolation position, or c.Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or d.Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLO SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENT5 4.6.3.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isola-tion time.Cg ARE gCAV CRT SHEARON HARRIS-UNIT 1'/4 6" 14

CONTAINMEHT SYSTEMS CONTAINMEHT ISOLATION VALVES SHNPP RPlls3C)8~e$86 M{lkf hf]3 3Blll lloI"l SURVEILLANCE RE UIREMEHTS Continued 4.6.3.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLO SHUTDOWN or REFUELING MOOE at least once per 18 months by: a.Verifying that on a Phase"A" Isolation test signal, each Phase"A" isolation valve actuates to its isolation position;b.Verifying that on a Phase"8" Isolation test signal, each Phase"8" isolation valve actuates to its isolation position;and YFplMc8nod V i ig t<<i~t signal, each normaliaad preentry purge makeup and exhausthvalve actuates to its isolation position~c g~'IAl~m4 al~&Pcs+w V.elie 5 4.6.3.3 The isolation time of each power-operated or automatic valve of Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5'.ga g~, iNJech0~5+esW ssfnlpli e pc4~e@~~g 4 oP/~~I~g~l ye geceiai~g+~~spa w>i+s Iso'+io<'p>>i+'/oL&y g~g I 5fewm XSoI~$ioa+eS 4 S g AA lp$ip~gp JpQ PcfL4Q P~e J Q w4wc%so lH~io~+c>p/end ppJ A ea iSol~ho~~+l~c hack+4-es+i+s iSol~+~Z yoSI SHEARON HARRIS-UNIT 1 3/4 6-15

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES'SH RPP P~!tent&ON$86 NETRATION NO.VALVE NO.CP&L (EBASCO)FUNCTION MAXIMUM ISOLATION APPLICABLE TIME (SEC)NOTES 1.PHASE A ISOLATION 1MS-25 MAIN STM LOOP A PRIMARY (MS-V122)SAMPLING PANEL 45 1, 3)7 1MS-27 (Ms-124)1MS-29 (MS-126)1CS-7 (CS-V511)1CS-8 (Cs-V512)1CS-9 (CS-V513)MAIN STM LOOP B PRIMARY SAMPLING PANEL MAIN STM LOOP C PRIMARY SAMPLING PANEL CVCS NORMAL LTDN ISOL CVCS NORMAL LTDN ISOL CVCS NORMAL LTDN ISOL 45 (0 lo lo 1, 3)7 1, 3)7 12 1CS-11 (CS-V518)1CS-470 (CS~V516)CVCS NORMAL LTDN ISOL j CVCS SEAL WTR RETURN&EXCESS LTDN lo NA 12 1CS-472 (Cs-V517)CVCS SEAL WTR RETURN&EXCESS LTDN NA'3 33 37 38 40 1SP-209 (SP-V4O8)1SP-208 (SP-V4O9)1CC-176 (CC-V172)1CC-202 (CC-V182)1RC-161 (RC-D525)lED-121 (WL-L600)GAS RETURN FROM P.A.S.S.SKID g2 GAS RETURN FROM P.A.S.S.SKID g2 CCW TO RCDT&EXCESS LTDN HEAT EXCH'CW FROM RCDT&EXCESS LTDN HEAT EXCH'EACTOR MAKEUP WTR TO PRT RCDT PUMP DISCH.IO IO 45 lO 4 1, 2 1, 2 TABLE3"6-1 (PLP)

C TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNPP g?W/lPj~s,i MAY 1986 NETRATION NO.VALVE NO.CP&L (EBASCO)FUNCTION MAXIMUM ISOLATION TIME (SEC)APPLICABLE NOTES 42 1ED-125 RCDT PUMP DISCH (WL-D650)74 1ED-94 (MD-V36)CTMT SUMP PUMP DISCH 74 76A 76B 1ED-95 (MD-V77)1SI-179 (SI-V554)1SI-263 (SI-V555)CTMT SUMP PUMP DISCH ACCUMULATOR FILL FROM RWST ACCUMULATOR TO RWST 45 lO 3, 4 3, 4 76B 1SI-264 (SI-V550)ACCUMULATOR TO RWST lo 3, 4 77A77B 77B 1SI-287 (SI-V530)1RC-141 (RC-D528)1RC-144 (RC-D529)N2 TO ACCUMULATORS PRT N2 CONNECTION j PRT N2 CONNECTION IO 3, 4 77C lED-164 RCDT-H SUPPLY'(WG-D590) 2 lo77C 1ED-161 (WG-D291)RCTD-H SUPPLY 2 78A 78A 1SP-948 (SP-V111)1SP-949 (SP-V23)RCS SAMPLE RCS SAMPLE 45 45 78B 1SP-40 (SP-V11)PRESSURIZER LIQ SAMPLE 45 TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNVP Pwlteir~r hfAY>986 ENETRATXON NO.78B VALVE NO.CPGL (EBASCO)1SP-41 (SP-V12).FUNCTION PRESSURIZER LIQ SAMPLE MAXIMUM ISOLATION TIME (SEC)45 APPLICABLE NOTES 78C 78C 78D 1SP-59 (SP-Vl)1SP-60 (SP-V2)1SP-78 (SP-V113)PRESSURIZER STEAM SAMPLE PRESSURXZER STEAM SAMPLE ACCUMULATOR SAMPLE 45 45 78D 1SP-81 (SP-V114)1SP-84 (SP-V115)ACCUMULATOR SAMPLE ACCUMULATOR SAMPLE 45 45 78D 1SP-85 (SP-V116)ACCUMULATOR SAMPLE 45 vw 79 80 88-88 91 1FP-35 (FP-V 4)1IA-216 (IA-V192')

1SP-201 (SP-V406)1SP-200 (SP-V407)1SW-240 (SW-B89)FIRE ATER ST PXPE SUPPL INSTRUMENT AIR SUPPLY LIQUID SAMPLE RETURN FROM PASS SKID gl LIQUID SAMPLE RETURN FROM PASS SKID 8'1 SERVICE WATER FROM NNS FAN COILS 45 45 3 f NA 91 92 105 1SW-242 (SW-B90)1SW-231 (SW-B88)1FP-347 (FP-B l.-)SERVICE WATER FROM NNS FAN COILS SERVICE WATER TO NNS FAN COILS FXRE WATER SPRINKLER SUPPLY 45 45 TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNPP A~/lc;!~4i MAY 1986 ENETRATION NO.VALVE NO.CP&L (EBASCO)FUNCTION MAXIMUM ISOLATION APPLICABLE TIMa (SaC)eorEs 108 108 109 1AF-155 (AF-V162)1AF-153 (AF-V163)1AF-159 (AF-V164)AUX.F.W TO S/G A (HYDRAZINE)

AUX.F.W.TO S/G A (AMMONIA)AUX.F.W TO S/G B (HYDRAZINE) 10 10 10 1, 3,7 1~3i7 1, 3I7 109 lAF-157 (AF-V165)AUX F.W TO S/G B (AMMONIA)10 1, 3,7 110 110 1AF-163 (AF-V166)'AF-161, (AF-V167)AUX.F.W.TO S/G C (HYDRAZINE)

AUX.F.W.TO S/G C (AMMONIA)10 10 1, Big 1, 3,7 73A 1'3B 73B 1SP-12 (SP-V300), 1SP-915 (SP-V348)1SP-9't I (SP-V301)1SP"917 (SP"V349)RAD MONITOR&H2 ANALYZER RAD MONITOR&Hg ANALYZER j 2 RAD MONITOR&H2 ANALYZER RAD MONITOR&H2 ANALYZER 45 45 86A 1SP-42 (SP-V308)HYDROGEN ANALYZER 45 3, 4 86A 86B 1SP-919 (SP-V314)1SP-62 (SP-V309)HYDROGEN ANALYZER HYDROGEN ANALYZER 45 3, 4 3, 4 86B 1SP-56 (SP"V315)HYDROGEN ANALYZER 45 3, 4 TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNPp R&(jp tn~i MAY>g86 ENETRATION NO.VALVE NO.CP&L (EBASCO)FUNCTION MAXIMUM ISOLATION APPLICABLE TIME (SEC)Aov'Gs 2.PHASE B ISOLATION 35 36 1CC-208 (CC-V170)1CC-297 (CC-V184)CCW TO RCP CCW FROM RCP IO NA NA 36 39 39 1CC-299 (CC-V183)1CC-249 (CC-V191)1CC-251 (CC-V190)CCW FROM RCP CCW FROM RCP THERMAL BARRIERS CCW FROM RCP THERMAL BARRIERS/O JO NA NA NA 3.SAFETY INJECTION ACTUATION'7 1CS-238 (CS-V610)1SI-3 (SI-V5O5)CVCS NORMAL CHARGING SI TO H GH H COLD LEG NA 17 51 1-4 (SI-O6)1BD-11 (BD-V11)S 0 HIGH HEAD COLD L S/G A BLO OWN NA 45 1, 3)'753 54 55 1BD-30 (BD-V15)~1BD-49 (BD-V19)1SP-217 (SP-V120)1SP-222 (SP-V121)1SP-227 (SP-V122)S/G B BLOWDOWN S/G C BLOWDOWN ,S/G A SAMPLE S/G B SAMPLE S/G C SAMPLE 45 45 45 45 1, 3i7 1, 3)7 1, 3i7 11 3 7 1, 3,7 TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNPP P Phl t)~hi MAY 1986 VALVE NO.PENETRATION CP&L NO.(EBASCO)FUNCTION 4.CONTAINMENT VENTILATION ISOLATION MAXIMUM ISOLATION APPLICABLE TIME (SEC)NOT'ES 57 57 57 58 58-58 59 98 CP-B1 (CP-B1)CP-B3 (CP-B3)CP-B4 (CP-B4)CP-B2 (CP-B2)CP-B7 (CP-B7)CP-B5 (CP-B5)CP-B8 (CP-B8)CP-B6 (CP-B6)CB-B1 (CB-B1)CB-B2 (CB-B2)CONTAINMENT ATMOSPHERE PURGE MAKEUP (8")CONTAINMENT ATMOSPHERE PURGE MAKEUP (42")CONTAINMENT ATMOSPHERE PURGE MAKEUP (42")CONTAINMENT ATMOSPHERE PURGE&MAKEUP (8")CONTAINMENT ATMOSPHERE PURGE EXHAUST (42")CONTAINMENT ATMOSPHERE PURGE EXHAUST (8")CONTAINMENT ATMOSPHERE PURGE EXHAUST (42")CONTAINMENT ATMOSPHERE PURGE EXHAUST (8")CONTAINMENT VACUUM RELIEF CONTAINMENT VACUUM RELIEF 3.5 15 15 3.5 15 3.5 15 3.5 3, 6 3, 6 3, 6 3, 6 5.CONTAINMENT SPRAY ACTUATION 23 1CT-50 (CT-V21)CONTAINMENT SPRAY fV A 24 1CT-88 (CT-V43)CONTAINMENT SPRAY TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNpp pgQtg)my~MAY 1986 VALVE NO.PENETRATION CPGL NO.(EBASCO)FUNCTION MAX1MUM ISOLATION APPLICABLE TIME (SEC)NOTES 6.MAIN STEAM LINE ISOLATION 1MS-80 (MS-V1)1MS-81 (MS-F1)1MS-231 (MS-V59)1MS-82 (MS-V2)1MS-83 (MS-F2)1MS-266 (MS-V60)1MS-84 (MS-V3)1MS-85 (MS-F3)1MS-301 (MS-V61)MSIV (S/G A)MSIV BYPASS pRblP TD Qp~~E p$Gg MS%0=8eNB;MSIV (S/G B)MSIV BYPASS p~a x 4~~H'O CNS G~MS$8=68ÃB MSIV (S/G C)MSIV BYPASS QQQZH~Q,O~DChl SER MS%MHBBB 10 45 10 45 10 45 1, 5 1,3,4)7 1, 3)7 1, 5 1, 3, 4,7 1, 3)7 1$3p 4 7 1, 3)7 7.MAIN FEEDWATER LINE ISOLATION 1FW-159 FEEDWATER LOOP A (FW-V26)1, 3g/1FW-307 (FW-V123)1FW-165 (FW-V89)FEEDWATER LOOP A BYPASS VALVE FEEDWATER LOOP A (HYDRAZINE)

[O 45 1, 3)7.1, 3)7 TABLE3-6-1 (PLP)

0 PENETRATION NO.VALVE NO.CP&L (EBASCO)1FW-163 (Fw-V90)1FW-277 (Fw-V27)TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES FUNCTION FEEDWATER LOOP A (AMMONIA)FEEDWATER LOOP B 8HNPP P&llel~e i MAY 1986 MAXIMUM ISOLATION APPLICABLE TIME (SEC)NOTES 1, 3)7 1, 3)7 1FW-319 (FW-V124)1FW-279 (Fw-V91)1FW-281 (FW-V92)1FW-217 (Fw-vzs)FEEDWATER LOOP B BYPASS VALVE FEEDWATER LOOP B (AMMONIA)FEEDWATER LOOP B (HYDRAZINE)

FEEDWATER LOOP C Io 45 1, 3,7 1, 3)7 1, 3)7 1, 3)7 108 1FW-313 (FW-V1Z5)1FW-223 (FW-V93)1FW-221 (FW-V94)'AF-64 (AF-V156)FEEDWATER LOOP C BYPASS VALVE FEEDWATER LOOP C (AMMONIA)FEEDWATER LOOP C (HYDRAZINE)

AUXILIARY FEEDWATER A PREHEATER BYPASS lO 45 45 10 1, 3)7 1, 3)7 1, 3)7 1, 3)7 109 1AF-102 AUXILIARY FEEDWATER B ('AF-V157)

PREHEATER BYPASS 10 1, 3)7 110 1AF-81~(AF-V158)AUXILIARY FEEDWATER C PREHEATER BYPASS 10 1, 3)7 8.REMOTE MANUAL VALVES 2 1MS-58 (MS-P18)1MS-60 (MS-P19)1MS-62 (MS-P20)S/G PORV (S/G A)S/G PORV (S/G B)S/G PORV (S/G C)NA.NA NA 1, y,g,-l 1, 3,9,7 1 p 3)'l)I TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES SHNPP P Ml)Q)&bl MAY$986 PENETRATION NO.VALVE NO.CP&L (EBASCO)1CS-341 (CS-V522)FUNCTION CVCS SEAL WATER TO RCP A MAXIMUM ISOLATION APPLICABLE TIME (SEC)NOTES NA 10 1CS-382 (CS-V523)CVCS SEAL WATER TO RCP B NA 15 15 1CS-423 (CS-V524)1RH-1 (1RH-V502) 1RH-2 (1RH-V503)

CVCS SEAL WATER TO RCP C RHR PUMP SUCTION (TRAIN A)RHR PUMP SUCTION (TRAIN A)NA NA 1, 4'A 1 4 t 1 16 1RH-39 (1RH-V500)

RHR PUMP SUCTION (TRAIN B)NA 1, 4 1620 21 22 1RH-40 (1RH-V501) 1SI-359 (SI-V587)1SI-107 (SI-v5oo)1SI-86 (SI-V501)1SI-52 (SI-V502)RHR PUMP SUCTION (TRAIN B)SI LOW HEAD TO HOT LEG SI HIGH HEAD TO HOT LEG SI HIGH HEAD TO HOT LEG SI HIGH HEAD TO COLD LEG NA NA NA NA NA 1;4 I 5$-'3 (5g-vivos)t.fggg f-IEA'0 CoLP LG(j~R (ss-'I (Sg-V50la)wo HSING IIEbo Covg L+~(g5-~(g s-VS)I r 5-72 V 9)~'@A a sMh!3m O<P'n.><<5'%At C.~<P~)~~g.B IW t='I R)7 TABLE3-6-1 (PLP)

4 VALVE NO.ENETRATION CPRL NO.(EBASCO)63 CM-B5 (cM-a5)TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES FUNCTION H2 PURGE EXHAUST MAXIMUM ISOLATION TIME (SEC)SHNPP P~)gt+Kl MAY 1986 APPLICABLE NOTES 9.MANUAL VALVES 17 34 41 1SI-43 (sx-vac)U.v.-6 (LT-V2)1SA-80 (SA-V14)SI-HIGH HEAD TO COLD LEGS ILRT ROTOMETER (LOCKED CLOSED)-SERVICE AIR (LOCKED CLOSED)NA NA NA 1, 4 3 4 3,4 42 44-45 45 61 1ED-119 (wr.-o~J)1SP-145 (SF-D164)1SP-144 (SP-D165)1SF-118 (SP-D25)1SP-119 (SP-D26)CM-B6 (GM-B6).RCDT PUMP DISCH BYPASS (LOCKED CLOSED)REFUELING CAVITY CLEANUP (LOCKED CLOSED)REFUELING CAVITY CLEANUP (LOCKED CLOSED)A REFUELING CAVITY CLEANUP (LOCKED-CLOSED)REPUELING CAVITY CLEANUP (LOCKED CLOSED)'hL H PURGE MAKEUP (LOCKED C OEED)NA NA NA NA 3 4 3 4 3,4 (~i 4 3)4, f t=P.f55-f-gg~us~SraPPOAe8'u A'<Y (Fp vs~)3,9 TABLE3-6-1 (PLP)

~l PENETRATION NO.62 63 90 96 109 110 VALVE NO.CP&L (EBASCO)l L7-10 (LT-V4)CM-B4 (CM-B4)1DW-63 (DW-V120)(LT-Vl)lAF-174 (,AF-V189) 1AF-173 (AF-V190)1AF-175 (1 AF-V191)TABLE 3'-1 (Cont'd)CONTAINMENT ISOLATION VALVES FUNCTION ILRT (LOCKED CLOSED)H2 PURGE EXHAUST (LOCKED CLOSED)DEMIN WATER SUPPLY (LOCKED CLOSED)ILRT (LOCKED CLOSED)WET LAY-UP TO STM GEN A AF HEADER WET LAY-UP TO STM GEN.B AF HEADER WET LAY-UP TO STM GEN.C AF HEADER MAXIMUM ISOLATION TIME (SEC)NA NA NA NA NA NA NA SHRPP P+\gig)~hl MAY 1986 APPLICABLE NOTES 34 3q 4 3)4 l)3)7 l,3,7 l)3, 7 10 CHECK VALVES 8 1CS-477 (CS-V515)CVCS NORMAL CHARGING NA NA 12 2'3 ,24 1CS"471 (CS"V67)1CT-53 (CT-V27)1CT-91 (CT-V51)CVCS SEAL WATER RETURN&EXCESS LETDOWN CONTAINMENT SPRAY TRAIN A CONTAINMENT SPRAY TRAIN B NA NA NA NA'A NA TABLE3-6-1 (PLP)

NETRATION NO.35 36 39 40 41 59 77A 79 80 90 92 94A VALVE NO.CPSL (EBASCO)'cc-211 (CC-V171)lcc-298 (CC-V51)lcc"250 (CC-V50)1RC-164 (Rc-V525)1SA-82 (SA-V15)CB-Vl (CB-Vl)CM-Vl (CM-Vl)1SI-182 (SI-V150)1SI-290'Sr-V188)1PP-357 (FP-V48)" 1AI-220 (1A-V33)1DW-65 (DW-V121)1SW-233 (SW-V142)(B)TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES FUNCTION CCW TO RCP CCW FROM RCP CCW FROM RCP THERMAL BARRIER DEMIN WATER TO PRT SERVICE AIR CONTAINMENT VACUUM RELIEF H2 PURGE MAKEUP ACCUMULATORY FILL FROM RWST N2 TO ACCUMULATORS FIRE WATER STANDPIPE SUPPLY INSTRUMENT AIR SUPPLY DEMIN WATER SUPPLY SERVICE WATER TO NNS PAN COILS EXCESS FLOW CHECK VALVE FOR CTMT VACUUM RELIEF SENSING MAXIMUM ISOLATION TIME (SEC)NA NA NA NA NA NA NA NA NA NA NA NA NA NA 3 HNP F'w/tg,lc<!MINNY 1986 APPLICABLE NOTES NA NA NA NA NA NA NA NA NA NA NA NA~NA 94B (B)EXCESS PLOW CHECK VALVE FOR CTMT VACUUM RELIEF SENSING NA TABLE3-6-1 (PLP)

~l VALVE NO.ENETRATION CP&L NO.(EBASCO)TABLE 3.6-1 (Cont'd)CONTAINMENT ISOLATION VALVES FUNCTION SHNPP@PAL/fc;)m~i MAY$986 MAXIMUM ISOLATION APPLICABLE TIME (SEC)NOTES 94C 95A 95B 98 (B)(B)(B)CB-V2 (CB-V2)EXCESS FLOW CHECK VALVE FOR CTMT VACUUM RELIEF SENSING EXCESS FLOW CHECK VALVE FOR CTMT VACUUM RELIEF SENSING EXCESS FLOW CHECK VALVE FOR CTMT VACUUM RELIEF SENSING CONTAINMENT VACUUM RELIEF NA NA NA NA NA 105 1FP-349 (FP-V46)FIRE WATER SPRINKLER SUPPLY NA NA 11.RELIEF VALVES 7 1CS-10 (CS"R500)CVCS NORMAL LETDOWN NA NA~I TABLE3-6-1 (PLP)

TABLE 3.6-1 (Cont'd)SHNF P 8<Vl l>w)MAY i986)Not subject to Type C leakage tests (2)This valve is not classified as a Containment isolation Valve because fission product release to the environment is prevented by both the closed system inside containment and the system pressure of 45 psig or more following a LOCA.This valve is included in this table because it receives a Phase"B" isolation signal to close it following an accident and because it is the first valve outside the containment.

(3)The provisions of Specification 3.0.4 are not applicable.

(4)This valve may be opened on an intermittent basis under administrative control.(5)The Main Steam Isolation Valves (MSIV's)are included for table completeness.

The requirements of Specification 3.6.3 do not apply because the OPERABILITY requirement for the MSIV's are governed by Specification 3.7.1.5.(6)May be opened only as permitted by Specification 3.6.1.7.(7)Fo<+l~~s votive+g~close.d 5y 5+et'N~4>c,4 i4 iQ loc.R+c.c~

J+o ke aiJ OPENER'DL E isoln+io~pa l ve 4o p, pug poSes<<rnPlin~ce

~;fg gee, ACT'/on/s+nken e<4, TABLE3-6-1 (PLP)

PLANT SYSTEMS PREACTION ANO MULTICYCLE SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION PE00F AHi3 P~PA:I flOPY pFV)$,t~~i trav, 596 3.7.10,2 The Preaction and Multicycle Sprinkler Systems listed on Table 3.7-3 shall be OPERABLE: APPLICABILITY:

Mhenever equipment protected by the Preaction and Multicycle Sprinkler System is required to be OPERABLE.ACTION: a.b.with one or more of the above required Preaction and Multicycle Sprinkler Systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged;for other areas, establish an hourly fire watch patrol.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.10.2 Each of the above required Preaction and Multicycle Sprinkler Systems~~shall be demonstrated OPERABLE: a.At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position,.b.At least once per 12 months by cycling each testable valve in the flow path through at least one compl.ete cycle of full travel, c.At least once per 18 months: 1.By performing a system functional test which includes simulated automatic actuation of the system, and: a)Verifying that the automatic valves in the flow path actuate to their correct positions on a thermal test'ignal, and b)Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.2~By a visual inspection of the dry pipe spray and sprinkler headers to verify their integrity; and By a visual inspection of each nozzle's spray area to verify the spray pattern is not obstructed.

SHEARON HARRIS" UNIT 1 3/4 7"30 Osl TABLE 3.7-3 PRE-ACTION AND MULTICYCLE SPRINKLER SYSTEMS/ZONES SHRPP P P>./)0)&h.)MAY$86 STEM/ZONE DESCRIPTION LOCATION/ELEVATION 1/A 1/B 1/C'/D 1/E 1/F 1/G', t/A 4/B S/A S/B AIRBORNE RADIOACTIVITY REMOVAL UNXTS-lA&1B SPRINKLER (1-C-1-CHFA

&1"C-1-CHFB)

CONTAINMENT FAN COOLERS 1A-SA&1B-SB SPRINKLER (1-C-1-BAL)

CONTAINMENT-FAN COOLERS 1A-SB&1B-SB SPRINKLER (1-C-1-BAL)

PRESSURIZER CABLE CONDUITS (1-C-1-BAL)

I PRESSURIZER CABLE TRAYS (1-C-1-RCP-1B)

ELECTRICAL CABLE PENETRATION AREA-1A SPRXNKLER (1-C-3-EPA)

ELECTRICAL CABLE PENETRATXON AREA-1B SPRXNKLER (1-C"3-EPB)

CONDUIT AND CABLE TRAYS-RC PUMP 1B AREA (1-C-1-RCP-1B)

PRESSURIZER AREA (1-C-1-BAL)

CONTAINMENT SPRAY AND RHR PUMP ROOM 1A SPRINKLER (1-A-1-PA)

CONTAINMENT SPRAY AND RHR PUMP ROOM 1B SPRINKLER (1-A-1-PB)

MISCELLANEOUS PUMP AND EQUXPMENT ROOM-SOUTH (1-A-2-MP)

MISCELLANEOUS PUMP AND EQUIPMENT ROOM-NORTH (1-A-2-MP)

ACCESS CORRIDOR CABLE TRAYS (1-A-3-COR)

MECHANICAL PENETRATION AREA (1-A-3-MP)

CNMW/221 CNMW/236 CNMT""/236 CNMW/236 CNMW/236 CNMW/261 CNMW/261 CNMW/261 CNMW/286 RAB/190 RAB/190 RAB/216 RAB/216 RAB/236 RAB/236 MHE SPRINKLER SYSTEMS LOCATED MITHIN THE CONTAINMENT BUILDING ARE NOT REQUIRED TO BE OPERABLE DURING THE PERFORMANCE OF TYPE A CONTAINMENT LEAKAGE.RATE TESTS.TABLE3-7-3 (PLP)

I Osl f SYSTEM/ZONE DESCRIPTION SHNPP PPQ(Q (~hl LOCATION/ELEVATION 5/c 5/E 6/A 6/B 6/C 6/D 6/E 6/F 7/B 8/A 8/B 10 11/A AUX.FEED WATER PUMPS AND COMPONENT COOLING WATER HEAT EXCHANGER AND PUMPS SPRINKLER (1-A-3-PB)

DECONTAMINATION AREA AND CORRIDOR CABLE TRAY SPRINKLER (l-A-3-COMB, 1"A-3-COME, 1-A-3-COMI)

HVAC CHILLER EQUIPMENT AREA AND CABLE TRAY SPRINKLER (1-A-4-CHLR)

CORRIDOR CABLE TRAY SPRINKLER (1-A-4-COMB

&1-A-4-COME)

CHARCOAL FILTER ROOM lA&CORRIDOR CABLE TRAY SPRINKLER (1-A-4-COMI

&1-A-4-CHFA)

CHARCOAL FILTER ROOM 1B SPRINKLER (1-A-4-CHFB)

ELECTRICAL PENETRATION AREA SA SPRINKLER'(1-A-EPA)

ELECTRICAL PENETRATION AREA SB SPRINKLER (1-A-EPB)CABLE SPREADING ROOMS A&B SPRINKLER (1-A-CSRA&1-A-CSRB)HVAC UNITS E-17&E-18 (12-A-5-CHF)

HVAC EQUIPMENT ROOM SPRINKLER (12-A-6-HV7)

HVAC UNITS E-19&E-20 (12-A-6-CHF-1)

EMERGENCY EXHAUST SYSTEM E-12&E-13 (5-F-3-CHFA

&5-F-3-CHFB)

FUEL POOL COOLING HEAT EXCHANGERS AND PUMPS (5-F-2-FPC)

DIESEL GENERATOR ROOM A 1A-SPRINKLER (1-D-1-DGA-RM)

RAB/236 RAB/236 RAB/261 RAB/261 RAB/261 RAB/261 RAB/261 RAB/261 RAB/286 RAB/286 RAB/305 RAB/305 FHB/261 FHB/236 DGB/261 11/B DIESEL GENERATOR FUEL OIL DAY TANK A ENCLOSURE DGB/261/280 1A-SPRINKLER (1-D-DTA)12/A DIESEL GENERATOR ROOM B 1B-SPRINKLER (1-D-1-DGB-RM)

DGB/261 2/a DIESEL GENERATOR FUEL OIL DAY TANK B ENCLOSURE DGB/261/280 1B-SPRINKLER (1-D-DTB)TABLE3-7-3 (PLP)

Phl (1 SYSTEM/ZONE 3/A13/B DESCR1PTION DIESEL OIL PUMP ROOM A 1A-SPRINKLER (1-0-PA)DIESEL OIL PUMP ROOM B 1B-SPRINKLER (1-0-PB)SHRPP p~S>tc)c,>>

LOCATION/ELEVATION DFOSB/242.25 DFOSB/242.25 TABLE3-7-3 (PLP)

PLANT.SYSTEMS FIRE HOSE STATIONS LIMITING CONQITION FOR OPERATION PROOF AHH t,EV]DP 00?Y P RIVI%)~~i MAY 1986 3.7.10.3 The fire hose stations given in Table 3.7-4 shall be OPERABLE." APPLICABILITY:

Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.ACTION: a~\With one or more of the fire hose stations given in Table 3.7-4 inoperable, provide gated wye(s)on the nearest OPERABLE hose station(s).

One outlet of the wye shall be connected to the standard length of hose provided for the hose station.The second outlet of the wye shal'1 be connected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station.Where it can be demonstrated that the physical~routing of the fire hose would result in a recognizable hazard to operating technicians, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station.Signs shall be mounted above the gated wye(s)to identify the proper hose to use.The above ACTION requirement shall be accomplished within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.b.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.10.3 Each of the fire hose stations given in Table 3.7-4 shall be demonstrated OPERABLE: a.At least once per 31 days, by a visual inspection of the fire hose stations accessible during plant operations to assure all required equipment is at the station.b.At least once per 18 months, by;1.Visual inspection of the stations not accessible during plant operations to assure all required equipment is at the station, 2.Removing the hose for inspection and re-racking, and 3.Inspecting all gaskets and replacing any degraded gaskets in the couplings."Fire hose stations within the containment are required to be operable only~~during refueling and maintenance outages.SHEARON HARRIS-UNIT 1 3/4 7-33 O'T L6AGT'uch PER Bpe4Rc, by~I.PARTIALL)oPmlQ+EmR HOSS~aZOH V ALVE TO VGRlF>YA~Y6 OPERk8lQ(Y f~D No~VJ'BLoCV-@&2, MC', Covuocwug w uosa sy~osr~vic.

wesy a-p.wsssuae or!50~<g oa Kw cwaaw 50 Psiq weave, m<e H~xl4vH Ql gE MA~ad oPGF-ATIQQ PRCWQ)~, VJH'iCHCUGR fs G~t YE'R.,

I I Osl TABLE 3.7-4 FIRE HOSE STATIONS LOCATION CNMT CNMT CNMT CNMT CNMT CNMT CNMT CNMT CNMT CNMT CNMT CNMT RAB~RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB ELEVATION 221 221 221 236 236 236 261 261 261 286 286 286 190 190 216 216 216 216 236 236 236 236 236 236 236 236 236 236 261 261 261 261 261 261 261 261 286 286 286 286 286 286 HOSE RACK NO.221-C"4 221-C-12 221-C-19 236-C-4 236-C-12 236-C-19 261-C-4 261-C-12 261-C-19 286-C-4 286-C-12 286-C-19 190-G-16 190-G-38 216-G"16 216-Fz-27 216-G-38 216-Gy-13 236-Gy"13 236-G-16 236-Fz-27 236-D-27 236-G-38 236-Kz-31 236-C-39 236-Fw-43 236-Jz-43 236-E-15 261-Gy-13 261-E-15 261-G-16 261-D-27 261-Kz-31 261-G-38 261-C-39 261-Fw-42 286-C-15 286.-E-15 286-G"16 286-E-38 286-C-39 286-Jv-41 CNMT-Containment Building FHB-Fuel Handling Building RAB-Reactor Auxiliary Building DGB-Diesel Generator Building SHEARON HARRIS-UNIT 1 3/4 7-34.

S OS1 h X LOCATION RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB RAB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB FHB DGB DGB DGB DGB DFOSB DFOSB ESWIS ESWISS TABLE 3.7-4 (Cont'd)ELEVATION 286 261 261'05 305 305 236 286 286 305 324 324 236 236 261 261 286 286 286 286 286 286 286 216 216 216 236 261 261 261 261 261 261 261 261 261 261 FIRE HOSE STATIONS SHNpp RF+jg)Aa)MAY)gag HOSE RACK NO.286-Fw-42 261-Jz-43 261-Fw-43 305-C-39 305-I-41 305-Fw-43 236-JZ-45 286-JV-45 286-FW-44 305-I-45 324-I-41 324-I-45 236-L-41 236-L-45 261-L-41 261-L-45 286-L-27 286-N-36 286-L-43 286-N-51 286-L-65 286-N-71 286-L-75y 216-L-41 216-L-45 216-L-71 236-L-71 261-N-73 261-M-75y 261-C-2 261-C-4 261-B-1 261-B-2 1-4H NNS~1-4V,.NNS~1-4AJ NNS~1-4AI NNS~~Yard Hydrant CNMT-Containment Building RAB-Reactor Auxiliary Building ESWIS-Emergency Service Water Intake Structure DFOSB-Diesel Fuel Oil Storage Building FHB-Fuel Handling Building DGB-Diesel Generator Building EDWISS-Emergency Service Water Intake Screening Structure SHEARON HARRIS-UNIT 3/4 7-35

ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING PROOF AND REVlHY COPY PP'~~~)SION MAY gag LIMITING CONDITION fOR OPERATION 3.8.3.1 The following electrical buses shall be energized in the specified manner with tie breakers open between redundant buses within the unit: Division A ESF A.C.Buses consisting of: 1.[6900]"volt Bus'1A-SA.2.[480]"volt Bus 1A2"SA.3.[480]-volt Bus 1A3"SA.b.Division B ESF A.C.Buses consisting of: 1..[6900]-volt Bus 1B-SB.2.[480]-volt Bus 1B2"SB.3.[480]"volt Bus 1B3-SB.C.d.'.g.APPLICABI[118]-volt A.C.Vital Bus 1DP-lA-SI energized from its associated inverter connected to 125-volt D.C.Bus DP-lA-SA*,[118]"volt A.C.Vital Bus 1DP-1A-SIII energized from its associated inverter connected to 125-volt D.C.Bus DP-1A-SA",[118]-volt A.C.Vital Bus 1DP-1B-SII energized from its associated inverter connected to 125-volt D.C.Bus DP-lB-SB*,[j18]-volt A.C.Vital Bus 1DP-IJ(-SIV energized from its associated inverter connected to 125-volt D.C.Bus DP-1B-SB",[125]"volt D.C.Bus DP-1A-SA energized from Emergency Battery 1A-SAp4a Q,bed'R/g-gg og/9-gag L125]"volt D.C.Bus DP-1B-SB energized from Emergency Battery 1B-SBa>>&844'C<l8-SQ oX/4-$8.LITY: MODES 1, 2, 3, and 4."Two inverters may be disconnected from their 125-volt D.C.bus for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as.necessary, for the purpose of performing an equalizing char ge on their associated Emergency Battery provided: (1)their vital buses are ener-gized and (2)the vital buses associated with the other Emergency Bat ery are energized from their associated inverters and connected to their asso-ciated 125-volt D.C.bus.SHEARON HARRIS-UNIT 1 3/4 8-14 I ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL E UIPMENT PROTECTIVE DEVICES pr-il)g)~a)

SAY CONTAINMEHT PENETRATION CONDUCTOR OVERCURREHT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3, and 4.ACTION: Mith one or more of the containment penetration conductor overcurrent protective device(s)given in Table 3.8-1 inoperable:

Restore the protective device(s)to OPERABLE status or deenergize the circuit(s) by tripping the associated backup circuit breake~or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoper-able circuit breaker racked out or removed at least once per 7 days thereafter; the provisions of Specification 3..0.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out or removed, or b.Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOMN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.SURVEILLANCE RE UIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices given in Table 3.8-1 shall be demonstrated OPERABLE: At least once per 18 months: 1.By verifying that the{6900-volt]

circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10K of the circuit breakers, and performing the following:

a)A CHANNEL CALIBRATION of the associated protective relays, b)An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and c)For each circuit breaker found inoperable during these functional tests, an additional representative sample of SHEARON HARRIS-UNIT 1 3/4 8-17

ELECTRICAL POWER SYSTEMS REVtStO~MAY-PRDOF AtID IIEVlnbt CDPY ELECTRICAL E UIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SURVEILLANCE RE UIREMENTS Continued 4.8.4.1 (Continued) at least 10K of all the circuit breakers of the inoperable type shall also be functionally tested unti 1 no.more failures are found or all circuit breakers of that type have been functionally tested.By selecting and functionally testing a representative sample of at least 10K of each type of lower voltage circuit breakers.Circuit breakers selected for functional testing shall be selected on a rotating basis.Testing of these circuit breakers shall consist of injecting a current with a value equal to 30(C of the pickup of the long-time delay trip element and 150K of the pickup of the short-time delay trip element, and verifying that the circuit breaker operates within the time delay band width for that current specified by the manufacturer.

The instantaneous element shall be tested by injecting a current equal to 120K of the pickup value of the element and verifying that the circuit breaker trips instantaneously with no intentional time delay.Molded case circuit breaker tes ing shall also follow this procedure except at time del a will be involved.Circuit breakers found nopera e uring unctiona es ing shall be restored to 2.mo(c44~+sc Cub 5aCakCaS~ill..OPERABLE status prior to resuming operation.

For each circuit breaker found inoperable during these functional tests, an*<~>>"~'~additional representative sample of at least 10K of all the iysQdkhucovs circuit breakers of the inoperable type shall also be function-,s>+~,w+o ally tested until no more failures are found or all circuit))Q floe breakers of that type have been functionally tested;and 3.By selecting and functionally testing a representative sample of each type of fuse on a rotating basis.Each representative sample of fuses shall include at least 10K of all fuses of that type..The functional test shall consist of a nondestructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria.Fuses found inoper-able during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation.

For each fuse found inoperable during these functional tests, an additional representative sample of at least 10K of all fuses of that type shall be functionally tested until no more failures are found or all fuses of that type have been functionally tested.b.SHEAR At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

ON HARRIS-UNIT 1 3/4 8" 18 Cl TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP A~V)StO~MAY 95 Item No.E ui ment Descri tion Primar Protection Secondar Protection MOV-2CT-V6SA-1 (Motor)15 A Breaker (Isolation Valve)15 A Breaker MOV-2CT-V6SA-1 (Valve Limit Switch Control)8 A Fuse 15 A Breaker MOV-2CT-V6SA-1 (Valve Limit Switch-ANN) 20 A Fuse 20 A Fuse MOV-2CT-V7SB-1 (Motor)15 A Breaker (Isolation Valve)15 A Breaker MOV-2CT-V7SB-1 (Valve Limit Switch Control)MOV-2CT-V7SB-1 (Valve Limit Switch-ANN) 8 A Fuse 20 A Fuse 15 A Breaker 20 A Fuse MOV-2S-V571SA-1 (Motor)(Isolation Valve)30.A Breaker 30 A Breaker MOV-2SI-V571SA-1 (Valve Limit Switch-IND&ANN)8 A Fuse 15 A Breaker MOV-2SI-V571SA-1 20 A Fuse (Valve Limit Switch-ANN) 20 A Fuse 10 MOV-2SI-V570SB-1 (Motor)(Isolation Valve)K MOV-2SI-V570SB"1 (Valve Limit Switch-IND&ANN)30 A Breaker 8 A Fuse 30 A Breaker 15 A Breaker 12 MOV-2SI-V570SB-1 20 A Fuse (Valve Limit Switch-ANN) 20 A Fuse

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8HNpp APlJ fq]pq MAY gag Item No.E ui ment Descri tion Primar Protection Secondar Protection 13 Containment Fan Cooler AH-37 (1A-NNS)1600 A Switch Gear Breaker 400 A Fuse 14 15 17 Containment Fan Cooler AH-38 (1A-NNS)Containment Fan Cooler AH-39 (1A-NNS)Rod Control Drive Mech Fan E-80 (1A-NNS)Rod Control Drive Mech Fan E-81 (1A-NNS)1600 A Switch Gear Breaker 1600 A Switch Gear Breaker 100 A Breaker 100 A Breaker 400 A Fuse 400 A Fuse 100 A Breaker 100 A Breaker 1&19 Pressurizer Heater Back-Up (Group D)90 A Breaker J1B Crane (Receptacles) 60 A Breaker 60 A Breaker 100 A Fuse21 22 Pressurizer Heater Back-Up (Group D)Pressurizer Heater Back-Up (Group D)Pressurizer Power Relief Isolation Valve MOV-1RC-V526SN-1 90 A Breaker 90 A Breaker 15 A Breaker 100 A Fuse 100 A Fuse 15 A Breaker 23 Pressurizer Power Relief Isolation Valve MOV1RC-V527SN-1 15 A Breaker 15 A Breaker 24 Full Length Rod Control (Control Bank A.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 25'Full Length Rod Control (Control Bank A.Group 2)Gripper Lift Coil 10 A Fuse 50-A Fuse 10 A Fuse 50 A Fuse Cl TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPp RPVjR,'A<J MAY]986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 26 27 Full Length Rod Control (Control Bank A.Group 2)Gripper 10 A Fuse Lift Coil 50 A Fuse Full length Rod Control (Control Bank A.Group 2)10 A Puse 50 A Fuse Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 28 Pull Length Rod Control (Control Bank C.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 29 Full Length Rod Control (Control Bank C.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 30 Pull Lenght Rod Control (Control Bank C.Group 2)Gripper.Lift Coil 10 A Fuse 50.A Fuse 10 A Fuse 50 A Fuse 31 Pull Lenth Rod Control (Control Bank C.Group 2)Gripper..Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 32 Full Length Rod Control (Shut Down Bank A.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Puse 50 A Fuse 33 Full Length Rod Control (Shut Down Bank A.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse TABLE 3.8-1.CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHgpp RF~/)Q tnt)MAY]ggg Item No.E ui ment Descri tion Pzimar Protection Secondar Protection 34 Pull Length Rod Control (Shut Down Bank A Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 35 Full Length Rod Control (Shut Down Bank A.Group 2)Gripper Lift Coil 10 A Fuse 50 A Puse 10 A Fuse 50 A Fuse 36 Full Length Rod Control (Shut Down Bank DE Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 37 Full Length Rod (Shut Down Bank Gripper Lift Coil Control D.Group 1)10 A Fuse 50.A Puse 10 A Fuse 50 A Fuse 38 Pull Length Rod Control (Shut Down Bank D.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 39 Full Length Rod Control (Control Bank A.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 40 Full Length Rod Control (Control, Bank A, Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP RF>jig t~<<MAY$86 Item No.E ui ment Descri tion Primar Protection

.Secondar Protection Full Length Rod Control (Control Bank A.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 42 Pull Length Rod Control (Control Bank A.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 43 Full Length Rod Control (Control Bank C~Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse Full Length Rod Control (Control Bank C.Group 1)Gripper Lift Coil 10 A Fuse 50'Puse 10 A Fuse 50 A Fuse 45 Pull Length Rod Control (Control Bank C.Group 1)Gripper.Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 46 Full Length Rod Control (Control Bank CD Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 47 Full Length Rod Control (Shutdown Bank A.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP PP~I)C,'!~kl MAY$86 Item No.E ui ment Descri tion Primar Protection Secondar Protection 48 Full Length Rod Control (Shutdown Bank A.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 49 Full Length Rod Contxol (Shutdown Bank A.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Puse 50 Pull Length Rod Control-(Shutdown Mn~.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 51 Pull Length Rod Control (Shutdown Bank C.Group Gripper Lift Coil 1)10 A Fuse 5Q A Fuse 10 A Fuse 50 A Fuse 52 Pull Length Rod Control (Shutdown Bank C.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 53 Pull Length Rod Control (Shutdown Bank CD Group 1)Gripper Lift Coil 10 A Puse 50 A Fuse 10 A Fuse 50.A Fuse 54 Pull Length Rod Control (Shutdown Bank C.Group 1)Gxipper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8HNpp REM)Stn~i MAY$86 Item No.E ui ment Descri tion Primar Protection Secondar Protection 55 Full length Rod Control (Shutdown Bank D.Group 1)Gripper 10 A Fuse Lift Coil 50 A Fuse 10 A Fuse 50 A Fuse 56 57 Containment Fan Cooler 225 A Breaker AH-3 (1B-SA)Containment Fan Cooler 225 A Breaker AH-3 (1B-SA)1600 A Breaker 1600 A Breaker 58 60 Containment Fan Cooler 225 A Breaker AH-3 (1A-SA)Containment Fan Cooler 225 A Breaker AH-3 (1A-SA)Containment Fan Cooler 225 A Breaker AH-4 (1B-SB)1600 A Breaker 1600 A Breaker 1600 A Breaker 62 Containment Fan Cooler 225 AH-4 (1B-SB)/Containment Fan Cooler 225 AH-4 (1A-SB)A Breaker A Breaker 1600 A Breaker 1600 A Breaker 63 Containment Fan Cooler 225 A Breaker AH-4 (lA-SB)Full length Rod.Control (Control Bank B.Group 1)1600 A Breaker Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 65 Full length Rod Control (Control Bank B.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse TABLE 3.8-1 CONTAINMEFZ:.!PENETRATION..CONDUCTOR OVERCURREMi97ROTECEIVBJINPfXCES 8HNPp p f-"(Cfc i~<<MAY gag Item No.E ui ment Descri tion Primar Protection Secondar Protection 66 67 Full Length Rod Control (Control Bank B.Group 1)Gripper Lift Coil Full length Rod Control (Control Bank B.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 68 Full Length Rod Control (Control Bank D.Group 1)Gripper 10 A Fuse Lift Coil 50 A Fuse 10 A Fuse.50 A Fuse 69 Pull Length Rod Control (Control Bank D.Group 1)Gripper Lift Coil'0 A Puse 50 A Fuse 10 A Fuse 50 A Fuse 70 Full Length Rod Control (Shutdown Bank B.Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 71 Full Length Rod Control (Shutdown Bank B.Group 1)r Gripper 10 A Fuse Lift Coil 50 A Fuse 10 A Puse 50 A Fuse 72 Pull Length Rod Control (Shutdown Bank BE Group 1)Gripper.Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 73 Full Length Rod Control (Shutdown Bank BE Group 1)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse

'TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES Item No.E ui ment Descri tion Primar Protection SHNPP P~lj)g t~<~MAY 586 Secondar Protection 74 Full Length Rod Control (Control Bank B.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 75 Full Length Rod Control (Control Bank B.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 76 Full Length Rod Control (Control Bank B.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 77 Pull Length Rod Control (Control Bank B.Group 2)Gripper Lift Coil 10 A Fuse 50.A Fuse 10 A Fuse 50 A Fuse 78 Pull Length Rod Control (Control Bank D.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse.10 A Fuse 50 A Fuse 79 Full Length Rod Control (Control Bank D.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 80 Pull Length Rod Control (Shutdown Bank BE Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP pp~s~pir~i MAY$86 Item No.E ui ment Descri tion Primar Protection Secondar Protection 81 Full Length Rod Control (Shutdown Bank B.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 82 Full Length Rod Control (Shutdown Bank B.Group 2)Gripper Lift Coil, 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 83 Full Length Rod Control (Shutdown Bank B.Group 2)Gripper Lift Coil 10 A Fuse 50 A Fuse 10 A Fuse 50 A Fuse 85 86 Reactor Coolant Pump (1A-SN)Reactor Coolant Pump (1A-SN)Lighting Panel (LP-105)Relay Trips.Feeder Breaker Relay Trips Feeder Breaker 70 A Breaker Relay Trips Upstream Breaker Relay Trips Upstream Breaker 150 A Breaker 87 88 89 90'ressurizer Heater Back-Up (Group-A)90 A Breaker Lighting Panel (LP-106)50 A Breaker Lighting Panel" (LP-101)60 A Breaker Lighting Panel (LP-102)60 A Breaker 50 A Breaker 125 A Breaker 125 A Breaker 100 A Fuse91 92 93 Pressurizer heater Back-Up (Group-A)Pressurizer Heater Back"Up (Group"A)1 Pressurizer Heater Back-Up (Group-A)Elevator Disc Switch 90 A Breaker 90'Breaker ,.90 A Breaker 100 A Breaker 100 A Fuse 100 A Fuse 100 A Fuse 100 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHg pp p~~l f 0 1M',(MAY$86 Item No.E ui ment Descri tion Primar Protection Secondar Protection 95 Po~er Receptacles 60 A Breaker 1-2 6 1-6 60 A Breaker 96 Power Receptacles 1-9&1-13 60 A Breaker 60 A Breaker 97 98 Power Receptacles 1-10 6 1-,14 Reactor Coolant Pump 1A"SN Oil BRG Lift Pump 60 A Breaker 30 A Breaker 60 A Breaker 30 A Breaker 99 Disk Switch for 5-Ton 50 A Breaker Monorail 50 A Breaker 100 101 102 103 Pressurizer Heater Back-Up (Group-A)Pressurizer Heater Back-Up (Group-A)Pressurizer Heater Back-Up (Group-A)Pressurizer'Heater Back-Up (Group-A)90 A Breaker 90 A Breaker 90 A Breaker 90 A Breaker 100 A Fuse 100 A Fuse 100 A Fuse 100 A Fuse 104'ower Receptacles 1-1 6 1-5 60 A Breaker 60 A Breaker 105 Power Receptacles 1-17 6 1-74 60 A Breaker 60 A Breaker 106 Power Receptacles 1-18&1-75 60 A Breaker 60 A Breaker 107 Rod Position Indication 50 A Breaker Distribution Panel 100 A Breaker~108 109 Pressurizer heater Control Group-C Pressurizer Heater Control Group-C 90 A Breaker 90 A Breaker 100 A Fuse 100 A Fuse I

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHg pp RF.>/)q!c g!MAY Item No.E ui ment Descri tion.Primar Protection Secondar Protection 110 112 Pressurizer Heater Control Group-C Pressurizer Heater'ontrol Group"C Pressurizer Heater Control Group-C 90 A Breaker 90 A Breaker 90 A Breaker 100 A Fuse 100 A Fuse 100 A Fuse 113 Reactor Coolant Drain 50 A Breaker Tank Pump-1A 50 A Breaker 114 115 Pressurizer Heater Control Group"C Pressurizer Heater Control Group-C 90 A Breaker 90 A Breaker 100 A Fuse 100 A Fuse 116117 Power Receptacles 1-76 60 A Breaker Containment Evacuation 3 A Fuse Horn 60 A Breaker 20 A Fuse 118 119 AOV-1RC-P527SN-1 5 A Fuse Containment Circular 225 A Breaker Bridge Crane 5 A Fuse 225 A Breaker 120 IRVH Cable Bridge Hoist 15 A Breaker 15 A Breaker 121 AOV-1RC-P528SN-1 (PCV-445-B) 5 A Fuse 5 A Fuse 122 123 AOV-1RC-P525SN-1 AOV-1RC-P525SN-1 3 A Fuse 6 A Fuse 3 A Fuse.20 A Breaker 124 125 AOV-1CS-L501SN-1 (LCV-459)3 A Fuse AOV-2RC-V504SN-1 (8032)3 A Fuse.3 A Fuse 3 A Fuse 126'OV-2CS-V519SN-1 (8141A)3 A Fuse I 3 A Fuse TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP REVtS)GN MAY f986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 127 AOV-2CS-V501SN-1 (8145)3 A Fuse 128 AOV-2CS-V502SN-1 (8146)3 A Fuse 129 130 AOV-2CS-V503SN-1 (8147)3 A Fuse AOV-7CS-D507SN"1 (8168A)3 A Fuse 131 MOV-2SI-V537SA-1 (8808A)Pos.SW.ANN 3 A Fuse 132 133 AOV-2SI-V532SN-1 (8875A)3 A Fuse AOV-2SI-V534SN-1 (8875C)3 A Fuse 134 AOV-2SI-V534SN-1 (8877A)3 A Fuse 135 AOV-2SI-V540SN-1 (8877C)3 A Fuse 136 AOV-2SI-V551SN-1 (8878A).3 A Fuse 137 AOV-2SI-V553SN-1

'(8878C)3 A Fuse 138 AOV-2SI-V541SN-1 (8879A)3 A Fuse 139 AOV-2SI-V543SN-1 (8879C)3 A Puse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 15 A Breaker 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3'A Fuse 140 Integrated Head 20 A Breaker Cooling, Fan E-.8 (1A-NNS)141 Integrated Head 20 A Breaker Cooling Fan E"81 (1A-NNS)142 AOV-2BD-F6SN-1 6 A Fuse (PCV-8400A) 20 A Breaker 20 A Breaker 15 A Breaker 143 Damper (CV-D9-1)144 Damper (CV-D13"1) 145 AOV-1'CS-L500SN-1 (LCV-460)6 A Fuse 6 A Fuse 3 A Fuse 15 A Breaker 15 A Breaker 3 A Fuse 146 AOV-6WL-D640SN-1 (7127)6 A Fuse 147 AOV-6WL-D649SN-1 (7144)6 A Fuse ,6 A Puse 6 A Fuse p

E TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE.

DEVICES SHNPp RFV)S!~~>MAY Sag Item No.E ui ment Descri tion Primar Protection

.Secondar Protection 150 Con.Rod Drive Mech.Fan E-80 (1A"NNS)6 A Fuse 148 AOV-6WL-D648SN-1 (7143)6 A Fuse 149 AOV-6WL-D647SN-1 (7141)6 A Fuse 6 A Fuse 6 A Fuse 15 A Breaker 151 Con.Rod Drive Mech.Fan E-81 (1A-NNS)6 A Fuse 15 A Breaker 152 Reactor Coolant Pump (1A-GN)Space Heater 15 A Breaker 30 A Breaker 153 Inst.Rack Cl-Rl 20 A Breaker 154 AH-37 (lA-NNS)Motor 15 A Breaker Space Heater 155 AH-38 (1A-NNS)Motor 15 A Breaker Space Heater 156 AH-39 (1A-NNS)Motor 15'Breaker Space Heater 20 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 157 158 Elevator Equipment Room Fan (E-3)(1X-NNS)SVT-SP-V334-1 (Rad.Mon.Sampling Valves)20 A Breaker 15 A Breaker 20 A Breaker 15, A Breaker 159 160 161 162 163 SV7-SP-V318-1 SV7-SP-V320-1 Containment Atmo.Rad.Mon.Valve (7SP-V322-1)

Containment Atmo.Rad.Mon.Valve (7SP-V324-1)

Containment Atmo.Rad.Mon.Valve (7SP-V326-1) 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker A 164 Containment Atmo.Rad.15 A Breaker Mon.Valve (7SP"V328-1) 15 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHf']Pp R~V~Str~!

MAY Sag Item No.E ui ment Descri tion Primar Protection Secondar Protection 165 Containment Atmo.Rad.15 A Breaker 15 A Breaker Mon.Valve (7SP-330-1) 166 167 168 169 AOV2RC-D528SA-1 (Limit Switch)8 A Fuse MOV-2CS-V516SA-1 (8112)8 A Fuse (Limit Switch)Containment Atmo.Rad.15 A Breaker Mon.Valve (7SP-332-1)

AOV-2RC-D528-SA-1 (8047)5 A Fuse 15 A Breaker 5 A Fuse 15 A Breaker 15 A Breaker 3.70 AOV-2CS-V511SA-1 (Sol.Valve&Limit Switch)5 A Fuse 5 A Fuse 171 AOV-2CS-V511SA-1

~(Limit Switch)172 AOV-2CS-V511SA-1 (Limit Switch)173 AOV-2CS-V511SA-1 (Sol~Valve)(8149A)174 AOV-2CS-V512SA-1 (Sol.Valve 6 Limit Switch)20 A Fuse 8 A Fuse 5 A Fuse 5 A Fuse 20 A Fuse 15 A Breaker 5 A Fuse 5 A Fuse 175 AOV-2CS-V512SA-1.(Limit Switch)176 AOV-2CS-V512SA-1 (Limit Switch)20 A Fuse 8 A Fuse 20 A Fuse 15 A Breaker 177 AOV-2CS-V512SA-1 (Sol.Valve)5 A Fuse 5 A Fuse 178 AOV-2CS-V513SA-1 (Sol Valve h Limit Switch)179 AOV-2CS-V513SA-1

'(Limit Switch)3 A Fuse 20 A Fuse 3 A Fuse 20 A Fuse TABLE 3.8-1~-.CQNTAINMEKI';PENETRATION CONDUCTOR~>~EXNKRCURREZRG'PROXECZI VE~'DEVICES SHNPP RFVISi~~~MAY 1986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 180 AOV-2CS-V513SA-1 (Limit Switch'8 A Fuse 15 A Breaker 181 AOV-2CS-V513SA-1 (Sol.Valve)3 A Fuse 3 A Fuse 182 MOV-2SI-V537SA-1 (8808A)(Limit Switch)8 A Fuse 15 A Breaker183 184 185 MOV-2SI-.V535SA-1 (8808C)(Limit Switch)MOV-2SI-V555SA-1 (8871)(Sol.Valve)MOV-2SI-V555SA;1

" (Limit Switch)(8871)MOV-2SI-V537SA-1 (8088A)(Limit Switch)8 A Fuse 3 A Fuse 8 A Fuse 20 A Fuse 15 A Breaker 3 A Fuse 15 A Breaker 20 A Fuse 187 188 189 MOV-2SI-V537SA-1 20 A Fuse (Stem.Oper.Pos.Switch)MOV-2SI-V535SA-1 20 A Fuse (Stem.Oper.Pos.Switch)AOV-2WL-L600SA-1 3 A Fuse (LCV-1003)(Sol.Valve&Limit Switch)20 A Fuse 20 A Fuse 3 A Fuse 190 AOV-2WL-L600SA-.

1'LCV-1003)..(Limit Switch)8 A Fuse 15 A Breaker 191 192 AOV-2WG-D590SN-1 (7126)3 A Fuse (Limit Switch&Sol~Valve)AOV2WG-D590-SN-1 (7126)8 A Fuse (Limit Switch)3 A.Fuse 15 A Breaker 193194 AOV-2SP-V300SA-1 AOV-2SP-V301SA-1 6 A Fuse 6 A Fuse 20 A Breaker 20 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP pP)l jQ0~gl MAY$86 Item No.E ui ment Des'cri tion Primar Protection Secondar Protection 195 MOV-'2CC-V184SA-1 (9481)8 A Fuse (Limit Switch)196 MOV-2CC-V191SA-1 (9483)8 A Fuse (Limit Switch)15 A Breaker 15 A Breaker 197 Service Water Containment Fan Cooler Isol.Valve (2SW-B89SA-1) 198 Hydrogen Purge (AOV-2CM-B5SA-1) 6 A Fuse 6 A Fuse 6 A Fuse 20 A Breaker 199 RA-1CR-3561A-SA 200 RA-1CR-3561C-SA 201 SV-2RC-V281SA-1 0.6 A Fuse.0.6 A Fuse&A Fuse 202 Reactor Support Cooling-'15 A Breaker Fan S-4 (1A-SA)Htr.15 A Breaker 15 A Breaker 20 A Breaker 15 A Breaker 205~Containment Atm.Rad.Mon.Valve 2SP-V405SA-1 203 SV-2RC-V283SA-1 204 SV-2RC-V284SA-1 6 A Fuse 6 A Fuse 6 A Fuse 20 A Breaker 20 A Breaker 20 A Breaker 206 207 AOV-2SP-V21SA-1 Containment Fan Cooler AH-2 Damper CV-D3SA-1 Position Switch 6 A Fuse 6 A Fuse 20 A Breaker 20 A Breaker 208 Cont.Pre-entry Purge 8 A Fuse Discharge Valve 2CP-B7SA-1

-15 A Breaker 209 Cont.Pre-entry Purge Inlet Valve 2CP-B3SA-1 210 Cont.Pre-entry Purge Inlet Valve 2CP-B3SA-1 6 A Fuse 8 A Fuse 20 A Breaker.15 A Breaker

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP 8~"Via'AN MAY 3986 Item No.E ui ment Descri tion Prima Protection Secondar Protection 211 212 AOV-2CP-B5-SA-1 Cont.H>Analyzer System Valve 2SP-V316SA-1 6 A Fuse 20 A Breaker 20 A Breaker 20 A Breaker 213 Cont.H>Analyzer System 9'alve 2SP-V386SA-1 20 A Breaker 20 A Breaker 214 Containment Fan Cooler AH-3 Damper CV-D5SA-1 Position Switch 6 A Fuse 20 A Breaker 215216 217 218 Cont.H>Analyzer System Valve 2SP-V388SA-1 Cont.H>Analyzer System Valve 2SP-V390SA-1 Cont.H>Analyzer System Valve 2SP-V392SA-1 Cont.H>Analyzer System Valve 2SP-V394SA-1 20 A Breaker 20 A Breaker 20 A Breaker 20 A Breaker 20 A Breaker 20 A Breaker 20 A Breaker 20 A Breaker 219 220 221 222 Cont.Fan Cooler AH-2 (1A-SA)Space Heate~Cont.Fan Cooler AH-2 (1B-SA)Space Heater Cont.Fan Cooler AH-3 (1A-SA).Space Heater Cont.Fan Cooler AH-3 (1B"SA)Space Heater 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker I M TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP R~>>t".-'r~l MAY 1986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 223 Primary Shield Cooling 15 A Breaker Fan S-2 (lA-SA)Heater 15 A Breaker 224 225 226 Hydrogen Recombiner Reactor Support Cooling Fan S-4 (lA-SA)MOV-1RH-V501SA-1 (8701B)(Isolation Valve)125 A Breaker 100 A Breaker 15 A Breaker 125 A Breaker 100 A Breaker 15 A Breaker 227 228ns MOV-2SI"V535SA-1 (8808C)(Accumulator"C" Discharge Valve).40 A Breaker MOV-2CS-V516SA-1 (8812)(RCP Seal Mater Return Isolation Valve)-15 A Breaker MOV-.2SI-V537SA-l 40 A Breaker (8808A)(Accumulator"A" Discharge Valve)40 A Breaker 40 A Breaker 15 A Breaker 230 MOV-1RH-V503SA-1 (8701A)(RHRS Inlet Isolation Valve)15 A Breaker 15 A Breaker 231.MOV-2CC"V184SA-1 (9481)(RCP Oil Heat Exchanger Isolation Valve)15 A Breaker 15 A Breaker 232 MOV-2CC-V191SA-1 (9483)(RCP Thermal Barrier Isolation Valve)15 A Breaker 15 A Breaker.233 MOV-2MD"V36SA-1 (Cont.Sump Isolation Valve)15 A Breaker 15 A Breaker 234 Primary Shield Cooling 100 A Breaker Fan S-2 (1A-SA)100 A Breakerns Reactor Coolant Pump (1B-SN)Relay Trips Feeder Brk.Relay Trips Upstream Brk, TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8 HNP F'F>>..~'ISIQW)

MAY 1986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 236 Reactor Coolant Pump (1B-SN)Relay Trips Feeder Brk.Relay Trips Upstream Brk.238 Primary Shield Cooling Fan S-2 (1B-SB)100 A Breaker 237 Hydrogen Recombiner"B" 125 A Breaker 125 A Breaker 100 A Breaker 239 240 241 Reactor Support Cooling Fan S-4 (1B-SB)MOV-2SI-V536SB-1 (8088B)(Accumulator"B" Discharge VaLve)MOV-1RH-V502SB-1 (8702A)(RHR Inlet Isolation Valve)100 A Breaker 40 A Breaker 15 A Breaker 100 A Breaker 40 A Breaker, 15 A Breaker 242 MOV-1RH-500SB-l (8702B)(RHR Inlet Isolation Valve)15 A Breaker 15 A Breaker 243 Valve 2BD-V2SB-1 245.246 248 249 250 Valve 2BD-V8SB-1 Valve 2BD-P6SB-1 Valve 2BD-V2SB-1 Valve 2BD-V5SB-1 Valve 2BD-V8SB-1 Valve 2BD-P8SB-1 251 Primary Shield Cooling Fan S-2 (1B-SB)Heater 244 Valve 2BD-V5SB-1 6i'A Fuse/6 A Fuse 6 A Fuse 6 A Fuse 8 A Fuse 8'A Fuse 8 A Fuse 6 A Fuse 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 252 Reactor Support-Cooling Fan S-4 (1B-SB)Heater.15 A Breaker 15 A Breaker 6

TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE-DEVICES SHNPP pt:,qg~~~~MAY 1986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 253 Valve 2BD-P7SB-1 6 A Fuse 254 Cont.Fan AH-1 (1A"SB)15 A Breaker Space Heater 15 A Breaker 15 A Breaker 255 Cont.Fan AH"1 (1B-SB)15 A Breaker Space Heater 15 A Breaker 256 257 Cont.Fan Cooler AH-4 (1$-SB)Space Heater A Cont.Fan Cooler AH-4 (1B"SB)Space Heater 15 A Breaker 15 A Breaker 15 A Breaker 1'5 A Breaker 258 Valve 1CS-V510SB-1 259 AOV-1CS-V509SB-1 (8154)3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse260 261 MOV-2SI-V536SB-1 (8808B)MOV-2SI-V536SB-1 (8808B).8 A Fuse 20 A Fuse 15 A Breaker 20 A Fuse 262 MOV-2SI-V536SB-1 20 A Fuse (Steam Oper.Pos.Switch)20 A Fuse 263 AOV-2SP-V308SB-1 264 AOV-2SP-V309SB-1 265 266 Valve 2SP-V90SB-1 Valve 2SP-V85SB-1 270 Valve 2SP-V81SB-1 271 Valve 2SP-V1SB-1 267 Valve 2SP-V80SB-1 268 Valve 2SP-V91SB-1 269'alve 2SP-V86SB-1 6 A Fuse 6 A Fuse 6 A Fuse 6 A Fusd 6 A Fuse 6 A Fuse 6 A Fuse 6 A Fuse 6 A Fuse 15 A Breaker 15 A Breaker 15 A Breaker 15'A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker Cll TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHNPP PU~!j~~~Pib)MAY 1986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 272 Valve 2SP-V11SB-1 6 A Fuse 15 A Breaker 273 Valve 2SP-VlllSB-1 6 A Fuse 15 A Breaker 274 275'276 279 280 Valve 2SP-V113SB-1 Valve 2SP"V114SB-1 Valve 2SP-V115SB-1 Cont.Fan Cooler AH-1 Damper CV-D1SB-1 Motor Cont.Fan Cooler AH-1 Damper CV-D1SB-1 Motor 6 A Fuse 6 A Fuse 6 A Fuse 5 A Fuse 6 A Fuse 15 A Breaker 15 A Breaker 15 A Breaker 20 A Breaker 20 A Breaker 281 Cont.Fan Cooler Damper CV-DlSB-1 Pose Swo 282 Cont.Fan Cooler Damper CV-D2SB-1.Pos.Sw.AH-1 6 A Fuse AH-1-3 A Fuse 20 A Breaker 3 A Fuse 285.Cont.Fan Cooler Damper CV-D7SB-1 286 Cont.Fan Cooler Damper CV-D7SB-1 AH-4 Motor AH-4 Motor 5 A Fuse 6 A Fuse 20 A Breaker 20 A Breaker 287 Cont.Fan Cooler AH-.4 Damper CV-D7SB-1 Pos.Sw.6 A Fuse 20 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHIPS p MAY 1986 Item No.E ui ment Descri tion Primar Protection Secondar Protection 288'ont.Fan Cooler AH-4 3 A Fuse 3 A Fuse Damper CV-D8SB-1 Pos.Sv.289 RA-1CR-356 1B-SB 290 RA-1CR-356 1D-SB 291 Valve 2SP-V90SB-1 292 Valve 2SP-V91SB-1 293 Valve 2SP-V85SB-1 294 Valve 2SP-V86SB-1 295 Valve 2SP-V80SB-1 296 Valve 2SP-V81SB-1 297 Valve 2SP-V11SB-1 302 SV-2RC-V280SB-1 303 SV-2RC-V282SB-1 304 SV-2RC-V285SB-1 305 Cont H>Analyzer System'Valve 2SP-V317SB 298 Valve 2SP-V1SB-1 299 Valve 2SP-VlllSB-1 300 Valve 2SP-V114SB-1 301 Valve 2SP-V113SB-1 0.6 A Fuse 0.6 A Fuse 8 A Fuse 8 A Fuse 8 A Fuse 8 A Fuse 8 A Fuse 8 A Fuse-8 A Fuse 8 A Fuse 8 A Fuse 8 A Fuse'8 A Fuse 6 A Fuse 6 A Fuse 6 A Fuse 20 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 20 A Breaker 20 A Breaker 306 Cont.H>Analyzer System'Valve 2SP-V387SB 20 A Breaker 20 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR-:,OVERCURREKZ.PROTECTIVE DEVICES 8HNt p p C~'L>j i~~'g cy j MAY$986 307 Cont.H>Analyzer System Valve 2SP-V389SB 20 A Breaker Item No.E ui ment Descri tion Primar Protection Secondar Protection 20 A Breaker 308 Cont.H>Analyzer System Valve 2SP-V391SB 20 A Breaker 20 A Breaker 309 Cont.H>Analyzer System Valve 2SP-V393SB 20 A Breaker 20 A Breaker 310 Cont.H>Analyzer System Valve 2SP-V395SB 311 Valve 2SP-V22SB-1 312 Valve 2SP-V408SB-1 313 Valve.2SP-V406SB-1 314 AOV-1RC-P529SN-1 (PCV-444B) 20 A Breaker6 A Fuse 6 A Fuse'A Fuse 5'A Fuse C 20 A Breaker 15 A Breaker 15 A Breaker 15 A Breaker 5 A Fuse 315 AOV-20S-W500SN-1 (8143)3 A Fuse 3 A Fuse 316 317 Incore Inst.Drive A 25 A Breaker Unit-TB Incore Inst.Drive B 25 A Breaker Unit-TB 25 A Breaker 25 A Breaker 318 Incore Inst.Drive C 25 A Breaker Unit-TB 25 A Breaker 3.19~Incore Inst.Drive D 25 A Breaker--Unit-TB 25 A Breaker 320 Incore Inst.Drive E Unit-TB 25 A Breaker 25 A Breaker 321 Incore Inst..Drive A Leak'Detection C 25 A Breaker 25 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8H happ PP-~l j~-1p p,]MAY fggg Item No.E ui ment Descri tion Primar Protection Secondar Protection 322 323 Incore Inst.Drive A 25 A Breaker Path Insertion Incore Inst.Drive A 25 A Breaker Leak Detection 25 A Breaker 25 A Breaker 324 Encore Inst.Drive B Path Insertion 25 A Breaker 25 A Breaker 325 326 Incore Inst.Drive C Path Insertion Incore Inst..Drive D Path Insertion 25 A Breaker 25 A Breaker 25 A Breaker 25 A Breaker 327 Incore Inst.Drive E Path Insertion 25 A Breaker 25 A Preaker329 330 331 332 MOV-2SI-V536SB-1 (8808B)Position Switch (ANN)AOV-2SI-V533SN-1 (8875B)AOV-2SI-'539SN-1 (8877B)AOV-2SI-V552S¹1 (8878B)AOV-2SI-V542SN-1 (8878B)3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 15 A Breaker 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 333 Integrated Head 6 A Fuse Cooling Fan E-80 (1B-NNS)15 A Breaker 334 335 Integrated Head 6 A Fuse Cooling Fan E-81 (1B-NNS)2SI-V631-SN-1

.3 A Fuse Va1ve Position Switch 15 A Breaker 3 A Fuseass 337 2SI-V631-.SN-1 Valve Position Switch Damper CV-D20-1 3 A Fuse 6 A Fuse 3 A Fuse 15 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8HMPP MAY]gag Item No.338 E ui ment Descri tion Primar Protection 2SI-V632-SN-1 3 A Fuse (Valve Position Switch)Secondar Protection 3 A Fuse 339 340 2SI-V636-SN-1 (Valve Position Switch)2SI-V634-SN-1 (Valve Position Switch)3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 341 342 2SI-V637-SN-1 (Valve Position Switch)2SI-V635-SN-1 (Valve Position Switch)3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 343 2SI-V638-SN-1 3 A Fuse (Valve Position Switch)" 3 A Fuse, 2SI-V626-SN-1'Valve Position Switch)2SI-V628-SN-1 (Valve Position Switch)3 A Fuse 3 A'use 3 A Fuse 3 A Fuse 346 2SI-V629-SN-1 (Valve Position Switch)3 A Fuse 3 A Fuse 347 2SI-V630-SN-1 (Valve Position Switch)3 A Fuse 3 A Fuse'348 349.350 351 RCP-IB-SN Space Heater 15 Integrated Head 20 Cooling Fan E-80 (1B.-NNS)Integrated Head 20 Cooling Fan E-81 (1B-NNS)AH-37 (1B-NNS).Motor 15 Space Heater A Breaker A Breaker A Breaker A Breaker 30 A Breaker 20 A Breaker 20 A Breaker 15 A Breaker 352os'H"38 (1B-NNS)Motor Space Heater AH-39 (1B-NNS).Motor Space Heater 15 A Breaker'15 A Breaker 15 A Breaker 15 A Breaker-26" TABLE 3.8"1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SHE pp Rg;,~~i~~pW p)MAY]ggg 354 CFC-AH-37 (1B-NNS)355 CFC-AH-38 (1B-NNS)356 CFC-AH-39 (1B-NNS)1600 A Breaker 1600 A Breaker 1600 A Breaker 357 CRDM Fan E-80 (1B-NNS)100 A Breaker 358 CRDM Fan E-81 (1B-NNS)100 A Breaker 359 Reactor Coolant.Drain'0 A Breaker Tank Pump 1B Item No.E ui ment Descri tion Primar Protection Secondar Protection 400 A Fuse 400 A Fuse 400 A Fuse 100 A Breaker 100 A Breaker 50 A Breaker 360 Containment Building Sump Pump 1B-NNS 361 Incore Instrument Drive Assemblies

-50 A Breaker 15 A Breaker 50 A Breaker 15 A Breaker 362 MOV-1RC-V528SN-1 (8000C)363 Lighting Panel LP-104 364 Lighting Panel LP-107 (N/E)365 Lighting Panel LP-103 366 Lighting Panel LP-123 367 Pressurizer Heater Back-up Group"B" 368 Pressurizer Heater Back-up Group"B" 369 Pressurizer Heater Back-up Group"B" 370 Pressurizer Heater Back-up Group"B" Po~er Receptacles f1-12, 1-16 15 A Breaker 70 A Breaker 50 A Breaker 50 A Breaker 60 A Breaker 90 A Breaker 90 A Breaker 90 A Breaker 90 A Breaker 60 A Breaker 15 A Breaker 150 A-Breaker 50 A Breaker IC0 440'Breaker 125 A Breaker.100 A Fuse 100 A Fuse 100 A Fuse 100 A Fuse 60 A Breaker TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES 8HNPP p g'~t~~~~MAY<OSS Item No.E ui ment Descri tion Primar Protection Secondar Protection 372 Power Receptacles 60 A Breaker 60 A Breaker$1-3, 1-7 373 Power Receptacles gl"4, 1-8 g~R<a>9 374 RCP-1B-SN Oil Bridge Lift Pump 375 Pressurizer Heater Back-up Group"B" 376 Pressurizer Heater Back-up Group"B" 377 Pressurizer Heater Back-up Group"B" 60 A Breaker 30 A Breaker 90 A Breaker 90 A Breaker 90 A Breaker 60 A Breaker 30 A Breaker 100 A Fuse 100 A Fuse 100 A Fuse 378~vs 380 381 Pressurizer Heater Back-up Group".B" Digital Rod Position Indication Cab"B" 120 V AC Supply Power Receptacles f'1-11, 1-15 Power Receptacles

$1-77, 1-78 90 A Breaker 50 A.Breaker 60 A Breaker 60 A Breaker 100 A Fuse 100 A Breaker 60 A Breaker 60 A Breaker 382 Stud Tensioner Hoist Motor (CRDM Terminal Box B1263)15 A Breaker 15 A Breaker 383 RCP-1C-SN 384 RCP-1C-SN Relay Trips Feeder Bkr.Relay Trips Feeder Bkr.Relay Trips Upstream Brk.Relay Trips Upstream Brk.385 Containment Fan Cooler 225 A Breaker 1600 A Breaker AH-2 (1A-SA)386 Containment Fan AH-2 (lA-SA)Cooler 225 A Breaker 1600 A Breaker CP TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVEvDEVXCES 3HNi"'p p<'-'<:-":~~<)MAY Item No.387 388 E ui ment Descri tion Primar Protection Containment Fan Cooler 225 A Breaker AH-2 (1B-SA)Containment Fan Cooler 225 A Breaker AH-2 (1B-SA)Secondar Protection 1600 A Breaker 1600 A Breaker 389 Fan S-l (1A-NNS)Filtration Unit MIS-lAR-7644 15 A Breaker 15 A Breaker 390 391 RCP-1C-SN Space Heater AOV-2CS-V521SN-1 (8141C)15 A Breaker 3 A Fuse 20 A Breaker 3 A Fuse 392 393394 AOV-2CS-V514SN-1 (8142)Valve Position Switch SM-1-LCV-408 RCP-C Stand Pipe'-LCV-408 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 3 A Fuse 395 AOV-ZCC-D224SN"1 (9472)3 A Fuse 3 A Fuse 396 397 Damper AR-D3-1 (Sol.Valve FSE-AR-D3-1)

Damper AR-D3-1 Limit Switch 6 A Fuse't 3-A Fuse 15 A Breaker 3 A Fuse 398 Charcoal Temp.Detection Fan S-1 (1A-NNS)6 A Fuse 15 A Breaker'399 400 Airborne Radioactivity Removal Unit S-1 (1A-NNS)BcAai~g RCP-1C-SN Oil Beikge Lift Pump 90 A Breaker 30 A Breakei 90 A Breaker 30 A Breaker

/4 TABLE 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES Item No.E ui ment Descri tion Primar Protection SHE pp MAY]F6 Seconda Protection 401 Containment Building 50 A Breaker Sump Pump 1A-NNS 402 Airborne Radioactivity 90 A Breaker Removal Unit S-1 (1B-NNS)50 A Breaker 90 A Breaker 403 Fuel Transfer Cont.Cab (Pump Motor)404 RCC Change Fixt (Gripper Hoist Ratio Motor)15 A Breaker 15 A'Breaker 15 A Breaker 15 A Breaker 405 Fuel Transfer Manipulator Crane 30 A Breaker 30 A Breaker.

~'

Osl TABLE 3.8-2 SHNPP RL.lfptr~i MAY Sg6 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTXON~3'/55 g7Ellic i (YES/NO)S'll 2CS<<V522)I c 5'-37Z CS-V523)IcS-9Zg 2CS-V524)lc5-I/2 2CS-V600)I 5-g I o (2CS-V601)

I c5 I 9 g (2CS-V602)lcs-ASS 2CS-V609)I c5-II l 2CS-L521)Ics-z'I2 (2CS-L522) le&-ZING'2CS-V585)

I cS-j55(2CS-L520)

I cs-2 9 I (2CS-L523)

I cs-233'2CS-V610)

I c5-I 7 p (2CS-V587)

Ics-I&9(2CS-V589) lcs-I7 I (2CS-V590) les-lIo3 pcs-v588i lcs-~IV PCS-V603)ICS-2 I7 (2CS-V604)

IcS-2 i V'2CS-V605)

I c5-2>o t2CS-V606) lcs-2"IoQCS-V611~

/c 5 g7 g(2CS-V586q l cg-79/pCS-V757(Ics-7/2(2CS-V759 I c5-7'2CS-V760 I cS-705 (2CS-V75 I cS-472 (2CS-V517)

I cs-q7o (2CS-V516)

S5.(2RH-V507~

I RH-4$(2RH-V506)

I g H-3 I (2RH-F513)

I gH-g9 (2RH-F512)

~PIl-Z (lRH-V503)

I pic-u 0 (1RH-Vsol)

I~~-I (1RH-VSO@IR Il-3'9 (1RH-Vsoo)

I 5 X-I t2SI-V503)

I SS'-I (2SI-V506)

ISS-Z (2SI-V504 I 5Z-3 (2SI-V505 2,94 (2SI-V537 2 0 9 (2SI-V535) 3 po (2SI-V571)

~Io (2SI-V573) 2.97 (2SX-V536) 3p I (2SI-V570)

I gg 3 f I (2SI-V572)

RCP A SEAL ISOL RCP B SEAL ZSOL RCP C SEAL ISOL CSIP A MINZFLOW ISOLATION CSXP C MINZFLOW ISOLATION CSIP B MINIFLOW ISOLATION CSIP TO RCS ISOLATION VCT ISOLATION RMST ISOLATION CSIPS MINIFLOW ISOLATION VCT ISOLATION RMST ISOLATION CSIP TO RCS ISOLATION CSIP SUCTION ISOLATION CSIP SUCTION ISOLATION CSIP SUCTION ISOLATION CSIP SUCTION ISOLATION CSIP DISCHARGE ISOL CSXP DISGORGE ZSOL'SIP DISCHARGE ISOL CSXP DXSCHARGE ISOL SEAL WATER INJECTION BORIC ACID TANK TO CSIP CSIP MXN1FLOW CSIP MINXFLOW CSIP MINXFLOM CSIP MXNIFLOW RCPS SEAL MATER RETURN XSOL RCP'EAL WATER ISOLATION RHR TO CSIP SUCTION RHR TO CSIP SUCTION RHR A MINX FLOW RHR B MINI FLOW RHRS INLET ISOLATION RHRS INLET ISOLATION RHRS INLET ISOLATION.

RHRS INLET ISOLATION BORON INJECTION TANK INLET ISOL BORON INJECTION TANK OUTLET ZSOL BORON INJ.TANK INLET ISOL BORON INJ.TANK OUTLET ISOL ACCUMULATOR A DISCHARGE ISOLATION ACCUMULATOR C DISCHARGE ISOLATION CNMT SUMP TO RHR PUMP A ISOL CNMT SUMP TO RHR PUMP A ISOL ACCUM B DISCHARGE ISOLATION CNMT.SUMP TO RHR PUMP B ISOL CNMT SUMP TO RHR PUMP B ISOL YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES, YES.YES YES SHEARON HARRIS-UNIT 3/4 8-21

C'i+OS1 TABLE 3.8-2 (Cont'd)SHNPP RP'>-"'r<<MAY 3986 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER I~j--I07 (2SI-V500) 52{2SI-V502)

I'5$'P&(2SI-V501)

I SX-32 6 (2SI-V577)

I5y-3>7 (2SI-V576)

I Sj:-gqo (2SI-V579)

IM 3gI (2SI-V578)

(2SI-V587)

Igy-32.2 (2SI-V575)

ISX-g 2,3 (2SI-V574)

Icc-I2+(3CC-B5)ICC.-I27 (3CC-B6)q l (3CC-B19)ICc.-I t3 (3CC-B20)I CC-(N7 (3CC-V165)(cc-I 47 (3CC-V167)

ICc-I7 Q (2CC-V172)I cc-po z (2CC-V182)

ICC->op (2CC-V170) 2"r 9 (2CC-V183)

]cc,-25 I (2CC-V190)Icc-zo7 (2CC-V169)

I cc 2'I7 (2CC-V184)I cc-s 1 9 (2CC-V191)I C T-r oS (2CT-V6)I cr-Io2.(2CT-V7)I CT'-2G (2CT-V2)I C.T-7 I (2CT-,V3)I C.T-g 0 (2CT-'V21).Ic.7 I 2.(3CT-V85)ICT-SF (2CT-V43)IcT-Ir (3GT-V88)Ic.7-L17 (2CT-V25)ICT-2.'/(2CT-V8)Ig-e5 (2CT-V49)I Cv" g>(2CT-V145)

(3AF-V18 7)Iyp-gM (3AP-v188)

InF-55 (2AP-v10)I a~-R3 (2AP-Vlg)I Ar 7'f (2AP-V23)(oF-I 37 (2AP-V116)

I AF I"t 3 (2AF-V117)

I A r=r'I I (2AF-V118) 5-70 (2MS-V8)5-7<(2MS-v9)<~a-3 I (3SW-B5)FUNCTION HH SI TO RCS HL HH SI TO RCS CL HH SI TO RCS HL LH SI TO RCS HL LH SI TO RCS HL LH SI TO RCS CL LH SI TO RCS CL LH SI TO RCS HL RWST TO RHR A ISOL RWST TO RHR B ISOL CCS NONESSENTIAL RETURN ISOL CCS NONESSENTIAL RETURN ISOL CCS NONESSENTIAL RETURN ISOL CCS NONESSENTIAL RETURN ISOL RHR COOLING ISOL RHR COOLING ISOL CVCS HX CNMT ISOLATION CVCS HX CNMT ISOLATION CCW-RCPS ISATION RCPS BEARING HX ISOLATION RCPS THER BARRIER ISOLATION CCW-RCPS ISOLATION~RCPS BEARING HX ISOLATION RCPS THER BARRIER ISOLATION CNMT SPRAY SUMP A RECIRC ISOL CNMT SPRAY SUMP B RECIRC ISOL CNMT SPRAY PUMP A INJECT.SUPPLY CNMT SPRAY PUMP B INJECT SUPPLY SPRAY HDR A ISOLATION NAOH ADDITIVE ISOLATION SPRAY HDR B ISOLATION NAOH ADDITIVE ISOLATION CNMT SPRAY HDR A RECIRC CNMT SPRAY PUMP A EDUCTOR TEST CNMT SPRAY HDR-B RECIRC CNMT SPRAY PUMP B EDUCTOR TEST AFWP A RECIRC AFWP B RECIRC AFW TO SG A ISOL AFW TO SG B ISOL AFW TO SG C ISOL AFWTD TO SG A ISOL AFWTD TO SG B ISOL AFWTD TO SG C ISOL AFWTD STEAN B ISOLATION AFWTD STEAM C ISOLATION NORMAL SW HDR'A ISOLATION BYPASS DEVICE (YES/NO)YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES SHEARON HARRIS-UNIT 3/4 8-21A v~f)

OS1 TABLE 3.8-2 (Cont'd)8HNPP Rp>,lie.,I

+5 I MAY 886 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION IS~-27@~S>-27O I5<-90 t Sw" 2,75 I Sw-g7"/isw-2, 77 ts~-3 ISQ-"I IS<-I>sw-2.I Su)-g 2.ISd-~7 Isw-9 I IS~-I O9 I5~-XU IS<-93)su)-2.2 7 ISA I I O)gyp-I Z'f i5w-l2.4 Is~-Ir'I I 5Q-I a7 Isd I 23 ISw-lrI Is'w-I 32-'ls<-t3o ICD-9g PEP-'fs (3SW-B8)(3SW-B15)t3SW-B6)(3SW-B13)(3SW-B14)3SW"B16)3SW-B3)t3SW-B4)(3SM-Bl)(3SW-B2)(2SW-B46)2SW-B47)2SW-B45)2SM"B49)2SW-B52)(2SW-B48)(2SW-B5 1)(2SM-B50)3SW"B70)3SW-B71)(3SW-B73)(3SM-B72)(3SW-B75 3 (3sw-B74)(3SM-B77)(3SW-B76)(2MD-V36)(2MD-V77)3CZ-B5 3CZ"B6 3CZ-B7 3CZ-B8 3CZ-B32 3CZ-B33 3CZ-B34 3CZ-B35 3FV-B2 3FV-B4 3CZ"B1.3CZ-B3 3CZ-B17 3CZ-B2 3CZ-B'4 3CZ-B18 3CZ-B14 3CZ-B26 3CZ-B25 3CZ-B13VALVE NUMBER FUNCTION NORMAL SW HDR A RETURN ISOL Sw HDR A TO AUX RSVR ISOL NORMAL Sw HDR B ISOL SW HDR A RETURN ISOL'w HDR B RETURN ISOL SW HDR B TO AUX RSVR ISOL EMER SW PUMP 1A MAIN RSVR INLET EMER SW PUMP 1B MAIN RSVR INLET EMER SW PUMP 1A AUX RSVR INLET EMER SW PUMP 1B AUX RSVR INLET Sw TO FAN CLR AH3 INLET Sw TO FAN CLR AH3 OUTLET SW TO FAN CLR AH2 INLET SW TO FAN CLR AH2 OUTLET SW TO FAN CLR AH1 INLET SW TO FAN CLR AH1 OUTLET SW TO FAN CLR AH4 INLET ,SW TO FAN CLR AH4 OUTLET SW TO AFWTD PUMP SW TO AFWTDWUMP SM TO AFMTD PUMP Sw TO AFMTD PUMP SW TO AFW PUMP A SUPPLY SW TO AFW PUMP A SUPPLY Sw TO AFM PUMP B SUPPLY SW TO AFM PUMP B SUPPLY CNMT SUMP ISOLATION~

CNMT SUMP ISOLATION RAB ELEC PROT INLET RAB ELEC PROT INLET RAB ELEC PROT EXHAUST RAB ELEC PROT EXHAUST RAB ELEC PROT PURGE MAKE-UP RAB ELEC PROT PURGE MAKE-UP RAB ELEC PROT PURGE INLET RAB ELEC PROT PURGE INLET FUEL HANDLING EXHAUST INLET FUEL HANDLING EXHAUST INLET CONTROL ROOM NORMAL SUPPLY ISOL CONTROL ROOM NORMAL EXHAUST ISOL CONTROL ROOM PURGE MAKE UP CONTROL ROOM NORMAL SUPPLY ISOL CONTROL ROOM EXHAUST ISOLATION CONTROL ROOM PURGE MAKE UP CONTROL ROOM PURGE EXHAUST CONTROL ROOM NORMAL SUPPLY DISCH CONTROL ROOM SUPPLY DISCHARGE-CONTROL ROOM PURGE EXHAUST P PP/I55@~V/C E (YES/NO)YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES YES~No+~vo+~mo~486-aO+/VO+QQ+VSe NO>%38 yg+488.<O+AS6 gO~PO+~uo~SHEARON HARRIS-UNIT 3/4 8-21B OS1 TABLE 3.8-2 (Cont'd)8HNP P RPJ f8JP~!ws$88 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER FUNCTION ypzss v>><<(YES/NO)3CZ-B12 3CZ-B10 3CZ-B9 3CZ-B11 3CZ"B23 3CZ-B21 3CZ-B22 3CZ-B24 3CZ"B19-3CZ-B20 3AV-B1 3AV-B2 3AV"B4 3AV-B5 3AV-B3 3AV-B6 3AC-82 3AC-B3 3AC-B1 CNTL RM EMER FLTR OUTSIDE AIR INTAKEf88 CNTL RM EMER FLTR OUTSIDE AIR INTAKE~CNTL RM EMER FLTR OUTSIDE AIR INTAKE~CNTL RM EMER FLTR OUTSIDE AIR INTAKESSS CONTROL ROOM EMER FLTR INLET CONTROL ROOM FLTR DISCHARGE CONTROL ROOM EMER FLTR DISCHARGE CONTROL ROOM EMER FLTR INLET CONTROL ROOM EMER FLTR DISCHARGE'F88 CONTROL ROOM EMER FLTR DISCHARGE 425 RAB EMER EXHAUST INLET YES RAB EMER EXHAUST OUTLET YES RAB EMER EXHAUST INLET YES RAB EMER EXHAUST OUTLET" YES RAB EMER EXHAUST BLEED YES RAB EMER EXHAUST BLEED YES RAB SWGR B EXHAUST YES RAB SWGR B EXHAUST YES RAB SWGR A EXHAUST YES~g 4.Vc~QG+~g 4-do+AJO~nJo>acluclc'egg, C.Otnp1e+n4ss OP iy~<<<)o+J by nSS J+o$4 v(ce I 5 A ccow pli gheJ by c)Pc~i~J<sl)n'HEARON HARRIS-UNIT 1 3/4 8-21C

DESIGN FEATURES SHNPP REvlS)A"-'AY

$86 t'R00F 55'IBfH IOIy DESIGN PRESSURE AND TEMPERATURE

'd 5.2.2 ,The containment building is designed and shall be maintained for a maximum internal pressure of 45.0 psig and a peak air temperature of t~'F.5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel rods clad with[Zircaloy-4].

Each fuel rod shall have a nominal active fuel length of 144 inches.and-ee~Wmnt@add dd df 1 hi to the ini ti al cor e 1 oading~8'.CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 52 shutdown and control rod assemblies.

The'hutdown and rod assemblies shall contain a nominal 142 inches of absorber material..

The nominal values of absorber material shall be 80K silver, 15K indium, and SX cadmium, or 95K hafnium with the remainder zirconium.

All control rods shall be clad with stainless steel tubing.5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained: ,a 0 In accordance with the Code requirements specified in Section[5.2]of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.For a pressure of 2485 psig,'and c.For a temperature of 650'F, except for the pressurizer which is 680 F.VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 9410 t 100 cubic feet at a nominal T of (588.8]F.5.5 METEOROLOGICAL TOWER LOCATION, 5.5.1 The meteorological station shall be located as shown on Figure 5.1-1.SHEARON HARRIS-UNIT 1 d 5" 6

~;~

OESIGN FEATURES SHNF P tQV>($~>.~~i MAY..Sg5 PROOF AND BBSY COPY 5.6 FUEL STORAGE CRITICALITY 5.6.1.The spent fuel storage racks are designed and shall be maintained with: a.A k ff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes an allowance for uncertainties as described in Section f4..3.2.6]of the FSAR, and b.A nominal 10.5 inch center-to-center distance between fuel assemblies placed in the PMR storage racks and 6.25 inch center to center distance in the BMR storage racks.5.6.1.The k ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed$0.98j when aqueous foam moderation is assumed.ORAIN AGE 5.6.2 The new and spent fuel storage pools are designed and shall be maintained to prevent inadvertent draining of the pools below eleva'tion 277.CAPACITY 5.6.3'he new and spent fuel storage pools are designed for a storage 11RRRPIIRP 1 11~d*1 1 1 PPllR BMR storage spaces in 48 interchangeable 7x7 PMR and 11x11 BMR racks.interchangeab1e racks will be installed as needed.Any combination of PMR ra'cks may be used.5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.?-1 are designed and shall maintained within the cyclic or transient limits of Table 5.7-1.capacity and These BMR and be SHEARON HARRIS" UNIT 1 5-7

6.0 ADMINISTRATIVE CONTROLS SHNPP ip+$/c~e<v 19 Nll03 klft)l'B)P('I fu3'f 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 OFFSITE 6.2.1 The offsite organizatjon for unit 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: a.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 MODE 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; e.A site Fire Brigade of at least five members" shall be maintained on h Fi Big ad)'embers of the minimum shift crew necessary for safe shutdown of the uni~any personnel required for other essential functions during a fire emergency; and g~gg+/fp/Q 7~gg,g g.l-/h5R"The Radiation Control Technician and Fire Brigade composition may be less than 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

TABLE 4.1.1-1 (continued)

REACTOR DESIGN COMPARISON TABLE THERMAL AND HYDRAULIC DESIGN PARAMETERS 15.Average in Core 16.Average in Vessel HEAT TRANSFER 17.Active Heat Transfer, Surface Area, ft.2 18.Average Heat Flux, BTU/hr.-ft.2 19.Maximum Heat Flux for Normal Operation, BTU/hr.-ft.2 20.Average Thermal Output, kW/ft.21.Maximum Thermal Output for Normal Operation, kW/ft.SHEARON HARRIS 592.6 588.8 48,600 189,800 440,400<'~5.44 12.6~VIRGIL C.SUMMER NUCLEAR STATION 591.2 589.0 48,600 189,800 440,400 5.44 12.6 22.Peak Linear Power Resulting From Overp'ower Transients, Operator Errors, Assuming a Maximum Overpower of 118 Percent kW/ft.18.0~<~~18.0M 23.Heat Flux Hot Channel Factor, Fq 24.Peak Fuel Central Temperature at 100 Percent Power F 25.Peak Fuel Central Temperature at Maximum Thermal Output for Maximum Overpower Trip Point, F 3250<4700 3250<4700

TABLE 4.1~1-1 (continued)

REACTOR DESIGN COMPARISON TABLE CORE MECHANICAL DESIGN PARAMETERS REFLECTOR'THICKNESS AND COMPOSITION SHEARON HARRIS VIRGIL C SUMMER NUCLEAR STATION 53.Top-Water plus Steel, in.54'ottom-Water plus Steel, in.55.Side-Water plus Steel, in.56.H20/U Molecular Ratio Core, Lattice (Cold)FUEL ENRICHMENT, W/0 57.Region 1 2 58.Region 2 59.Region 3 10 10 15 2.42 2.10 2.60'3.10 HO HO 2.42 2.10 2.60 3.10 n n (x)~(~)~pQ,~pePZ~Q)~~~ggg gl~k4IL See Section 4.3.2~2.6.~JIB~~H<This is the value of F pg SHNPP FSAR finally, as previously discussed, this upper bound envelope is based on procedures of load follow which require operation within an allowed deviation from a target equilibrium value of axial flux difference.

The procedures are detailed in the Technical Specif'ications and are followed by.relying only upon excore surveillance supplemented by the normal monthly full core map requirement and by computer-based alarms on deviation and time of deviation from the allowed flux difference band.Allowing for fuel densification effects the average linear power at'2775 Mwt is 5.44 kW/ft.From Figure 4.3.2-2 the conservative upper bound value of normalired locaL power density is 2.32 corresponding to a peak linear power of~~~12.9 kW/Et.at 102 percent power.n../5 Z.WS Accident analyses f are presented in Chapter 15 of the FSAR.The results th analyses determined a limiting value of total peaking factor, FQ~of 2~32 under normal operat ion, inc 1 uding load f o 1 1 owing maneuvers~Thi s value is erived from the conditions necessary to satisfy the limiting conditions specified in the LOCA analyses of FSAR Sectio 15.6.5.As noted above, an upper bound envelope of F~x Power equal to.32 x K(Z), as shown in FSAR Figure 4.3.2-21, results from operation in accordance with Constant Axial Offset Control procedures using ex-core surveillance only.2..zS To determine reactor protection system setpoints with respect to power distributions, three categories of events are considered:

namely rod control equipment malfunctions, operator errors of commission, and operator errors of omission.In evaluating the three categories of events, the core is assumed to be operating within the four constraints described above.The Eirst category comprises uncontrolled rod withdrawal (with rods moving in the normal bank sequence)for full length banks.Also included are motions of the fuLL length banks below their insertion limits, which couLd be caused, Eor example, by uncontrolled dilution or reactor coolant cooldown.Power distributions were calculated throughout the occurrences, assuming short term corrective actions that is, no transient xenon effects were considered to result from the malfunction.

The event was assumed to occur Erom typical normal operating situations which include normal xenon transients't was further assumed in determining the power distributions that total core power level would be limited by reactor trip to beLow 118 percent.Since the study is to determine protection limits with respect to power and axial offset, no credit was taken for trip setpoint reduction due to flux difference.

Results are given in Figure 4.3.2-22 in units of kW/Et.The peak power density which can occur in such events, assuming reactot'rip at or belo~118 percent, is 4.3.2-11 Amendment No.21 Core Avera e Linear Power kW/ft includin densification effects TABLE 4~3+2-2 NUCLEAR DESIGN PARAHETERS First Cycle 5.43 Total Heat Flux Hot Channel Factor, F Nuclear Enthalpy Rise Hot Channel Factor, F" AH 1.55 Reactivit Coefficients+

Doppler-only Power, Coefficients, pcm/X Power (upper limit)(See Figure 15~0.4-1), Lower Limit Doppler Temperature Coefficient, pcm/F Hoderator Temperature Coefficient, pcm!F Boron+efficient, pcm/ppm Rodded Hoderator Density Coefficient pcm/gm/cc Desi n Limits-19.4 to-12.6-10 2 to-6.7-29 to-1~4<0-16 to-7<0.43 x 105 Best Estimate-14.2 to-10.5-11.5 to-8.2-2.1 to-1 4-0 to-35-13 to-9<O.27 x 1O5 Dele ed Neutron Fraction and Lifetime g ff BOLi (EOL), BOL, (EOL)p sec.0.0075 (0.0044)i9.4 (ia.i)Control Rods Rod Requi.rements Haximum Bank Worthy pcm~Haximum E)ected Rod Worth Boron Concentrations (PPH)See Table 4.3.2-3<2000 See Chapter 15 (Zero Powers keff~1.00, Cold, Rod Cluster Control Assemblies Out, 1X h>Uncertainty Included 1430 48/76/17545.1 2.4 2.3 2.2 2.1 2,0 1.9 O 1.8 X 107 a 1.61.4 2.28 AT 0'.28 AT 6'.14 AT 10.87'.50 AT 12'.3 1.2 1.0 0 1 2 3 4 5 6 7 8 9 10 11 12 BOTTOM CORE HElGHT (FT)TOP Figure 4.3.2-21.Maximum FQ x Power Versus Axial Height During Normal Operation

SHNPP FSAR 4.4.2.11.4 Surface Heat Transfer Coeffici'ents The fuel rod surface heat transfer coefficients during subcooled forced convection and nucleate boiling are presented in Section 4.4.2.7.1.

4.4.2.11.5 Fuel Clad Temperatures The outer surface of the fuel rod at the hot spot operates at a temperature of approximateiy 660F for steady state operation at rated power throughout core life due to the presence of nucleate boiling.Initially (beginning-of-life), this temperature is that of the clad metal outer surface.During operation over the life of the core, the buildup of oxides and crud on the fuel rod, surface causes the clad surface temperature to increase.Allowance is made in the fuel center melt evaluation for this temperature rise.Since the thermal-hydraulic design basis limits DNB, adequate heat transfer is provided between the fuel clad and the reactor coolant so that the core thermal output is not limited by considerations of clad temperature.

4.4.2.11.6 Treatment of Peaking Factors~g.g'he total heat flux hot channel factor, Fq, is defined as he ratio of the maximum to core average heat flux.The design value'F0 as presented in Table 4.3'-2 and discussed in Section 4.3.2.2.6', is.for normal operation.

This results in a peak linear power of 2.o W/ft.at full power conditions.

lz.4 As described in Section 4~3.2.2.6, the peak linear power resulting from overpower transients/operator errors (assuming maximum overpower of 118 percent)is 18.0 kW/fthm The centerline temperature kW/ft must be below the U02 melt temperature over the li ctime of the rod, including aLLowances fot'ncertainties The fuel temperature design basis is discussed in Section 4.4.1~2 and results in a maximum allowable calculated centerline temperature of 4700F.The peak linear power which would result in centerline meir, is>18.0 kW/ft.The centerline temperature at the peak linear power resulting from overpower transients/operator errors (assuming a maximum overpower of 118 percent)is below that required to produce melting.4.4.2-15

SHNPP FSAR TABLE 4.4.2-1 (continued)

THERMAL AND HYDRAULIC COMPARISON TABLE Desi n Parameters Shearon Harris V.C.Summer Heat Transfer Active heat transfer, surface area (ft.2)Average heat flux (Btu/hr.-ft.2)

Maximum hest flux for normal'operation (Btu/hr.-f t.2)48,600 189,800 440,400(a) 48,600 189,800 440,400 Averge linear power (kM/ft.)5e44 5.44 Peak linear power for normal operation (kM/ft.)12.6(a)12.6 Peak linear power resulting from overpower transients/operator errors, assuming a maximum overpower of 118X (kV/ft)18.0 18.0 Peak linear power which would result in centerline melt (kW/fthm)Power density (kR per liter of cote)(d)Specific power (kM per kg.uranium (d)Fuel Central Tem erature>18.0 104.5 38.4>18.0 104.5 38.4 Peak at peak linear power for prevention of centerline melt (F)<4700<4700 Pressure Drop (e)Across core (psi)'Across vessel, including noszle (psi)23.4+2.3 41~0+4.1 23'+2'40.7+4'6p NOTES;Q,I W IT 7a)This licit is asseeiatad with the value ef Pq 2.32.g5 4lv&J fH~~~$P~f.(b)See Section 4.3~2.2.6 (c).See Section 4.4.2.11.6

~~w (d)Based on cold dimensions and 95 percent of theoretical density fuel.(e)Based on best estimate reactor coolant flow rate as discussed in Section 5~1~4.4.2-17

6,2,1.5 Minimum Containment Pressure Anal sis for Performance Ca abilit Studies of Emer enc Core Coolin S stem The containment backpressure used for the limiting case CD=0.Q, DECLC break for the ECCS analysis presented in Section 15.6.5 is pre'sented in.Figure 6.2.1-302.The containment backpressure is calculated using the methods and assumptions described in"Westinghouse Emergency Core Cooling System Evalulation Model-Summary," WCAP-8339, Appendix A.Input parameters including the containment initial conditions, net free containment volume, passive heat sink materials, thicknesses, surface areas, starting time, and number of containment heat removal systems used in the analysis are described belov.The anaylsis vas performed assuming the loss of offsite power as the most limiting condition.

As indicated in WCAP-8471, the three loop plant limiting case break (CD=0.4 DECLC)yields lower calculated PCT values with offsite po~er available (reactor coolant pumps run case)than if offsite power is lost (reactor coolant pumps trip case).This results Erom core thermal hydraulics during blowdovn and is true even'though calculated containment pressure may be lowerin the offsite power available case due to Easter actuation of the engineered sefeguards.

The applicability of the generic conclusion regarding offsite power status to the Shearon Harris ECCS analysis is presented in detail below.'review'of the original three-loop plant generic sensitivity runs demonstrated the large benefit in calculated clad temperature vhich exists at.end oE blovdown in the offsite pover available case.Calculated clad temperature at end of bLovdovn at the limiting fuel rod elevation (7.25 ft)is 1528 F vith offsite power lost;vith offsite pover available, the calculated clad temperature at the equivalent location is only 1453 F at end of blovdovn.Hence, the blovdovn performance calculated vith offsite pover available produces a clad temperature result at end of blovdown vhich is 75 F better than with loss of offsjte power assumed.This benefit will remain in effect throughout the core reflood transient, during vhich time the PCT is calculated.

During the core refLood transient the reactor coolant pumps are assumed to be in the Locked-rotor configuration independent oE the availability of oEfsite power.The impact oE containment pressure on ECCS performance is important only during the core reflood transient.

If offsite power is presumed available,the start times oE the containment fan coolers and sprays at Shearon Harris vill be reduced by ten seconds.The ten seconds of additional heat removal by these systems vill reduce calculated containment pressure during ref lood by less than 0.4 psi;the impact of this pressure reduction on calculated PCT is less than 25 F.Overall, then, the total effect of assuming offsite power to be available during a large break LOCA event at Shearon Harris is to obtain a more favorable result.The Westinghouse ECCS performance analysis generic assumption of loss of offsite po~er is Limiting for Shearon"Harris, and the results presented in the FSAR demonstrate compliance vith 10 CFR50.46 for this limiting case.6.2.1-26b 6.2.1.5.1 Mass snd Energy Release Data The mass/energy r'eleases to the Containment during the blovdown and ref lood portions of the limiting break transient are presented in Tables 6.2.1-59 through 6.2.1-61.The mathematical models vhich calculate the mass and energy releases to the Containment are described in Section 15.6.5.Since the requirements of Appendix K of 10CFR50 are very specific in regard to the modeling of the RCS during blovdovn and the models used are in conformance vith Appendix K, no alteracions to those models have been made in regard to the mass and energy releases.A break spectrum analysis is perEormed (see references in Section L5.6.5)that analyzes various break sizes, break locations, and Moody discharge coefficients Eor the double ended cold leg guillotines which do affect the mass and energy released to the Containment.

This effect is considered for each case analyzed.Dur'ing refill, the mass and energy released to che Containment is assumed to be zero, vhich minimires the containment pressure.During reflood, the effect of steam-water mixing betveen the safety injection water and the steam flowing through the RCS intact loops reduces the available energy released to the containment vapor space and therefore tends to minimize containment pressure.6.2.1.5.2 Inicial Containment Internal Conditions The folloving initial values were used in the analysis'.

Containment pressure Containment temperature RWST temperature Service water temperature Outside temperature Initial Relative Humidity 14.7 psia 90 F 40 F 3g M F-2 F 100 X The initial temperature condition that may be encountered under limiting normal operating conditions used in the ECCS performance analysis was assumed to be 90 F.An evaluation determined that the containment cannot Eall below 80 F, and the normal expected average containment temperature estimated at 100 F.The 90 F value vss chosen because it vas shown to be a conservatively lou value'consistent uith representative normal full paver operar.ion oi other nuclear piancs.The normal operating range ior containment pressure is expected to be betveen negative 1 inch vg to positive 4 inch vg vith the nominal pressure expected to be slightly positive.The value of 14.7 psia vas assumed for the ECCS performance analysis.The containment is the atmospheric type per Item d of SRP 3.8.The normal containment purge and makeup systems along vich che containment cooling system vill maintain the containment vithin the normal operating range.The Normal Containment Purge Exhaust is Eirst adjusted to sllov the system to draw down the containment atmosphere to a slight negative pressure (to prevent outleakage).

Mhen the containment pressure is reduced to-0.25 in.vg, one of the tvo 100 percent capacity makeup fans vill automatically start.The static pressure controller vill regulate the respective supply fsn inlet damper to modulate snd maintain the containment pressure setpoint.The pressure transmitter for controlling this 6.2.1-26c

))The Frictional r<<s(sLa<:ce assnciaterl with duct e<t(ra<>c..

<<>xir asses, f (it.ers, duc(w>rk bends and skin fric:(<<n has nnt.been cons(d.red.

c)No fa.c<)as(down ef fec(ar;considered

~d)in>>rtia is cor.s(dered

~St.eady sta(e flow is es(ahl~hed (m.".ed'a'el:

at.t.he t.ime of t.he LOC>>.ouL Lhe p<<<rg>>~stem d<<" t.-A m(x(ure of s earn a>ro<<<;h L<<<<<purge lines duri Lhe 6.r)3 seconds that.Lhe is<lat(nn>al ves are assume l L~remain open.The'ect.nf t.he compos((ion nf t!te~being exhausted or, cur ta(nme<L pressur>>as b>>u".bound>>d by invest~(.tpt (r.g t.he two ex(rem>>cas<s, a(r alo".e and steam alo e~'M(Lh(n several seco 8s of th~incept.ion of t.he LOCK, con(aianenL pressur will have increas.<tn the point, that crit(cal F(nw will occur in the p<<rg lin<y~Tn bo<<rd L e calo<<lat.d gas mixture exha<<<s(ud t.hriu"h t.he purge lines, t,he Xr(ical f w ra(es of stean and air wer~calc<<la(ed d<ring t.he first 6.03 s o s F t.he CD~0.4 OFCLG break t.ransient. Using Lttesn flow raLe<t, r tical flow was then c<<nservaL(ve'.y assumed to b>>ir.effect.from Lim~rn.Fquation 4.18 in Reference 6.2.1-13 was employed t.n cal ulzte t.he'calif l rat.e of air Lhro<<<gh the purge lines.Fig<<re 14 of Ref>>re 6.2.1-14 was plied to compute Lhe criLical flow ra(e of st.earn t.hruu e purge lines.T.tntal mass released during the 6.03 seconds Lha(t<, valves are presume<1<>pe is calculated as 331 lhm a(r or 239 ibm st.ea.~he impact or.containment pr sure at 6.03 sec<)nds result.ing frnm t.1'oss of air or steam is less than.05 psi in either case.The eff t of a contai.nment pressure reduction o this magnitude nn t.h.calculated p k cla9 temperature (PCT)is less Lhan 1 deg-The PCT fnr Lh<DECLG$0=0 case is 2181 deg-F aL an Fg of 2.11-Therefore, there is no FQ penal(i(nd margin w(th respect.Lo (OCFR50.46 PCT requirement.s o<

>t.ers 'to other parameters have a substsnL(al eFfect, nn t.he m(nimum cnnta(nmenL press<<<ce analysis'6.2.1.6 Test(n and Ins ect'on S:ruct ra'".Le<ri(y Lests anrl in--erv(ce surveillance requirement.s are discussed in Section 3.8.1~7 for the Cont.ainment. Building and in Section 3.8.2.7 for t.he Class'.iC cnmpnnen(s (penetrations, locks, and hatch<s).CnnLainmenL leakage test.ing is discussed in Sect.ion 6.2,6.Test.i<<~-and inspection requirem>>n(s For t.he Vacu<<m Rel,(ef Syst.em and engi<te<.'red safe(y feat.ures t;hat.affect.the funct.ional. capabil(t.y of the ConLainment are'iscussed in Sect.inns I.6.2 an<i 6.6.Preoperat.ional testing is des<:ribed i".Section 14.2.12.6'.1~7 Inst.rumentat.inn A 1(caLinr.Pressure sensing i.nst.rument.s mo<<it.nr t.he cont.ainmenL atmosphere and iniriz(e the con(a(nment isnlatlon, s<<Fet.y i<tjec(ion, an<i cont.ainment. spray act.<<at.(on signals according Ln the logic disc<<ssed in Section.-7.3 Radiation mnnit.nrs mnnitor cont.ainmer,L at.mospher<: and.(s<>late.conLainmenL;purge.thro<<gh t.he conLainment vent(laL(on isolat.ion signal;.cs discussed in Section 7.3. TABLE 6.2.1-59 BLOWDOWN MASS/ENERGY RELEASES DECLG C=O.PV Time (sec.)Mass Flow lb/sec.Enez Flow Btu/sec.0.0.05 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0 18.0 20.0 a2~o 2+009 29'.o 3Org, 0.0 5;Szl f~x 10 V.ZCO x 10 2,C9 2-x 10 Z.r vC&r9+5 x 10/.ZV9~x 10/.JCo q.~x 10>7~x 10'.czar~x 103 c~t5'r~x 103 Y.ssF x 10 Q PS%'10 M 3J'/1~x 10 p,77/p'o QiP r/3 gyp wro 0.0 c.f ZP SWAN x 10 x 10 r.0"4 V~+2-x 107 r.zf v x 107/,/2%~x10 F.s~5~x 10~Cef M~x 10 C,.~<+'x 106 5'.z/r~~x 106 g.gc5~+2-x 10 3.o pr.10~'~'Po Z~x10 p~/g7 Ago C C y,~y~M/4'Z/X/O 5 5 g gy~/0 6.2.1-164 TABLE 6.2.1-60 REFLOOD MASS AND ENERGY RELEASES DECLG C"-0.4 Time (sec.)Mass Floe lb/sec.Ener Flow Btu/sec.'fX a(44 69-RZ-"1 I g$.Kl 48-.~4 5.6 I/0 g.p3 K6~)(3 , o'3 k8+eK/5+,9 5 NkrH8 34,g, Z 3 em<-0.RO&~Z.'l 5'L 3~Z~3+4~3r P.F g,~~32he V~1 SY+7g 0.w.5 o~x 10~x 10>.5'1~x10 l H9 4-r&Q x 105 I 95'10~I.~x 10~r SP 1~x 10 l tE~x 10~6.2.1-16S TABLE 6.2.1-61 BROKEN LOOP ACCUMULATOR MASS AND ENERCY~YfLEXSES TO CONTAINMENT (C=0.4)Time (sec.)Mass Flow lb/sec.Ener Flow Btu/sec.0.000 1.010 2.010 3.010 4.010 5.010 6.010 7.010 8.010 9.010 10.010 11.010 12.010 13.010 14.010 15.010 16.010 17.010 18.010 19.010 4678.13~4&6 9/4 7.o/~44-.+8 3 7 5$, Z S-9+59~Eh7 3 (fan y'215%8 3 2.~3./g M20%3 3~~Z f4 2%4&9 2.g p!.7 g~4-15~g9,88 Q C/f.fo 2r 81.+1 1 2-9/V.Zw Z.52c.pp 2 QR 2.'Ly7.c;I 2J.51~Z./7 s"./g L/o 8.7+20%7.+o l990.7o i%~z.zx~11~1/~>'5: S 5~~8/I'VV.v 0 278910;37 2bErf~3 2944~'J', Z'3~V<E'3 l~~!5 f>Z+./7/3>ss~.~C 1~~/-.>CVZ.rJ/2 5/'/.5o I 2;os I'.e 7//fc E'5.Zt.1~3-.H i!g A"c,.S'y 14453%46 I/ZC<7.C 1100%444/o>>c g.~-g~'~7 2.s 5 yz.oq~407.SQ-, 9 Z7C z 1/180066.Z4-r/'~'-~~u~!7!t<<<<1~S-.+e i4'~>>////S///I 47/~rj 7 2 Swy//o 3~i/z 20.010 21.010 7 1~2/j'02.~/I/~(.2.5k/y7~e'/.5/10'~~/oc,"o E5/*Enthalpy of accumulator water is 59.6 Btu/ibm.6.2.1-166 //J 5'~X (LZ.o(o~pZS.wc roz s-~i r<Z3.0 (0/Co++~6 yoc 7FC.7)2.f.Ot'0 IC S P.z.~+z'rc p'-Z7 ZC o(o~4 Z P./<P 7 dC'7.ZZ Z<.o/o/~~f-7f~s g7p.iv Z7.o I o/7lc.<F 3 P-ohio/P3 cM l5 5<gz Zgio/'b/fZ.sY j 5-5-~, o z. Z I CII II O II y I Z lo n+o M Ol m mZ'lI O I7 Z C A Eh rz mm O ao OZ F$z m+~2 rc Z I O P U CO K D K 4 I Z W R I-z 0 CP D CII ll~0 C CA Ill M I-z LLJ O LL LLI O K LL LO z K I I-LL.<o, LLL CV X I LL z~Cll w 2 3 LLJ I Oz O U I R z I-z 0 U TIME (SEC)SHEARON HARRIS NUCLEAR POWER PLANT Carolina Power 5 Light Company FINAL SAFETY ANALYSIS REPORT CONTAINI"ENT WALL HEAT TRANSFER COEFF I CIEHT LFIGURE 6.F 1-'304 Z I~~I s I~s I~se I~tn tt 0 AM rz m-mm 4 ti'n~~~V r."~O Z~~+OZ f)os my Ot 1st t)1+CO m~.Z 0 O O Z I O ttl A r Gl 0~~I-z:-'.I-..01: O i.os.e.l:t~I~~e I~ll:I: 'jll~~:: I Ij ol~I~I elj~I lee I:I~lj!i I:o I~~~II.I ls:~I i I~~~~~~~~~~~I~o sl j."~I~*~I!j::ijl',::rr', pi s:il~ei iill hei:1st I~~~!Ill iili."Ii o IVI~~'t!I l I'I.tl 1+(11 I I'j, tjjj!L'I illi jlsi ilji ti:~I I j'i::i: jli::llr iiii'!l'j I::i.'!!I!1::li:: Il"oi::sl~I~~s~I~~~I~~~~~~e;11~oet:!il'.'sjIi.I:~;I~~i"'el iilo rs ill:Ii is l Ii s~o~~loll s;:::I I',ll jilt its I l': I~l I)I'l I Il ji i I~ls I I l I I il I I I l~~iJ i!itl I I i l I io~ilt;lilt~~I ii:: I I Is I!!t I, I I!I'II'I I l Ill ill!:.I'I~~~~s I~p~11 I~ji~~ijii iii I I!!i I i'o'III I:: Is.: lji!~~be Ij!i I 11!i~I~I I I I: I I I e I!;I ,.'II l i I:: el ji~I!I:"j 11'io'l~I-" Itll lli I~j:l)ji il!II I I l ji jtj', I 11 Ilti 1 I.I;::~i.11;:~I~~I li I: I: I: I;Ss II iii!I ill I: I~'li~I il: j!li I I;:l.Io~ee',I:~~I li!Il:!i:~I I I~~~I I I~I Is II~~;I'il~~IIjl:lli'ill I~I'11:i I:I~I-~os Itji~mo~~ol: le: s lii!Ij:: ii'1 I~~~~!ji'ji'1 il 0 C O m Ul j.i~'!ll;J;SI4 I~!::'8 I~II i]6 TIME (SEC)Ssi 24 I'sl: I'.:.'i:~I::,e'~~~tel Ij~I:I ll e l~~I I 0e~.Ii I~40 TABLE 15..0.3-2 (Continued) pauits Computer Codes Utilized Reactivity Coefficients Assumed Hoderator Hoderator Temperature Density (hk/F)(hk/gm/cc) Doppler Initial.HSSS Thermal Power Output Assumed (HMt)Loss of coolant accidents resulting from the spectrum of postulated piping breaks Mithin the reactor coolant pressure boundary SATAN-VI, WFLASH WREFLOOD, COCO, LOCTA-I GART See Section 15.6.5, references See Section 15.6.5, references 2775 25 Ln o fag I CO a.See Figure 15.0.4-1 g s A minimum of (percent margin is applied to the values shown for analysis purposes.c.pcm gg 1 x 10 Ap d.Analysis based upon rated poMer average reactor coolant system vessel temperature of 588.8 F.~(See Section 15.0.3.2)e.LOCA analysis performed at best-estimate T(590.0 F). ~~~~~~~l5.b.5 I>.b.5.1 LOSS VF COOLANT ACCI,Dt.NTS* Identification of Causes and Fre uenc Classification h loss-oX-coolant accident (LOCA)is the result nf a pipe rupture of the Keactnr Coolant System (RCS)pressure boundary.A ma)or pipe break (large break)is defined as a rupture with a total cross sectional area equal to or greater than 1.U ft.2.This event is considered a limiting fault, an ANS Condition IV event, in that it is not expected to occur during the lifetime of t.he plant, but is postulated as a conservative design basis.A uinor pipe break (small break)is defined as a rupture of the reactor coolant pressure boundary with a total cross-sectional area less than 1.U ft.Z in which the normally operating charging system flow is not sufticient to sustain pressurizer level and pressure.This is considered a Ahh Condition III event in that it is an infrequent fault that may occur during the life of the plan't.The ncceptance criteria for the loss-of-coolant accident is described in lv L'Fk 5U Paragraph 4b (Reference 15.6.5"1)as follows: j a)The calculated peak fuel element clad temperature is below the requirement of 2ZUU VS b)The amount nf fuel eLement cladding that reacts chemically with water ur steam does nut exceed one percent nf the total amount of Zircaloy in the reuc ter~c)The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.The localized cladding oxidation limits ot 17,percent ~re nor exceeded during or after quenching. a)Tne core remains amenable to cooling during and after the break.e)The core temperature is reduced and decay heat is removed for an extended period of time, as required by the longlived radioactivity remaining in the core Tiiese criteria were established to provide siginificant margin in ECCS performance following a LOCA.(Reference 15.6i5-2)presents a recent study in regard to the probability of occurrence of RCS pipe ruptures.In all cases, small breaks (less than 1~0 fthm>)yield results with more margin to the acceptance criteria limits than large break.1 hobo 5,2 Se uence of Events and S stems 0 rations bhoulct a mayor break occur, depressurization of the RCS results'in a pressure decreas'e in the pressurizero The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached.A safety in)ection actuation si na3 fs generated when the appropriate setpoint is reached.These countermeasures limit the consequences of the accident in two ways:*Additional information is contained in the TNI Appendix, I a)Reactor trip and borated water inject,ion compLement void Eormation in causing tapid reduction of po~er co a residual level corresponding to fission product decay heat.b)Injection of boraced~ater provides for heat transfer from the core and prevents excessive clad temperatures. In both Large and small break LOCA analyses loss-oE-offsite po~er coincidenc with the accident is assumed.The single failure in defining SI flow rates subsequently considered is the loss oE a dieseL generator; thus, only one train of ECCS flow of the two actually present is considered to be available. Therefore, for boch large and small break LOCAs, KCCS flow co che core is at a conservativeLy Low value following ics automatic actuation, especiaLly since all wacer delivered to the broken loop is considered to spill directly to the concainmenc sump.Notwithstanding r,hese conservatisms, conformance with the 10CFR50.46 acceptance criteria is demonstrated in the Large and small break LOCA Analyses.No other postulated single EaiLure wo ld have as great an effect on PCCS flow deLivery.Single failures do not have a significant eftect on final containment water levels following a postuLated LOCA.The ECCS cerminacion and reinitiation criteria provided in the Harris Plant Emergency Operating Procedures (KOP's)are designed to minimize any possiblity of an operator error to improperly or prematureLy shut off safety injection from challenging core cooling.Termination criteria for high pressure safety injection flow (HPI)following a LOCA event call Eor a shutoff of all HPI when the RCS pressure is stable or increasing <<nd subcooling exists,-the pressurizer level is on span plus errors and steam generators are being fed auxiliary Eeedwater or have indicated Level above the U-tubes.For a break as smalL as a 0.5" equivalent diameter hole, as soon as the HPI is terminaced, a rapid depressurizacion of the.system occurs.EOPs will.direct the operator to immediaceLy reinitiate sat'ecy injection and to pertorm a controLled cooldown with SI flow, thereby ensuring that the core will, remain covered and adequately cooled.Any break postulated to occur in the KCCS line is bounded by the spectrum of breaks presented in the FSAR Standard assumptions used in defining safecv n jeccion flow tor eicher a Large or small cold leg break LOCA analysis Lncludet 1)The spilLing ot the broken loop accumulator directly to containment. 2)The spillina of the safety injection Line attached to the broken cold Lee.3)A single"tailure condition such that onLy one train ot saiety injection pumps operates'ach ot che three RCS cold legs has an injection line attached.into the RCS is computer based on the tollowing Logic.FLuw delivered One train of ECCS pumps starts and delivers tlow into the reactor coolant)yscem through cwo branch injection lines. One br'a@eh injection line spills to containment backpressure. The branch injection line with minimum system resistance is selected to spill to minimize delivery to the core.t The flow delivered into the reactor through the reactor coolant pump seals is assumed to be lost and, therefore, seal injection is not included in the total core delivery.Safety injection flows computed via this methodology are conservatively low for any postulated break location.Descri tion of Lar e LOCA Transient The sequence of events following a large break LOCA are presented in Figure 15.6.5-1~Before the break occurs,%he Unit is in an equilibrium condition, i.e., the heat generated in the coz%is being removed via the secondary system.During blowdown, heat from fission product decay, hot internals and the vessel continues to be transferred to the reactor coolant.At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed-nucleate boiling.Thereafter, the core heat transfer is based on local conditions with trans.ition boiling and forced convection to steam as the major heat transfer mechanisms'he heat transfer between the Reactor Coolant System and the secondary system may be in either direction depending on the relative temperatures. In the case of continued heat addition to the secondary, secondary system pressure increases, and the main steam safety valves may actuate to limit the pressure.Make-up water to the secondary side is automatically provided by the Auxiliary Feedwater System.The safety.injection actuation signal isolates the steam generators from normal feedwater flow and initiates emergency flow from the Auxiliary Feedwater System.The secondary flow aids in the reduction of reactor coolant system pressure.Mhen the Reactor Coolant System depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops.Since the loss of off-site power is assumed, the reactor coolant pumps are assumed to trip at the inception of the accident.Previous sensitivity studies have demonstrated the conservatism of this assumption for large break LOCA analyses.The effects of pump coastdown are included in the blowdown analysis.The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2280 psia)falls to a value approaching that of the containment atmosphere. Prior to or at.the end of the blowdown, some amount of injection water begins to enter the reactor vessel lower plentum.At this time (called end of bypass)refill of the reactor vessel lower plenum begins.Refill is complete when emergency core cooling water has filled the lower plenum of the reactor vessel which is bounded by the bottom of the fuel rods (called bottom of core recovery time.)15.6.5-3 SHNPP FSAR Qo CRAB&~The reflood phase of the transient is defined as the time period lasting from the end of refill until the reactor vessel has been filled with eater to the extent that the coz'e temperature rise has been terminated. From the later stage of blovdoun and then the beginning of ref lood, the safety injection accumulatoz tanks rapidly discharge borated cooling eater into the RCS, contributing to the filling of the reactor vessel dovncomer. The dovncomer vater elevation head provides the driving force required foz the z'eflooding of the reactor core.The RHR (lo~head)and charging (high head)pumps aid the filling of the dovncomer and subsequently supply eater to maintain a full douncomer and complete the ref looding process.Continued operation of the ECCS pumps supplies eater during long-term cooling.Coze temperatures have been reduced to long-term steady state levels associated ~ith dissipation of residual heat generation. After the vater level of the refueling eater storage tank (EST)reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching from the injection mode to the gold leg reciz'culation mode of operation in which spilled borated eater is dyad from the containment sumps by the pumps and returned to'he RCS cold legs.The Containment Spray System continues to operate to further reduce containment pressure.Approximately 24 houzs aftez initiation of the LOCA, the ECCS is realigned to supply eater to the RCS hot legs.in Order to control the boric acid concentration in the reactor vessel.Descri tion of Small Break LOCA Transient As contrasted with the large break, the bio~down phase of the small break occurs over a longer time period.Thus, for the small break LOCA there are only three characteristic stages, i.e., a gradual bloudoun in which the decrease in eater level is checked, core recovery, and Long-term recirculation. 15.6.5.3 Core and S stem Performance 15.6.5.3.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50 (Reference 15.6.5-1). Lar e Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases: 1)blovdoun, 2)refill, and 3)reflood.There are thz'ee distinct transients analyxed in.each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient vithin the Containment, and the fuel and clad temperature transient of the hottest fuel rod in the core.Based on these considerations, a system of inter@elated computez codes has been developed for the analysis of the LOCA. SHNPP FSAR The description of the various aspects of the LOCA analysis methodology is given in MCAP-8339, Reference 15.6.5-3.This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure corn li nce vith the acceptance criteria-The SATAN-VI, MREFLOOD, COCO, , BART, and LOCTA-IV codes, which m e used in the LOCA apalysis, are described in detail in References 15.6.5-4 through 15.6.5-7,~A'15.6.5-31 These codes are used to assess the core heat transfer geometry an to determine if the core remains amenable to cooling throughout and subsequent to the blovdovn, refill, and reE.lood phases of the LOCA.The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during hioudoun and the WREFLOOD~~Pcomp tuer cod~.~c~used to calculate this transient during the refill and reflood phases of the accident.The COCO computer code is used to calculate the containment pressure transient during all three phases of the LOCA analysis.Similiarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three phases.'ATAN-VI is used to calculate the RCS pressure, enthalpy, density and the mas's and energy flov rates in the RCS, as well as energy transfer between the primary and steam generator secondary systems as a function of time during the blowdovn phase of the LOCA.SATAN-VI also calculates the accumulator mass and internal'ressure and the pipe break mass and internal energy flow rates that are assumed to be vented to the Containment during blovdovn.During blowdovn,. no credit is taken for rod insertion or boron content of the injection vater.The core vill shutdovn due to void formation. At the end of the blowdown phase, these data are transferred to the MREFLOOD code.Also at the end of blovdovn, the mass and energy release rates during blovdovn are transferred to the COCO code for use in the determination of the containment pressure response during this first phase of the LOCA.Additional SATAN-VI output data from the end of blovdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.provide e rea~CS du ng'he r ood vill e 4 sti ed f c therma hy a~sxmulat of the'eactor core and phase CA~Figure.6.5"2 g11u rates hov BASH r MR in culati tra sient u s of c inlet xs a n gral 4 of the E S evaluation model.replacing MREFLOOD to~W~flov, dn alp, and p esp re or t Ins anta eo values of.'ac mula r 1 a fent t perature d essure, a d 1 wer p um refil'e ovide roposc'd ECCS mo 1, has b~n relegat boundary condit ons.A mare detailed MCAP-10266 (Reference 15.2'-32).otail d'fuel rod'm el, LOCTA V ow safety n'4c on f ow, nt inment e mime o o letio.of ea or vessel o ASH by LOOD/C 0 vhic , in the solely providi~'g hesy requirA'd descrip ion of thCcode if avail'ble,'in The MREFLOOD code provides mass and energy discharge rates from the reactor coolant system to the containment during a core reflood transients A brief overviev is presented here.A complete description of the code is available in MCAP-8170 (Reference 15.6.5-5)The basic geometric configuration in MREFLOOD divides the primary coolant system into three sections: the reactor vessel, the broken loop, and a second loop which combines all unbroken loops.The reactor vessel region is further divided into a downcomer, lover plenum, and core.Using the injection 15.6.5-5 v'ag i~'n%characteristics of the ECCS as input, the code calculates the dogdncomer and core water levels as the reflood.transient continues. Other basic input to MRKFLOOD includes geometric data and initial and boundary conditions in the core, steam generators, and containment'. The COCO code is a mathematical model of the containment. Selection of various options in the code allous the creation of models of particular containment buildings. COCO is described in detail in MCAP-8327 (Reference 15;6.5-6). COCO is run simultaneously with MREFLOOD, which provides the neces ary mass and ener in uts nt on a continuous ba is.In t x an ysis th h, MiLQL~isonly.a su a , ru ni arallel~e m in ns t an l,co e, B~~D 6.n" r fl od, e MR L C+0 s~is ed o~prov'ide ntainme'und oedi's reQiredht gAQ.The LOCTA code is a computer program that evaluates fuel, cladding, and coolant temperatures during a LOCA.A more complete description than is presented here can be founl in MCAP-8301 (Reference 15.6.5-7). y D re'd rel , th S mod l ses>a o fied ersio of CT ie ignif imp)f d be'c(et'AR t sfe fo lcu'o corr at i 1 s vhp aced at Qd gdgT oode~li employs rigorous meohaniseio smde~ro generaee hearr sanfse r coefficients appropriate to the actual flov and heat transfer regimes experienced by the LOCTA Euel rods.This is considered a more flexible, realistic approach than relying on a static empirical correlation. Small Break LOCA Evaluation Model The MFLASH program used in thy analysis of the small break loss of coolant accident is an extension of the FLASH-4 (Reference 15.6.5-15) code developed at the Mestinghouse Bettis Atomic Power Laboratory. The'MFLASH (Reference 15.6.5-16) Program permits a detailed spatial representation of the Reactor Coolant System (RCS).The RCS is nodaliaed into volumes interconnected by flovpaths. The broken loop is modeled explicitly vith the intact loops lumped into a second loop.The transient behavior of the system is determined from the governing conservation equations of mass, energy and momentum applied throughout the system.A detailed description of MFLASH is given in Reference 15.6.5-16. The use of MFLASH in the analysis involves, among other things, the representation of the reactor core as a heated control volume vith the associated bubble rise model to permit a transient mixture height calculetione The multi-node.capability of the program enables an explicit and detailed spatial representation of various system components. In particular, it enables a p'roper calculation of the behavior of the loop seal during a loss-of-coolant transients 15.6.5-6 Clad thermal analyses are performed with the LOCTA IV Code (Reference 15.6.5-7)which uses the RCS pressure, fuel rod po~er history, steam flow past the uncovered part of the core and mixture height history from the WFLASH hydraulic calculations as input.Figure 15.6.5-44 gives the safety injection flowrate for the small break analysis.Figure 15.6.5-45 presents the hot rod power shape utilixed to perform the small break analysis presented here.This power shape was chosen because it provides an appropriate distribution of power versus core height and also local po~er is maximized in the upper regions of the reactor core (10 ft.to 12 ft.).This power shape is skewed to the top of the core with the peak local po~er occurring at the 10.0 ft.core elevation. This is limiting for the small break analysis because of the core uncovery process for small breaks.As the core uncovers, the cladding in the upper elevation of the core heats up and is sensitive to the local power at that elevation. The cladding tepperatures in the lower elevation of the core, below the two phase mixture@height, remains low.The peak clad temperature occurs above 10 ft.Schematic representations of the computer code interfaces are given in Figures 15.6.5-2 and 15.6.5-3.The small'break analysis was performed with the approved October, 1975 verison of the Westinghouse ECCS Evaluation Model (References 15~6~5-7, 15.6.5-16', and 15.6.5-17 and 15.6.5-25). 15.6.5.3.2 Input Parameters and Initial Conditions Table 15.6'-2 lists important input parameters and initial conditions used in the analysis.The analysis was performed with the upper head fluid temperature equal to the reactor coolant system cold leg fluid temperature, achieved by increasing the upper head cooling flow (Reference 15.6'"18)~In order to achieve upper head temperatures in the T cold xone, bypass flow was diverted into the vessel head region.h study was performed and documented in Reference 15.6.5-26 to determine the amount of bypiss flow necessary to achieve T cold conditions in the head.hs described in Section 2 of Reference 15.6.5-26, an analytical model for upper head temperature calculation was developed for both UHI and non-UHI plants'o estimate the upper head region fluid temperature with the anaLytical model, numerous boundary conditions must be known.The boundary conditions used were based on experimental data obtained from a series of three hydraulic tests conducted at the Westinghouse Forest Hills facility.These tests were the UHI flow distribution test, the 1/7 scale UBI upper internals test and 1/7 scale 414 flow.test.To provide experimental verification of the analyticaL modeli i 1/5 scale model upper head temperature test was developed as described in Section 3 of Reference 15.6.5-26. Results for both UHI and non-UHI pLant shoved good agreement with analyticil predictions. Further confirmation of the analytical procedures was obtained by an in-pLant heid fluid temperature Neasurement 15.6.5-7 SHHPP FSAR program as described in Section 4 of Reference 15.6.5-26. The program included meas:irements from 2, 3, and 4 Loop plants.Boch UHI and non-UHI plants were measured.A'll three types of upper core place designs (flat, top hat, and inverted top hat)were included as well as both neutron shield configurations (thermal shield and neutron pad).As reported in Section 4 of Reference 15.6.5-26, good agreement was reached betveen measurements and the analyticaL modeL for che above spectrum of non-UHI plant, types.this provides good assurance that the upper head fluid temperatures have been adequately calculaced by the analytical modes described in Reference 15.6.5-26 A break spectrum sensitivity study is presented in Reference 15.6.5-19. The bases used to select the numerical values that are input parameters co the analysis have been conservatively determined from extensive sensitivity, studies (References 15.6.5-19 through 15.6.5-22). In addition, the requirements oE Appendix K regarding specific model features vere met by s Lecting models which provide a significan. overall conservatism in the r analysis'he assumptions ~de pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core peaking factors, the containment pressure, and the performance of the ECCS.Decay heat generated throughout the transient is also conservatively calculated. The pressurizer heaters are not assumed to operate during the large and small break LOCA analyses.During the blovdown depressurizacion phase of the large break LOCA transient, Liquid flashing in the pressurizer Eorces a very rapid surge of mass out of the pressurizer, leaving only steam vithin a Eev tens of seconds.The effect of the pressurizer heaters being energized vould be to transfer some heat to the fluid surging out oE the pressurizer. Higher enchalpy Eluid from the pressurirer mixing vith the broken loop hot leg fluid will result in a very smalL pressure increase in the RCS during the period of time that fluid is surging from the pressurizer. The impact of modelling pressurizer heaters during the Large break would be to extend the end of blowdown time by a very small amount The integrated heat transfer from energized pressurirer heaters nver che period of depressurization blovdown is~in the order ot one sixth nf the metal heat release Erom the upper head.Fuel heat release is siicnilicantly ltreater than the heat input from pressurizer heaters.Additionally, the break flow controls the rate of system depressurization. Overall, pressurizer heaters should have a negligible effect~n the Large break transient results.The loss oE off"site pover assumption has been shown in Reference 15.2.8-1 to result in more Limiting peak cladding temperatures since the reactor coolant pumps lose po~er.The pressurizer heaters vould noc be energired during the'.uss if off site power.Small break LOCA's result in a slow system depressurization characterczed by discincc mixture leveLs within the-system.ALL smaLL break LOCA's rely upon the steam generators for some decay heat removal.In fact, the primary RCS pressure wtlt depend upnn a balance between the amount uE energy puc inco che system from core decay heat and metaL heat and the amount of energy that is removed from the primary by heat transfer through the steam generators and energy removal through the break.IE an additional heat source vere present, the syscem would respond by resulting in a very slightly higher RCS pressure 15.6.5-8 during that time that the heat source was active.".)e higher pressure would result in a higher break flow during the time the heat source is active.This may result in an earlier core uncovery and higher peak cladding temperature although the system tends to compensate for changes in the mass flow.A higher break mass flow rate results in m re mass removal and an earlier core uncovery.This also results in an earlier transition to two phase and steam break flow which causes earlier accumulator injection for a shorter duration of core uncovery.The earlier core uncovery at hi ner decay heat levels tends to increase peak cladding temperatures while the shorter duration of uncovery tends to decrease peak cladding temperature. The loss of off site power has also been determined to be limiting in terms of peak cladding temperature for small break LOCA's.Pressurizer heaters are not operable during the loss of off site power.Other systems which are not operable for the loss of off site po~er are the reactor coolant pumps and the steam dump control system.FSAR small break LOCAs result in the secondary pressure rising to the secondary safety valve setpoints~If the steam dump control system were operable, the RCS pressure would be significantly reduced giving a considerable ben<fit in terms of peak cladding temperature. r So the effect of pressurizer heaters being energized during a large break or small break LOCA would have a smal'ffect on the results and the more limiting assumption of the loss of offsite~ower precludes operability. 15.6.5.3.3 Results Lar e Break Results Based on the results of the LOCA sensitivit'y studies (References 15.6.5-20 through 15.6.5-22), the limiting large break was found to be the double-ended cold leg guillotine (DECLC).Therefore, only the DECLC break is considered in the large break ECCS performance analysis.Calculations were performed for a range of Moody break discharge coefficients. The results of these calculations are summarized in Tables 15.6.5-1 and 15.6.5-3.Factors affecting break flow in a westinghouse PMR and the lower limit of break discharge coefficient based on experimental data are discussed in Reference 15.6.5-21'onclusions in that report are that a best estimate value of the Hoody discharge coefficient is about 0'and that varying the discharge coefficients from a maximum of 1.0 to a minimum of 0.4 covers all uncertainties associated with the prediction of the break flow in case of a guillotine type severance of~cold leg pipe.The position to limit the break discharge coefficient to that range has been reviewed and approved by the NRC.Therefore, analyzing a LOCA for break discharge coefficients less than 0.4 is not consistent with experimental data or with the established procedure for a 10CFR50 Appendix K evaluation of ECCS performance. ~A corn~as of the core 1 od transien Figures 15.6.-6;15.6.5-2 and 15.6.5-2 y d core av age r d temper re rynsients xg res 15.6-3 I)'6.5 2,'15'.3)sh ws that e 0..4ECLC 1 ds t.the m sev re z'efl d tra sient e to th hjgh'tial c ad tern ratur at th begin ing;re ood.e ho'ssembly calc ations al o ref ct the arne riatio of i xtia1 cl d'erature w reak size.How er', the ari'on of a clad t eeyerstur r the het sejnhly tel~let e~i'ess se'I~nf 15.6.5-9 SHNPP FSAR ower at eak an ea siz e t.obit t f r exh/by th6 ot rod~ith break'ze.The mass and energy release data for calculated peak clad temperature are the break resulting in the highest presented in Section 6.2.1.5~0.8 DECLC bec e ightly higher d higher bio'age and hig r around.S'e sensitivi o pea clad mpe is ery'nc the ef ood tr ient sho s defx ite'mp veme with 1 ger b it'i E t a~t'the a 1ysis the th e br s esen ed sufEi p'dd n, u&xcient ists e a o o.khe sm v riat'o Figures 15.6.5-4 through 15.6.5-30 present the parameters of principal interest from the large break ECCS analyses.For all cases analyzed transients of the following parameters are presented: a)Hot spot clad temperature. b)Coolant pressure in the reactor core.c)Mater level in the core and downcomer during reflood.d)Core reflooding rate.e)Thermal power during blowdown.E)Containment pressure.For the limiting break analyzed, the following additional transient parameters are presented: a)Core flow during blowdown (inlet and outLet).b)Core heat transfer coefficients. c)Hot spot Eluid temperature. d)Mass released to Containment during blowdown.e)Energy released to Containment during blowdown.f)Fluid quality in the hot assembLy during bLowdown.g)Mass velocity during bl owdown.h)Accumulator ~ater Elow rate during blowdown.i)Pumped safety injection water flow rate during ref lood.4/4 8 The maximum clad temperature 'calculated for a large break is F which is less than the acceptance criteria limit of 2200 F oE 10 CFR 50.46.The maximum local metal water reaction is 2.37 percent which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46~The total core metal-water reaction is Less than 0.3 percent for aLl breaks, as compared with 15.6.5-10 the 1 percent'riterion of 10 CRF 50.46, and the clad temperature transient is terminated at a time Mhen the core geometry is still amenable to cooling.As a result, the core temperature Mill continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time Mill be provided.Full ECCS floM assumed in the LOCA analysis has not been noted to result in a higher peak clad temperature in 3 loop plants.Thus, the analysis presented remains conservative Reference 15.6.5-27 provides a moie detailed discussion. Small Break Re sul t s hs noted previously, the calculated peak clad temperature resulting from a small break LOCA is less than that calculated for a large break.Based on the results of the LOCA sensitivity studies (Reference 15.6.5-28), the limiting small break+as found to be less than a 10 in.diameter rupture of the RCS cold leg.In addition,.sensitivity studies have indicated little or no uncovering Mill occur for break sizes that are 2 in.or less.A range of small break analyses are presented Mhich establishes the limiting small break.The results of these analyses~are summarized in Tables 15.6.5-4 and 15.6.5"5.Figures 15.6.5"31 through 15.6.5-43 present the principal parameters of interest.for the small break ECCS analyses.'For all cases analyzed the folloMing transient parameters are presented: 15.6.5->oa SHNPP FSAR p/d Ch Afy68 a)b)c)d)RCS pressure Core mixture height.Hot spot clad temperature. Core power after reactor trip.For the limiting break analyzed, the following additional transient parameters are presented: a)Core steam flowrate.b)Core heat transfer coefficient. c)Hot spot fluid temperature. The maximm calculated peak clad temperature for all small breaks analyzed is 1808 F.These results are well below all acceptance criteria limits of 10 CFR 50.46 and in all cases are not limiting when compared to the results presented for large breaks.A complete spectrum of Small Break Loss of Coolant Accidents were examined in WCAP-9600,"Report on Small Break Accidents for Westinghouse NSSS~" The studies in that report indicated the maxiaum PCT occurred for the 3" break, thus the PCT does increase as break size decreases for the FSAR cases, but then H decreases as break sized decrease below 3~15.6.5.4 Radiological Consequences of a Postulated Lossmf-Coolant Accident 15.6.5.F 1 Containment Leakage To demonstrate in a conservative manner that the operation of a nuclear power station does not present any undue radiological hazard to the general public, a hypothetical accident involving a gross release of fission products is evaluated. No mechanism for such a release has been postulated because it would require a number of simultaneous failures to occur in'the engineered safety features.The following core fission product inventory is assumed to be released into the Containment as described in TID-14844'. 100 percent of the noble gases and 50 percent of the halogen'umerical values for the total core fission product inventory of the isotopes considered in calculating the radiation doses are listed in Table 15,0,9-1.The radiological evaluation of this accident is divided into two parts: internal (thyroid)dose" from inhalation of iodines in the containment leakage plume, and external (whole body)exposure as a result of immersion in the leakage plume.The integrated thyroid doses and the integrated whole body doses are calculated using methods and assumptions in conformance with Regulatory Gui,de 1.4.These assumptions are outlined below:.15.6.5-11 Amendment No.5 ~SHNPP FSAR TABLE 15.6.5"1 LARGE BREAK-TIME SE UENCE OF EVENTS EVENT OCCURRENCE TIME (SECONDS)DECLGo CD=0'4 DECLG, CD=0.6 DECLG, CD=0.8 Accident Initiation Reactor Trip Signal Safety Injection Actuation Signal Start Accumulator Injection End of ECC Bypass End of Blowdovn Bottom of Core Recovery Accumulators Empty Start Pumped ECC Injection 0.0~443 1.03 15.4 30.02 30.15 42.012 50.599+ocR 26.03 0.0.433.840 11.4 24.58 24.62 36.427 45.~25.840 0.0.427.74 9.12 21.89 21.89 33.890 5$'7 42%58 25'4 15.6.5-12 TABLE 15.6.5-1A Lar e Break Time Se uence of Events)Event Time (sec)1)Reactor trip signal;steam generator throttle valve closed signal;turbine trip signal 2)SI signal (on high containment pressure)(19.2 psia)3)Accumulator injection 4)Safety injection begins 5)Containment fan coolers begin 6)Containment spray begins Reactor Tri Si nal-occurs on compensated pressurizer pressure signal (1860 psia)Accumulator In'ection-injection begins Mhen RCS pressure drops to 600 psia.No failures assumed.Safet In'ection Si nal-occurs on high containment pressure (19.2 psia)~4 eD3 Safet?n'ection-There is a~second delay before injection begins.Delay is conservative and includes the folloMing: 2.0 sec signal generation time 14.5 begin charging floM 19.5 begin full SI floM 24.5 begin RHR flo~The limiting, single failure in westinghouse Evaluation Model Analyses is the failure of an RHR pump to start.Therefore, RHR pump operates in this analysis.See attached curve for SI flou during Morst large break transient.(Figure-15.6.5-18) Accumulator In'ection-Mhen pressure in RCS reaches 600 psia.See attached curve for accumulator injection floM no failures.(Figure 15.6.5-16) 15.6.5-12a TABLE 15.6.5-1A (Cont'd)Containment Heat RemovaL S stem.'n C-4 cooLers a oolers fan operate Initiated at 26.43 sec on High containment pressure signal at 1.03 sec DeLay oE 25.4 sec includes: 2.0 sec delay to start fan coolers 23.4 deLay to get power up 25.4 sec 10 sec diesel startup 5.3 sec sequences 8.1 sec fan coolers to reach EuLL.speed Fan coolers cooled by service water at 33 F (min).HEAT REHOVAL TABLE Temp (F)Q (BTU/sec)150 7208 3 180 220 12355.6~20930.6 258 29555'Containment S ra: Flow Temp=F PH total 25 Actuated on Hi-3 containment pressure (12 psig)NOTE: Spray reaches EuLL ELow 54.27 sec.25 15.6.5-12b Amendment No.25 TABLE 15.6.5"2 INPUT PARAMETERS USED IN THE ECCS ANALYSIS Core Power Peak Linear Power (Includes 102X factor)2775 Mwt (g,gl o l~h96 kW/f t.Total Peaking Factor FQ Axial Peaking Factor, F<Po~er Shape I,+709 7 Large break-chopped cosine Small Break-See Figure 15.6.5"45 Full Assembly Array Accumulator Water Volume (nominal)Accumul a t or Tank Volume (nomi na l)Accumulator Gas Pressure (minimum)Safety Injection Pumped Flow 17 x 17 1050 ft./accumulator 1450 ft./accumulator 600 psia See Figures 15.6.5-18 and 15.6.5"44 Containment Parameters See TabLes 6.2.1-62, 6.2.1-63, and Figures 6.2.1"303 and 6,2.1"304 Initial Loop Flow Vessel Inlet Temperature Vessel Outlet Temperature Reactor Coolant Pressure Steam Pressure 10073.71 lb./sec.558.1 F 622'F 2280 psia 952.0 psia Steam Generator Tube Plugging Level*2X is added to this power level to account for calorimetric error.15.6.5-13 SHNPP FSAR TABLE 15.6.5-3 LARGE BREAK Results Peak clad temperature (F)Location (ft.)DECLGR CD 0'DECLGR CD 0'l406 7~oo DECLG, CD=Oe8~ZO3 424 7.oa Maximum local clad/Mater reaction (Z)Location (ft.)6.0 3.'3 5~C.So Total core clad/Mater reaction (X)Hot rod burst time (seconds)Location (ft.)(~3 50.4 6.0 (~3 5Oe f<~3~57 F~4,.S-o 15.6.5-14 TABLE 15e6e5%SMALL BREAK LOCA TIME SE UENCE OF EVENTS 3 inc (See.)4 ine (See.)6 ine (See.)Start Reactor Trip Signal Top of Coze Uncovered r kccuuulator In5ection Bdgins Peak Clad Temperature Occurs Top of Core Covered OS 0 22.7 586 1380 1415 1487 OS 0 14.8 299 660 681 1331 0 0 10e 3 108 255 275 285 15.6.5-15 SHNPP FSAR TABLE 15.6.5-4A Small Break Time Sequence of Events (DECLG, CD~0.4)Event Time 1)Break occurs (3" small break)0.0 2)Reactor Trip Signal;steam generator throttle valve signal 18.5 3)Steam generator throttle valve closed (0.5 sec delay)4)Reactor trip occurs (4.2 sec delay)19.0 22.7 5)Normal feedwater floy begins to decrease 6)Sl signal setpt.reached 23.5 27.8 7)Normal feedwater flow stops 28.5 8)Steam generator safety valve low opening pressure reached 28.3 9)SI begins 10)huxiliary feedwater flow begins 11)accumulator in)ection begins 52.8 78.5 1380.0 No containment heat removal systems are modeled since a small break exhibits choked flow and thus the containment conditions would have no affec on the RCS.15.6.5-15a SHNPP FSAR TABLE 15.6.5-4A (Cont'd)Discussion of Events and Dela s for Small Breaks Reactor trip signal-setpt.reached at compensated pressuriser pressure 1860 psia 4.2 sec delay until trip occurs.Plant specific design limit.Steam generator throttle valve-0.5 sec delay time in closing-design criteria Steam flow to turbine stops after 0.5 sec after trip.Normal feedwater flow-signal to shut off flow is reactor trip signal.5 second delay until flow begins to decrease.Another 5 sec delay ro close valves complety.Steam Generator Safety Valves-Low opening pressure~1190.7 psia Flow at this pressure~101.95 lb/sec/S.G. High opening pressure 1287.2 psia Flow at this pressure~1274.7 lb/sec Linear variation in flow between these 2 pressures. These flows and opening pressures are plant specific.pressure~1760 psia.Assume 1 diesel fails to start 25 sec delay until safety in)ection begins.Delay conservative for all plants;includes: 2.0 sec signal generation time 14.5 sec charging flow begins 19.5 sec'I flow begins 24.5 sec RHR flow begins SI flow during break plot attached (Figure 15.6.5-44). huxiliar Feed'low-52.77 lb/sec/steam generator Begins 60 sec after trip signal at 18.5 sec~78.5 sec Standard aux.feed start time.Accumulator In ection-begins when RCS pressure reaches 600 psia.Occurs at 1380 sec.15.6.5-15b SUPP PSAR TABLE 15.6.5-5 SMALL BREAK LOCA RESULTS Resul ts 3 in>>Peak Clad Temp., F Peak Clad Location, ft.Local Zr/bpO Reaction, (Max)X Local Zr/820 Location, fthm Local Zr/820 Reaction, X Hot Rod Burst Time, sec.Hot Rod'Burst Location, ft.1808 11>>75 4.16 ll>>75<0.3 N/A N/A 1401 ll>>0 0.44 ll~25<0.3 N/A N/A 1119.11.00 0.3745 1 1>>00<0.3 N/A N/A 15-6.5-16

REFERENCES:

SECTION 15o6o T W.T.et al.,"LOFTRAN Code Descziption, WCAP-7907, 15.6.1-1 Burnett, T.~~~e a~~June 1972.15+6.5-1 15'.5-2 15.6.5-3 15.6.5-4 15i6~5-5 15e6o5-6 15-7 15+6.5-8 15.6.5-9 15+6~5-10 15@6+5 11 15e6.5-12"Acceptance Criteria for Emergency Coze Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix R of 10 CFR 50.Federal Register, Volume 39, Number 3, January 4, 1974.Reactor a e y Re S f ty Study-hn Assessment of Accident Risks in U.S.C 75/014 Commercial Nuclear Power Plants," WASH-1400, NURE g i October, 1975.Bordelon, F.M., Massie, H.W.and Zozdan, T.A., Westinghouse ECCS Evaluation Model-Summary," WCAP-8339 (Non-Proprietary), July, 1974.Bordelon, F..M., et al., SATAN-VI Program: Comprehensive Space-Time Dependent Analysis oi Loss of Coolant," WCAP-8302, June, 1974 (Proprietary) and WCAP-8306, June, 1974 (Non-Proprietary)

~tally, R.D.et al., Calculational Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code)," WCAP-8170, June, 1974 (Proprietary) and WCAP-8171, June, 1974 (Non-Proprietary)

Bordelon, F.M.and Murphy, E.T"Containment Pressure Analysis Code (COCO)," WCAP-8327, June, 1974 (Proprietary) and WCAP-8326, June 1974 (Non-Proprietary)

~~F M t al LOCTA-XV Program: Loss of Coolant~y and Transient Analysis'CAP 8301~June)1974 (Proprietary) a WCAP-8305, June, 1974 (Non-Proprietary)

~PWR FQCHT final Report, WCAP-7931, October 1972.Bordelon, fe Mo~et al i Westinghouse ECCS Evaluation Model" Supplementary Inforastion," WCAP-8471-P-A, April, 1975 Westinghouse ECCS Evaluation Model October 1975 Version, WCAP-8622, November 1975 (Proprietary)

~and WCAP-8623, November)975 (Non-Pr'prietary)).

Letter from C.Eicheldinger of Westinghouse Electric Corporation to D.Bi Vassallo of the Nuclear Regulatory Coaaiseionf Letter Number NS-CE-924 dazed January 23, 1976.Eicheldinger, C., Westinghouse ECCS Evaluation Model, februarY 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-h (Non-Proprietary Version), february 1978.

REFERENCES:

SECTION 15.6{Cont'd)15.6.5-13 Letter from T.M.Anderson of Westinghouse Electric Corporation to John Stole of the Nuclear Regulatory Commission, Letter Number NS-TNA-1981, November 1, 1978.15.6.5-14 15,6.5-15 Letter from T.M.Anderson of Westinghouse Electric Corporation to to John Stole of the Nuclear Regulatory Commission, Letter Number NS-TMA-2014, December 11, 1978.Porsching, T.A., Murphy, J.H., Redfield, J.A., and Davis, V.C.,"FLASH-4: A Fully Implicit FORTRAN-IV Program for the Digital Simulation of Transients in a Reactor Plant," WAPD-T.'i-840; Bettis Atomic Power Laboratory (March, 1969).15.6.5"16 Esposito, V.J., Kesavan, K.and Maul, B.A.,"WFLASH-A FORTR'N IV Computer Program for Simulation of Transients in a Multi-Loop PWR," WCAP"8280, Revision 2, July, 1974 (Proprietary) and WCAP-8261, Rdbision 1, July, 1974 (Non-Proprietary).

15.6.5-17 Skwarek, R., Johnson, W.Mayer, P.,"Westinghouse Emergency Core Cooling System Small Break October 1975 Model," WCAP-8970-P-A (Proprietary) and WCAP-8971 (Non-Proprietary), hpril 1977.15.6.5-18 Letter from T.M.hnderson of Westinghouse Electric Corporation to John Stole of the Nuclear Regulatory Commission, Tatter Number NS-TMA-20'30, January, 1979.15.6.5-19"Westinghouse ECCS Evaluation Model Sensitivity Studies,"%CAP-8341, July, 1974 (Proprietary), WCAP-8342, July,)974 (Non-Proprietary).

15.6.5-20 Julian, H.V., Tabone, C.J., Thompson, C.M.,"Westinhouse ECCS Three Loop Plant (17 x 17)Sensitivity Studies," WCAP-8853, September, 1976 (Non-Proprietary).

15, 6.5-21 15.6.5-22 Salvatori, R.,"Westinghouse ECCS-Plant Sensitivity Studies," MCAP-8340, July, 1974 (Proprietary) and WCAP-8356, July, 1974{Non-Proprietary).

Buterbaugh, T.L., Julian, H.V., Tome, h.E.,"Westinghouse ECCS Three Loop Plant (17 x 17)Sensitivity Studies," WCAP-8572, July, 1975 (Proprietary) and WCAP-8573, July, 1975 (Non-Proprietary).

)5.6.5-23 15.6.5-24 15.6.5-25 Murphy, K.G., Compe, K.M., Nuclear Power Plant Control Room Ventilation System Design For Meeting General Criterion, 13TH hEC hir Cleaning Conference (1973)~Stole, J.F., Letter to T.M.hnderson (Wescinghouse), Transmitting Safety Evaluation Report of Westinghouse ECCS Evaluation Model, February, 1978 Version, August 29, 1978.Stole, J.F., Letter to T.M.hnderson (Westinghouse), Transmitting Safety Evaluation of Wesginghouse ECCS Small Break, October, 1975 Model, June 8, 1978.

REFERENCES:

SECTION 15.6 (Cont'd)15.6.5-26 HcFetridge, R.H., D.C.Carner,"Study of Reactor Vessel Upper Head Region Fluid Temperature," WCAP-9404, Rev.1, December 1978.15.6.5-27.

Letter from P.Rahe of Westinghouse Electric Corporation to~~~~R.Tedesco of the Nuclear Regularory Commission, Letter Number NS-EPR-2538, December 12, 1981.15.6.5-28 15.6.5-29"Report on Small Break Accidents for Mestinghouse NSSS," MCAP-9600.

MASH 1258."Numerical Guides for Design Objective and Limiting'onditions for Operation to Hect the Criterion As Lou As Practicable for Radioactive Material in Light"Water-Cooled Nuclear Power Reactor Effluents." Volume 2, July 1973, U.S.Atomic Energy Commission.

15.6.5-30 ORNL-TH-21?, Part IV,"Design Considerations of Reactor Containment Spray Systems.~Calculation of Iodine-Mater Partition Coefficients." L.F.Parsly, January 1970, U.S.Atomic Energy Commission.

15.6.5-31 Collier, C., et.al.,"BART-Al: A Computer Code for the Best-Estimate Analysis of Ref lood Transient," WCAP-9561, January 1980.(Westinghouse Proprietary).

15.6.Kab i, J C e for 984 (Mes R., g al~, aly s of P in ouse Pro"BASH'R.Loss--Co rie'.y).teg'rated o e and Ref l l ht cid ts," CA-10 6, 0 B L 0 W D 0 W N R E L 0 0 D bREAK OCCURS REACTOR TRIP ICOMPENSATED PRESSURIZER PRESSUREI PUMI'ED SAFETY INJECTION SIGNAL IHI I CONT.PRESS.OR LO PRESSURIZER PRESS.I PUMPED SAFETY INJECTION BEGINS IASSUMING OFFSITE POWER AVAILABLEI ACCUMULATOR INJECTION CONTAINMENT HEAT REMOVAL SYSTEM INITIATION (ASSUMING OFFSITE POWER AVAILABI.EI END OF BYPASS END OF BLOWDOWN~UMPED SAFETY INJECTIUN BEGINS IASSUMING LOSS OF OFFSITE POWERI r BOTTOM OF COfC RECOVERY CONTAINMENT HEAT REMOVAL SYSTEM INITIATION IASSUMING LOSS OF OFFSITE POWERI ACCUMULATORS EMPTY CORE QUENCHED L 0 N G T E R M C 0 0 L I N G SWITCH TO COLD LEG RECIRCULATION ON RWST LOW LEVEL ALARM IMANUAL ACTIONI SWITCH TO LONG TERM RECIRCULATION IMANUAL ACTIONI SHEARON HARRIS NUCLEAR POWER PLANT Carolina Power 5 Light Company FINAL SAFETY ANALYSIS REPORT SEOUENCE OF EVENTS FOR LARGE BREAK LOSS-OF-COOLANT NALYSl S FIGURE 1'5~6.5" 1

R e>C n ce rX~o r n>>v)~E Ol g-3~CO tlat<g 0~h SOD THEIDIAI,, NECHLHICLI CONDItiONS DURING bQNRWH I ROD CEOIIETRY I HEAT TRANS EER COEPPICIEL,~~TRNEER CONDITIONS DURING REFLOOD END OF BlDQQggg REPILLiREPMOD (ROB)rn>0 XOÃ>.SZ Pl PI>Ql X 1l ZO OPl 0 Pl 0 rm Itl A n O N'Il Cl g)Ul fll ShTAH NLSS, aHERGY aEmaSE CONDITIONS NIRIHQ bMHDCNH CONDITIONS AT Eob I I I CORE IN~~e INLY EHTHLLPY I I I I I I'6tEFMOD/COCO IlIREE hOOD CLlCllLATES bRELX IILSS g ENERGY%ÃIXLSE I I I I I I I I I t COCO CAWUIATES COIITLIHNEHT NSSURE I I I I I I I I I I I I

W F L A S H L 0 C T A CORE PRESSURE, CORE FLOW, MIXTURE LEVEL, AND FUEL ROD POWER HISTORY 0 TIME CORE COVERED SHEARON HARRIS NUCLEAR POWER PLANT Carolina Power Si Light Company FINAL SAFETY ANALYSIS REPORT lHTERFhCE DESCRlPTlON FOR 5RALL BREhK l%OEL FIGURE Is.e.5-3 2 I 2 tn>OOI~~a R~2~~a 02 III~III~II III t)I~CO m~2 0 II SHEARON HARRIS (CQL)0.4 DECLG BART-LOCTA CLAD AVG.TEMP.HOT ROD BURST, 6.00 FT()PEAK, 6.75 FT()n r O I O n r Gl~2888 IJ 1688 8 I y 1888 I P 688 d 68 88 188 128 148 168 168 288 228 248 268 11IIC I SCCI 04/30/86 2500 2000 1500 K 1000 0 0 10 TIME (SEC)30 SHEARON HARRIS NUCLEAR POWER PLANT Carohna Power 5 Light Company FINAL SAFETY ANALYSIS REPORT Ar CY CORE PFE55'JRE-DECLG (CD=0./)F IGURE I5.6.5-26 z r CO O gn~f Pt 0 rn~ttl~-3 Ul tl 4s m<'0 0 D K C O tll rZ tll tll mO OZ m+D+Ch z mg mr ro-SD o OH n>rz p Cl I&em~D D 28-l2.8 LI i1.5 1.5 8 ee tee lse gee 25d 588 558<88 use 588 Sse ilttC tSCCI Ul al Ul I gl X'M O~r g C)7 r<cn~-9 Ch D 4s m w O X C men r mm A mO OZ o m~M 2\P I QO 0 o Z~mlfl h4 I-" n~I.S X L 5d led les 2ee 251 sbd 558 dies 45e SCO 556 Tl~l5CC)

1.75 1.50 1.25 O~l)00 O 0.75 0.50 0.25 0 0 10 TIME (SEC)20 SHEARON HARRIS NUCLEAR POWER PLANT Carolina Power 5 Light Company FINAL SAFETY ANALYSIS REPORT CORK POMER TRANSIENT DKCLS (CO=OsÃ)FIGURE 8 I5~6~5~

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z I O m 8~r g lO r n+o lh l/l a 17 0 R A Eh r y mm OZ a m 0 g)I+CA z ddbE~S bddE AS O dbdE~S bdbEiS~ldbbE S 2.S S 7.S lb.l2,S lS llaS 2d 22 5 25.2>.5 bb S2.S T lhE t SEC I Ul~'ll Ol e I m tA

Z tn~O r Z~~~n O lh a m w 0 37 Z C O Eh rZ mm A mO OZ a m s~37 r-z bdbE~8~SbbEib dddE'8 ISbbE F 8 i IdbbE.8 8 O bbbE~7 8 2.S S.1.5 lb.I2.5 IS.I2.5 28 22.S 25.2t.5 Sd.52.5 TIJOU ISECI Ul n m Z M O Ill y I Z~~n<o CII th m'0 O Z C O Ch I X m III n em mO OZ Ill II g)~CO Z II I O O.C r 51 2 5 O Pl A f A)68 IIII'SCC)Ul Ol*0 Ul C fll Ul

?Z e>C o cs rX m~mm~~n~+~o Z~~+OZ 5 m~Ch VI 0+CO mZ Z 0~~Ck M Vl~~SN6 b>Pl 0 pw A)c leee b K 2 5 S 7.5 III.I 2.5 IS.I W.S 28 22.5 25.2l.5 30.32.5 Tlttf ISECI Ul tA 0 Ul, C I m Ol

z r z cn~C Ocn r r 1 l o z~~<oz rO~m CO g I7 o th D~CO mZ z 0 L7 Ul Ul C Pl I O O I O Pl A I o C8 28 8-28 s-48 8 9-88>-88-Iee IBB I88 Y tME ISECI I82 Ul A Ol Ul 37 m Z I Cn~Cl C m~I Z~7~o+o CII Ul u m~O Z C n ol I mm n xm a OZ Iz XI'll~~CO Z 10~I~Og)4'5~r ill A A Ul I O C U X A O O 0 CJ Llj~CO 6 0 0 40 80 120 160 200 TIME (SEC'I 240 360 Ul Cll Ul.C tll a)

z~-g tn~C neo rx"~O D~o z~~o oz I"~<Z rO>+Z CO DD'0 D D5 0 D'%g 0 e//P Q I R P I$88 y I088 I P Sdb d g o..'28<8 68 88 ldd I 28 I<8 I68 I dd 284 228 2~8 26d TINE ISECl~'ll Ol Ul C lTl lO

z I g CO~C oco rX m 5 mm~I"o~>"~O OZ C~o>>m-3 g y Vl m w O Il 2500 J 2000 8 1500~1000 I O Cl I Gl 500 0 10 TIME (SEC)Ul gl~CI e g yJ m o R r N~C Ace~g rX m~mm~o n~mx C ol g r+D Q CA m<0 o nr A2)0 l5.~\(!2.5 lI.7.$n O O 2+5 I 58 l88 l58 Pdd 258 588 558>88 458 588 SSI Tl%(SIC> Z Z e~C neo rx m~mm~o Z~+OZ~he m o)~Vl D~CII m<Z O D 20 QO O Cl Ol~$5 Z CC IX D 4.O e Sd Ibd)Se 208 2Se sbe 588~le 45e See SSS i JkC ISECl z ce>c oem n~Cg n m O~o Z~>o OZ f)0 illy cn~3 cog g~CO IlS<0~1.75 n O fg O g O P 1.25 O~1.00 O 0.75 0.50 0.25 0 0 10 TIME (SEC)15 20.

Z 2 e p C o cs rZ rn m gn~~o~g pO Z~~+OZ~<no m+e tn t)PX7 g CA mc 0 Il U CO 20 K D a.15 z Z 4 10 I Z 0 O 0 0 80 120 200 TIME{SEC)240'320 360 Ul P n e c I m R 4n>C Oco~o Z~~+OR 0 m2'In g m 4 R 0 7,0 S I R P I))g IOOO I P$88 d 8 2d 48$8 88 I 88)28)48 I$d)88 208 228 248 2$8 TIME)SEC)m~'Tl n m Ul z co>p m 8 I r O C p Ch CO m D z lll C D 0 D X C A Ol rZ Fll m O~o OZ a m DD r~EO z 1500 K D 1000 K I a Pl A r Cl 0 0 10 TIME (SEC)Ul~ll P a Ul D Ng m

z I c neo rZ mC mm g pe<r n~m+Ih+Ut tt p+CO m<Z It 28m'37>m mr XO O ro m c~'al UlZ Ul 4~Atn n I I Gt I lloS lb.(I?oS Ids'7 5 It fn Ct~go-a O e z n S.2.5 8 58 Idd ISd 2dd s58 dbd.558<dd 458 588 55d Tlat%ISECI Ul p Ot Ul I lg m

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Z I O O~n~-3 Ol a D g m c'0 O C O Ce F mm vO OZ m~CO Z 2.00 1.75 1.50 A o Pl 9 o K Pl Z Ul 1.25 0 1.00 0 0.75 0.50 O f Gl O cA 0.25 0 0 10 TiME{SEC)15 20 25 N Cll C l~m 5l~

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0 3000F tL O 1000 C 0 0, 100 TIME tSEC)SHEARON HARRIS NUCLEAR POWER PLANT Carolina PoNrer Sc Light Company FINAL SAFETY ANALYSIS REPORT CORE AVERAGE ROO TFP%ERATVRE DECI.S I CD=0 e P)F IGURE 15,Co5 50h 13000 F La 0~~1000 C I 100 TIME (SEC)SHEARON HARRIS NUCLEAR POWER PLANT Carolina Povirr&Light Company FINAL SAFETY ANALYSIS REPORT CORE AVERAGE ROD TFPlPERATLÃE DECI.G (CD=0~$L')F IGURE r5.6.5-OOa

1400o F H///100 T)PAE (SEC)CORE AVERAGE ROO TEPPERATt'RE OKCLG (CO-"O.g)

8WCmegP Z.SHRPP~/1 c',trig FES$86 3/4.2.POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE WOOF>m tZme un LIHITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFO)shall be maintained within the following target band (flux difference units)from the target AFO value: a.k 5X for core average accumulated burnup of less than or equal to ooo~MWO/MTU;and brac b.+3X,"12K for core average accumulated burnup of greater than 3888.MWO/MTU.The indicated AFO may deviate outside the above required target band at greater than ar equal to 50K but less than 9(C of RATED THERMAL POWER provided the indi-cated AFO is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed 1 hour during the previous 24 hours.'he indicated AFD may deviate outside the above required target band at greater than LSX but less than 50K of RATED THERNL POWER provided the cumulative .penalty deviation time does not exceed 1 hour during the previous 24 hours.APPLICABILITY: MODE 1, above 15K of RATED THERNL POWER."~ACTION: With the indicated AFQ outside of the above requfred target band and with THERMAL POWER greater than or equal to 9'f RATED THERNL POWER, within 15 minutes either: 1.Restore the fndicated AFD to within the target band limits, or 2.Reduce THERNL POWER to less than 90%of RATED THERNL POWER.b.With the indicated AFD outside of the above requfred target band for more than 1 hour of cumulative penalty deviation time during the previous 24 haurs or autsfde the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 9(C but equal to or greater than SCAN of RATED THERMAL POWER, reduce: 1.THERNL POWER to less than 5QX of RATED THERNL POWER~ithin 30 minutes, and 2..The Power.Range Neutron Flux&%-High Setpofnts to less than or equal to MC of RATED THERMAL POWER within the next 4 hours."See Special Test Exceptions Specification 3.10.2.""Survef11ance testing of the Power Range Neutron Flux Channel may be performed pursuant ta Specfffcatfon 4.3.L.1 provider.'he indicated AFO is maintained within the Acceptable Operation Limits of Figure 3 2-1.A tatal of 16 hours operation may be accumulated with the AFO outside of the above required target band during testing without penalty dev4atfon. SHEARON HARRIS-UNIT 1 3/4 2-1

POWER OISTRIBUTION LIMITS~>E",~~~~~i~PROOF AND RDtH 0uPY 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F (Z)LIMITING CONOITION FOR OPERATION 3.2.2 F~(Z)shall be limited by the following relationships: 28 F~(Z)<[2.4C][K(Z)]for P>0.5 Fq(Z)<t,(4v64)l EK(Z)3 for P<0.5 Where: THERMAL POWER , and RATEO THERMAL POWER K(Z)"-the function obtained from Figure 3.2-2 for a given core height location.APPLICABIL STY: MOOE 1.ACTION: With F~(Z)exceeding its limit: a.Reduce THERMAL POWER at least IX for each IX F~(Z)exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours;POWER OPERATION may proceed for up to a total of 72 hours;subsequent POWER OPERATION may proceed provided the Overpower hT Trip Setpoints have been reduced at 1east I.".for each 2X F~(Z)exceeds the limit.b.Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above;THERMAL POWER may then be increased provided F~(Z)is demonstrated through incore mapping to be within its limit.SHEARON HARRIS-UNIT 1 3/4 2-5 ).2500 f.0000 O.~5OO I P 0.5000 0.25M TOTAL FQ 2.28 CORE HEIGHT 0.000 6.000 10.870 12.000 i K{Z)1.000 1.000 0.939 0.658 0.0 C)C7 ED AJ ED C7 40 CORK HE NACHT (FT)Pt~~3a 2 2.Cl C3 CI

3/4.2 POWER DISTRIBUTION LIMITS PROOF AHO IIEjjEN COPY BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1)maintaining the minimum ONBR in the core greater than or equal to 1.30 during normal operation and fn short-term transients, and (2)limiting the fission gas release, fuel pellet temperature, and cladding mechanical proper ties to within assumed design criteria.In addition, limiting the peak linear power density during Conditfon I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200oF is not exceeded.The definitions of certain hot channel and peaking factors as used in these specifications are as follows: F~(Z)Heat Flux Hot Channel Factor, fs defined as'the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allotting for manufacturing tolerances on fuel pellets and rods;Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power'to the average rod power;and F (Z)Radf'al Peaking Factor, is defined as the ratio of peak power density to average power densfty in the horizontal plane at core elevation Z.3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFQ)assure that the F~(Z)upper bound a.a 8 envelope of 400&times the normalized axial peaking factor fs not exceeded during either normal operation or in the event of xenon redistribution following power changes.Target flux difference (TARGET AFD)is determined at equilibrium xenon condi-tions..The rods may be positioned within the core in accordance with their respective insertion lfmits and should be fnserted near their normal position for steady-state operation at high power levels.The value of the target flux difference obtafned under these conditfons divided by the fraction of RATED THERMAL POWER fs the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER Ie'vels are obtained by multiplying the RATED THERMAL POWER value.by the appropriate fractfonal THERMAL POWER level.The periodic updating of the target flux difference value fs necessary to reflect core burnup considerations. SHEARON HARRIS-UNIT 1 8 3/4 2-1 PRDOF AHD REVIE'"/tlGP'f,>ge6 e I TNN SIOUhl SOh ILLU0ThATIOtt OttLT OO ttOT UCC SOh OSthATIOtt

I I 9 ILS0 I Teryn Slam Oitt~s I I I f l" t" I I I I I I I I~I f~i i r eg%+10%0+10%tg+og%NOICATtO AXNL SLUX OI SflltOOCE FIGURE B 3/4 2-E SHEARON HARRIS-UNIT 1 TYPICAL" INDICATED AXIAL FLUX DIFFERENCE YERSUS THERMAL POSER FOR BURNUP GREATER%N~HtttD/HTU QcoO B 3/4 2-3

Un R-oddeg.Allowable Fxi<versus Care Height X 2.88 1.38 88 Calc.Max Allow.X X~,~~v X: X 1.73 C7-1.b8 1.58 I~~~~X'" X x: X X X'----.-X X: X:X 1.6y X ,'X l.<8<7 L 5 b Core Height (ft)~F~y-I.l Plover e0~/z.<,'~D+C confro)r0 J5

Iexiono flkxP(REL) versus Core Height Quriog Normal Operation~1~11 f<<OOWtl ON11<<ttt+I>>1 0 Otto~~~Q).2.2f):~~~1>>~2'W W H N~1<<t~0<<At~W>>ft>>1 WH>>ft~OH 4OON 2 X<<11~ow~~ow X Y Xw>>1X ox'\o X~g~XXX~f~gX X: 0~~~~~~O}Ot X': XX X.-"iiv7i:i j)=.=i.M X: 1~t~t~10 0 0 0 0~t 4 11 11 t 1 1 1<<%0 0 0~A Htf~W<<t W f WQ H~~N.'()~]S)i.Ill 0 0 0 0 0 0 0 0 t>>>>H1 0 tt H ft>>>>>><<f 0>>0 0 i.K>>>>WA~>>1<<0>>Nttt ft>>e 11 0~W 111 ftt N N~1 OOHOO~1~~0~H>>>>H>>l l l I~Ott>>W~tttt>>1 1~NO N~1<<1~Ot~~0 0 0 0 0*A 0 0 0 0 0~0~OOOO......K.~4 11~~1 i 2 3 5 b I 8 9 i8 ii i2 Core Height (ft) l1 i': 86060600 Attachment 1

READABILlTY OF THE MCB SLBs, BPSLBs, AND TSLBs I Based upon the MCB profile and the NUREG-0700 anthropometric criteria, a fifth percentile female (eye height at 56.5 inches above the floor)would have a viewing distance to the top lens of the various light boxes (LB)that would be the hypoteneuse of a right triangle with side A equal to 83 inches minus 56.5 inches and side B equal to 25.5 inches.The top lenses of the LBs are located 83 inches from the floor.Top LB Lens C A=83-56.5~26.5 Eye B=25.5 The hy'poteneuse, C,.equals the square root of the sum of A-squared and B-squared. C=(26.5+25.5)T 36.8 inches A viewing distance of 36.8 inches would normally require a character height of 36.8 times.000 equals 0.107 inches for adequate discrimination. Characters in the LBs are currently engraved at 0.125 inches in height.This is approximately 85%of the recommended height (l5%under size).Readability of the LBs was reviewed by operators and human factors personnel and it was felt that, given adequate control over the other readability requirements of width, stroke width, spacing, and transluminance, the lenses were accurately readable from the normal operating positions. (3705ANS/ccc)

READABILITY OF THE MCB MLBs Characters within the Monitor Light Boxes (MLB)are below criteria;character height is~~~093.Based on the uses of the MLBs in the CR the HEDAT feels they are adequate.Operations views these lights to verify valves are properly actuated on SI signal.When an SI signal is actuated a bank of lights for each function will go on.The operators verify that all lights in each bank are on.Any light that is not on will be readily apparent to the operator.The primary indication of these valves is the position indication associated with each valve switch for every MLB status light in the control room.(37p5AWS/ccc)

Attachment 2

SHEARON HARRIS NUCLEAR POWER PLANT COLOR CODING MATRIX COLOR RED MCB ACP AEP-1 OPEN g RUNNING g BREAKER CLOSED DANGER i FIRST OUT i ALARM g ALERT BACKPANELS ERROR i POWER ON g RESET GREEN SHUT i STOP g BREAKER TRIPPED POWER ON Y'ELLOW/AMBER/ ORANGE CAUTIONS POST ACCIDENT MONITORING WARNING WHITE/BLUE INFORMATION I NF ORMATI ON 4 POWER ON

Attachment 3

AEP-I REVIEW A review of Emergency Operating Procedures was conducted by plant operations personnel to determine which steps require actions on the part of an operator to interface with the AEP-l.Each action was then reviewed to determine the necessity of using the AEP-l (e.g., are other controls or indicators used)and whether there were any consequences of either misuse or nonuse of the AEP-l controls or indicators. In addition, a review of all AEP-l indicators and controls was conducted to determine safety consequences of nonuse or misuse and whether the AEP-I control or indicator was the primary control or indicator for the required actions.For all AEP-I controls or indicators, excluding the reactor vessel and pressurizer vent valve controls, no safety consequences from misuse or nonuse were discovered. The reactor vessel and pressurizer vent valve controls are well labeled, demarcated, and are physically separated from the rest of the AEP-l controls.Additional enhancement/demarcation of the reactor vessel and pressurizer vent valve controls is under consideration. Attached are the results of this AEP-I review.(3750AWS/pgp ) AEP-I LIGHT BOXES ALB-023 (Annunciator Light Box)has coordinate axis labels and will be read from directly in front.Operations has no trouble reading it.DRPI (Digital Rod Position Indication) will be used during rod motion to recover dropped control rods and to verify a reactor trip.In all of the cases DRPI will be the secondary source of information, with the primary source being the ERFIS computer.SLB-08 (Status Light Box)B Train indication of RAB HVAC damper position.These dampers are slaved to the two RAB normal supply fans which trip on a SI signal.During normal fan operation, the ERFIS computer will be used to verify the damper positions as the fans are started.The Status Light Box indications will be used as a secondary indication. If the damper position is misread on a fan start, temperature alarms on the ERFIS computer and/or alarms on the RMS computer would indicate the problem.If such an alarm is received, the damper will be locally checked before any action is taken.There is no safety consequence of error from misreading any damper position.The same actions are taken on a SI signal, except it is for a fan stop and damper closure.SLB-10 A Train, same as SLB-08.SLB-09 B Train, indication of chilled water valves, HVAC dampers in the ESW intake structure, and HVAC dampers in the Fuel Handling Building (FHB).The indication will be used to verify proper automatic actions (open dampers or valves)during normal operation. There are no safety consequences from a misreading of the SLB since the indications have ERFIS temperature alarms to alert the operators that the automatic actions did not occur.If the alarms are received, the dampers or valves will be locally checked before any action is taken.SLB-I 1 A Train, same as SLB-09.(3705AMS/pgp) SLB-l2 Indication of HVAC dampers in the RAB.SLB-Ol2 provides indication of inlet and outlet dampers for various fans throughout the plant.These indications would be used to verify automatic actions during normal fan starts.Fourteen of these status lights are backups to the air flow indication of the AEP-Ol.The rest of the lights have trouble alarms that annunciate if the proper automatic actions do not occur.In either case, the only consequence of these automatic actions not occurring would be a fan trip.This would also give an alarm in the Control Room.If a fan does trip, it would be monitored upon fan restart.There is no safety consequence of an incorrect reading of these indications.(These indications are not used in the EOPs).(3705AM S/pgp)

AEP-l CONTROLS Sample Isolations (Steam Generator, Accum., RCS, PRZ, CNMT atmos, and CNMT sump)-All sample lines have redundant isolation valves so there are no consequences from opening the wrong isolation valve.The sample isolation valves are normally open with the only consequence of an inadvertent closure being a small delay in obtaining a sample.There are no safety consequences from inadvertently operating the control of a sample isolation valve.The sample isolation valves receive a Phase A cnmt.isolation signal.The primary means of verifying the Phase A closure is on the ERFIS computer.The back up means is with an extra operator in the control room after the initial phases of the transient are over.Post Accident Sample System (PASS)-The PASS is only mentioned once in the Emergency Operating Procedures (EOPs).This is in EOP-020 Step 7.This step says"Initiate Evaluation of Plant Status:...Obtain samples of RCS, SG, and CNMT Sump." The sample of the RCS would be obtained with the PASS.If the operator does not properly line up the PASS valves from the Auxiliary Equipment Panel One (AEP-OI), the result would be that the Chemist would not be able to draw his sample without delay.The Chemist would then notify the Control Room of his inability to draw the sample.Upon receiving this information, the operator would correct his valve line-up at AEP-Ol and the Chemist would then draw his sample.The basis for this procedure step (from the V/estinghouse Owners'roup) states"Since an evaluation of plant status may require some time to complete... it is initiated early in the recovery..." This sample would be used to help determine the long-term recovery actions.Should it require two hours (twice the normal time)to obtain this sample, it would still be completed before plant recovery was delayed.There is, therefore, no safety consequence from incorrect operation of this control.Stm Gen Bldn Isol-Same as the sample isolations except these receive a SI signal instead of a Phase A isolation. (3705AMS/pgp )

Chem Add to Stm Gen-These isolation valves must be opened and the ammonia and/or hydrazine metering pump locally started to add chemicals to the steam generators. If a valve is accidentally closed, the only consequence would be a delay in adding chemicals to the stm gen.There are no safety consequences from incorrectly operating the controls to these valves.These valves receive a feedwater isolation signal or a SI signal and are verified the same way as the sample isolation valves.RCDT-These are redundant isolation valves like the sample isolations and also receive a Phase A isolation which is verified in the same manner.Cnmt Fan Clrs-These valves isolate normal service water to the non-safety cnmt fan coolers.The only consequence for inadvertently closing one of these valves would be the loss of cooling water to the non-safety fan coo1ers.This would be detected by a cnmt ambient temperature alarm.These are redundant valves, so there would be no affect from inadvertently opening one of these valves.There is no safety consequence from incorrectly operating a control of the Cnmt fan coolers.These valves receive a Phase A signal and would be verified in the same manner as the sample isolation valves.Fuel Pool Cooling Pumps-Inadvertent stoppage of a pump would be immediately detected by a low flow annunciator. There would also be a high temperature annunciation before there was any danger of the pool overheating. If a pump was inadvertently started, the only consequence would be an increase in pool cooling flow.There are no safety consequences from incorrectly operating the controls of a fuel pool cooling pump.Chilled Water Isolation Valves-These normally-open valves isolate the non-essential portion of the chilled water system from the essential portion.The only consequence of inadvertent closure of a valve would be a high temperature alarm on the ERFIS computer.These valves close on a SI signal and will be verified in the same manner as the sample isolation valves.The operator must accidentally open (3705AWS/pgp )

four valves before there is any potential for a problem.There is no safety consequence. from incorrectly operating the controls of the chilled water isolation valves.Essential Chillers and Expansion Tanks-There are two redundant 100%capacity chillers and associated expansion tanks.If make up water is inadvertently stopped or started to either expansion tank, the results'will be a hi-hi or lo-lo level alarm.The affect of inadvertently stopping the running chiller would be the activation of several annunciators and eventual initiation of high area temperature alarms on the ERFIS computer.If an idle chiller is inadvertently started, the chiller's automatic control system would shut it off because there would be no service water to cool the chiller.There are no safety consequences from incorrectly operating the controls of a chiller.Both chillers start on a safety injection signal and would be verified in the same manner as the sample isolation valves.RAB HVAC-Local Air Handling Units (AHU)-These are two 100%capacity automatic cooling trains.There will normally be one train in operation with each AHU cycling on and off from the ambient temperature in each area.If a fan is inadvertently stopped, the'result would be a computer high temperature alarm for the respective area or the area fan automatically starting.If a fan is inadvertently started, the result would be the fan automatically stopping or extra cooling in a room in the plant.There are no safety consequences from incorrectly operating the controls of one of the AHUs.Both trains of the AHUs start on a SI signal and would be verified in the same manner as the sample isolation valves.RAB HVAC-Normal Supply and Exhaust Fans-There are two 100%capacity supply fans and four 50%capacity exhaust fans.Each fan has flow instrumentation and flow alarms that will alert the operator if the wrong fan is inadvertently stopped.Both supply fans and/or all four exhaust fans must be stopped for a period of time before there is any potential for a problem.If a fan is inadvertently started, the only (3705AWS/pgp)

consequence would be an increase in the air flow through the RAB.There are no safety consequences or potential for radiation release from incorrectly operating the control of a fan.These fans stop on a SI signal and would be verified in the same manner as the sample isolation valves.il RAB HVAC-Emergency Exhaust Fans-These are two 100%capacity exhaust fans used during emergency operation. Both fans must be stopped for a period of time before there is any potential for a problem.If a'fan is inadvertently started during normal operation, the only result would be an increase in the air flow through the RAB.There are no safety consequences from incorrectly operating the controls of the emergency exhaust fans.Both of these fans start on a Sl signal and would be verified in the same manner as the sample isolation valves.RAB HVAC-Room Exhaust Fans-Each room has two 100%capacity exhaust fans.Both fans must be stopped for a period of time before there is any potential for a problem.The only consequence of inadvertently starting a fan would be an increase in the air flow through the individual room.There are no safety consequences from incorrectly operating the control of a fan.RAB HVAC-Smoke Purge Fans-There are two smoke purge fans which are used after a fire is extinguished to assist in the recovery effort.During fan use, only fire brigade personnel will be in the affected area.The result of accidentally stopping a smoke purge fan would be an increase in the time needed to remove the smoke.If a fan is inadvertently started, the only result would be an increase in air flow through the RAB.There are no safety consequences from incorrectly operating the control of a smoke purge fan.Fuel Handling Building HVAC-There are two parallel trains of normal HVAC for the FHB.If a fan is inadvertently stopped, there is a high temperature annunciator for the spent fuel pool (SFP)area and ERFIS alarms for the SFP pump room temperatures. Accidental starting will only result in an increased air flow.There are no safety consequences of incorrectly operating the controls of the fans.(3705AWS/pgp) ESW Intake Structure HVAC-There is one train of HVAC for each train of ESW.These fans are used when the associated ESW pump is running or the room temperature is above 90'F.No single failure can disable both trains of HVAC.If one train is inadvertently stopped, it would be detected by an auxiliary operator on his normal rounds or by a high temperature alarm on the ERFIS computer.Continuous operation of these fans is not required so it is acceptable for these fans to be stopped until the operator makes his next set of rounds.If a fan is started, the only consequence would be an increased air flow through the building.There are no safety consequences from incorrectly operating the control of one fan.These fans start on a SI signal and would be checked in the same manner as the sample isolation valves.Diesel Fuel Oil Pump Exhaust Fan-There are two exhaust fans in each train.Both fans in one train must be stopped before there is any potential for a problem.If one fan is inadvertently stopped, it would be detected by an auxiliary operator on his normal rounds or by a high temperature alarm on the ERFIS computer.Continuous operation of these fans is not required so it is acceptable for these fans to be stopped until the operator makes his next set of rounds.There is no safety consequence from incorrectly operating the control of one of these exhaust fans.These fans start on a SI signal and would be checked in the same manner as the sample isolation valves.Pressurizer and Reactor Head Vent Valves-During normal operation, if these valves were incorrectly opened (two series valves must be open for an RCS release path), the Control Room would receive an alarm.This alarm would either be immediate or within approximately 30 minutes depending on which valves were opened.In either case, the alarm would be received before any action would need to be taken to correct the situation. The vent valves can be aligned to discharge to two different locations. The preferred location is to the Pressurizer Relief Tank (PRT).When there is vent flow to the PRT, there will be a"REACTOR VESSEL VENT FLOW" alarm on the main (3705AWS/pgp )

control board.This alarm will be used to verify flow and, therefore, a proper valve line-up.The second location, which should only be used if the PRT cannot be lined up, is to discharge to the CNMT atmosphere. EOP use of the head vent valves occurs in two different ways.The first use is as an RCS bleed path, that is to deliberately create a controlled hole in the RCS.The EOPs tell the operator to open all pressurizer (PRZ)PORVs and vent valves.If the vents are not properly lined up to the PRT and there is no flow, the flow alarm will not be received.If the vents are supposed to be discharging to the CNMT atmosphere (secondary source), but are not, the error would be comparatively insignificant because the PORVs should have about 25 times the flow as the vents.The second use of the vents is to release a gas bubble from the RCS (increase the water level in the vessel).Implementation of this procedure would be a very slow and deliberate process in which the water level in the vessel will be constantly monitored. If the valves are not properly lined up, the operator monitoring the vessel level will notice it is not increasing and stop the venting process to check the valve line-up for the vents.(3705AMS/pgp)

EMERGENCY OPERATING PROCEDURES ACCESSING AEP-I CONTROLS AND ANNUNCIATORS PATH I/30 FRP J.I/01 EPP 01/19"Verify" or"Check Phase A Isolation" The primary means of verifying or checking a proper Phase A isolation is on the ERFIS computer.The back up means is with an extra operator checking each valve position indication in the control room.EPP 01/09 EPP 13/01 EPP 10/00 EPP 15/02"Check" or"Verify SG status...blowdown...sample line isolation Closed" or"check any valves...as a possible coolant loss flow path" or"Check Secondary Pressure Boundary...blowdown...sample...hydrazine and ammonia-Closed". The primary means of verifying these automatic isolations is with the ERFIS computer.The back up means is with an extra operator checking each valve position indication in the control room.FRP C.I/10 FRP C.I/IS FRP C.2/03 FRP C.3/03 FRP H.l/15 FRP I.3/19 EPP Ol/03 Close"Reactor vessel vent...PRZ vent valves" or"Open reactor...PRZ vent valves" or"Align one reactor vessel vent..." These controls are well labeled, demarcated, and are physically separated from the rest of the.controls of AEP-OI.(3705AMS/ccc)

FRP C.3/02 FRP C.3/03 CAUTION EPP 01/27 EPP 20/02 Statement removed.FRP 3.2/02"Sample Sump V/ater for Activity Level, If Possible" It will take approximately 60 minutes to draw the sample and analyze it for activity.If the operator unisolates the wrong line, the chemist will call the control room and notify the operator that he can not draw a sample.The operator will then unisolate the sump sample lines.There is no urgency in drawing the sample and no safety consequence if the operator does not open the proper valves.EPP 01/22"Isolate CNMT" This is in EPP 01, which is only used during a loss of all AC power.The primary means of checking the CNMT isolation valves is with the ERFIS computer (operating on batteries). If a valve is found open, it will be checked against any control room indication (if operable w/o AC power)then locally closed in the plant.EPP-15/30"Realign Plant Systems for Normal Operations, As Appropriate" This step is performed after the accident has been terminated. It is the initial step in returning the plant systems to their normal status.EP P-16/01"Identify Ruptured SG" using steamline radiation monitor, sample for activity levels, unisolate blowdown and check for high radiation. The steamline monitors are the primary source of information to determine which SG has the ruptured tubes.The secondary means is by sampling for gross activity.It will take approximately 60 minutes to draw a sample and analyze it.If the operator unisolates the wrong sample line, the chemist will call the control room and tell the operator to unisolate the correct sample line.There is no urgency in drawing the sample and no safety consequence if the operator does not open the proper valve.(3705AWS/pgp)

EPP-20/07"Initiate evaluation of plant status...obtain samples of RCS, S/Gs and CV sump as needed" It will take approximately 60 minutes to draw and analyze each of these samples.These samples will help determine the plant's long term recovery method.There is no urgency in obtaining these samples if the plant Technical Support Center feels they are necessary, and there is no safety consequence in delaying the sample.(3705AWS/pgp )

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