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References
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{{#Wiki_filter:F I I l Attachment C PROPOSED TECNNICAL SPECIFICATIONS (strikeout for deletions and highlight for additions) San Onofre Unit 2 I t j 4 l 1 9906110178 990608 ~~ PDR ADOCK 05000361 P PDR

r-Boration Systems - Operating 3.1.9 SURVEILLANCE REQUIREMENTS ' SURVEILLANCE-FREQUENCY Verify)the boron concentration in the BAMU 7 days SR 3.1.9.1 tank (s is within limits. 1 SR 3.1.9.2- , Verify the volume of borated water 7 days containedintheBAMUtank(s)iswithin limits. SR '3.1.9.3'- ' Verify that each flow path is operable and 31 days thateachvalve(manual power operated or automatic,that1snotlocked, sealed,or '~ otherwise secured) in the above required flow paths is in its correct position. SR' 3.1.9.4 Verify thatLeach autcmatic valve in the 24 months above required flow paths actuates to its correct. position on an SIAS test signal. Mid3iR([(. yif{f $ K c G $ 5 [ N j M [jjlspE N @ )A?E5@jsj with6the. Inservice Testigg R,.,_ _, ogtag SAN ONOFRE--UNIT 2 3.1-21 Amendment No. 127~ .i. 3

ECCS -Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l I SR 3.5.2.4 Verify ECCS piping is full of water. 31 days SR 3.5.2.5 Verify the following ECCS pumps develop the In accordance indicated developed head and/or flow rate. r e 0 0 Full Flow Full Flow Mini flow Eggt GPM Head (Ft) Head (Ft) HPSI-P017 650 a 2142 HPSI-P018 650 a 2101 HPSI-P019 650 a 2103 i LPSI-P015 a 406.1 LPSI-P016 a 406.1 '!crify ce:h cher~Nump d;velops a flew SR 3.5.2.6 In accordance of i 40 gpm. N with the hservice Testing Program SR 3.5.2.7 Verify each ECCS automatic valve in the 24 months flow path actuates to the correct position on an actual or simulated actuation signal. SR 3.5.2.8 Verify each ECCS pump starts automatically 24 months on an actual or simulated actuation signal. SR 3.5.2.9 Verify each LPSI pump stops on an actual or 24 months simulated actuation signal. (continued) (continued) SAN ONOFRE--UNIT 2 3.5-6 Amendment No. 127

l MSSVs 3.7.1 Table 3.7.1-2 (page 1 of 1) MainSteamSafetyValves(LiftSettings) VALVE NUMBER LIFT SETTING

Steam Generator Steam Generator (psig) fl

2 2PSV-8401 2PSV-8410 1085**

2PSV-8402 2PSV-8411 1092 2PSV-8403 2PSV-8412 1099 2PSV-8404 2PSV-8413 1106 2PSV-8405 2PSV-8414 1113 2PSV-8406 2PSV-8415 1120 2PSV-8407 2PSV-8416 1127 l 2PSV-8408 2PSV-8417 1134 2PSV-8409 2PSV-8418 1140 The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. Each MSSV has an as-foundtoleranceof+2%/-3%. Following testing according to LCO 5.5.2.10,MSSVswillbesetwithin+/-1%ofthespecifiedliftsetpoint. ..i.}E..i! E.h.!$,.$. 9,5 5N. 3.,.5Y5,,.10 ''E "' '3-Io"" I 'It 36ttII'i GI I E3D Y s f. l (continued) j SAN ONOFRE--UNIT 2 3.7.4 Amendment No. 127 i )

~ l Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING. LIMITS REPORT (COLR) (continued) 3.b.2 Letter, O. D. Parr (NRC).to A. E. Scherer (CE), dated December 9, 1975 (NRC Staff Review of the Proposed Combustion Engineering ECCS Evaluation Model Changes) (Methodology for Specification 3.2.1 for Linear Heat Rate) i 4.a.1 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137P, August 1974 4.a.2 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 1-P, January 1977 1 4.b.1 Letter, K. Kniel (NRC)' to A. E. Scherer (CE), dated September 27, 1977 (Evaluation of Topical Report ' CENPD-133, Supplement, 3-P and CENPD-137, Supplement 1-P) (Methodology for Specification 3.2.1 for Linear Heat Rate) SE i 24E . 5!Id5MNF511PSE!!lf ~^~~~^~^ ~~ 5. " Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A, May 1988 (Methodology for Specifications 3.2.4 for Departure From Nucleate Boiling Ratio, and 3.2.5 for Axial Shape Index) c. The core operating limits shall be determined such that all a >plicable limits e.g., fuel thermal-mechanical limits, core tiermal hydraulic imits, Emergency Core Cooling i System ECCS) limits, nuclear limits such as SDM, transient analysi limits, and accident analysis limits) of the safety analysis are met. I (continued) i SAN ONOFRE--UNIT 2 5.0-28 Amendment No. 127 t

Attachment D PROPOSED TECHNICAL SPECIFICATIONS (strikeout for deletions and highlight for additions) San Onofre Unit 3 j j

Boration Systems - Operating ' 3.1.9 . SURVEILLANCE REQUIREMENTS SURVEILLANCE.' -FREQUENCY-Verify)theboronconcentrationintheBAMU 7 days -SR 3.1.9.1 tank (s _is within limits. SR 3.1.9.2'- Verify the-volume of' borated water 7 days containedintheBAMUtank(s)iswithin 11mi ts'. SR 3.1'. 9. 3 Verify that each flow path is operable and I 31 days that each valve (manual, power operated or automatic, that is not locked, sealed, or otherwise secured) in the above required 1 flow paths.is in =1ts. correct position. J l JSR' 3.1.9.4 Verify that each automatic valve in the 24 months-above required flow paths actuates to its correct position on an SIAS test signal. MEM}})l5 !HRjfpWh]M5}fgI@l(({0$$$] Is?AEWd55

withsthe, Inservice _ _ _

185tiMlf.togram I el (continued) i i SAN ON0FRE--UNIT-3 3.1-21 Amendment No. 116

ECCS -Operating l 3.5.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify ECCS piping is full of water. 31 days SR 3.5.2.5 Verify the following ECCS pumps develop the In accordance indicated developed head and/or flow rate, with the Inservice Full Flow Full Flow Miniolow Testing Program Eggt GPM Head (Ft) Headi;Ft) HPSI-P017 650 a 2093 HPSI-P018 650 a 2132 HPSI-P019 650 a 2099 LPSI-P015 a 396 LPSI-P016-a 396 SR 3.5.2.6 3rif3;cect, ct.efj ir,g,pa+ develops e flow ,5.jc gder.cc I exAIM -- c,- vr - c

.i nayj;,

?. s'_',' H..I.,I'L.. _3. 1 SR 3.5.2.7 Verify each ECCS automatic valve in the 24 months flow path actuates to the correct position l on an actual or simulated actuation signal. SR 3.5.2.8 Verify each ECCS pump starts automatically 24 months on an actual or simulated actuation signal. SR 3.5.2.9 Verify each LPSI pump stops on an actual or 24 months simulated actuation signal. (continued) (continued) SAN ONOFRE--UNIT 3 3.5-6 Amendment No. 116 L

l MSSVs 3.7.1 Main Steam Safety Val (ves (Lift Settings) Table 3.7.1-2 page1of1) l VALVE NUMBER LIFT SETTING

Steam Generator Steam Generator (psig)

1
2 3PSV-8401 3PSV-8410 1085**

3PSV-8402 3PSV-8411 1092 3PSV-8403 3PSV-8412 1099 3PSV-8404 3PSV-8413 1106 3PSV-8405 3PSV-8414 1113 1 3PSV-8406 3PSV-8415 1120 3PSV-8407 3PSV-8416 1127 3PSV-8408 3PSV-8417 1134 3PSV-8409 3PSV-8418 1140 The lift setting pressure shall correspond to ambient conditions of the j valve at nominal operating temperature and pressure. Each MSSV has an as-found tolerance of +2%/-3%. Following testing according to LC0 210,MSSVs,wi11, be, set,within +/-1% of the specif1ed,1ift, set ?oint._ 5:f? g .'f. '.'. '.' ".' ". m'. ". ?.."e s' " ' ' ". ' ' ' ', " ' ' " "" "" "" '"""" ' "" "3 ~ F"'" v. . r. f... i l l SAN ONOFRE--UNIT 3 3.7.4 Amendment No. 116

1 Reporting Requirements 5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3.b.2 Letter, O. D. Parr (NRC) ff Review of the Proposed. to A. E. Scherer (CE), dated December 9, 1975 (NRC Sta Combustion Engineering ECCS Evaluation Model Changes) (Methodology for Specification 3.2.1 for Linear Heat Rate) 4.a.1 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137P, August 1974 4.a.2 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 1-P, January 1977 September. Kniel (NRC) to A. E. Scherer (CE), dated 4.b.1 Letter K CENPD-133, Supplemen(Evaluation of Topical Reportt, 3-P ani 27, 1977 1-P) ) (Methodology for Specification 3.2.1 for Linear Heat Rate) [S!'I! l TilF6sFI(M WC'CIAHliiW '%) A05 ' ~ ~ 'I Seitak (SCA E[valsat#8QQ* *fg{gJ g I 7J.* Decemberg16(;199ff ~ ten Modeli3 TAC ^ the' e "~ ' ^ ^ ' ~ ^ ' ~ 5. " Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A,May1988 (Methodology for Specifications 3.2.4 for Departure From Nucleate Boiling Ratto, and 3.2.5 for Axial Shape Index) The core operating (limits shall be determined such that all c. applicable limits e.g., fuel thermal-mechanical limits, core thermal hydraulic limits. Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient anal sis limits, and accident analysis limits) of the safety anal sis are met. (continued) SAN ONOFRE--UNIT 3 5.0-28 Amendment No. 127

) Attachment E 1 PROPOSED TECHNICAL SPECIFICATIONS (Changes Incorporated) San Onofre Unit 2 i l

Boration Systems - Operating 3.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify)the boron concentration in the BAMU7 days SR 3.1.9.1 tank (s is within limits. SR.' 3.1.9.2 Verify.the volume of borated water 7 days containedintheBAMUtank(s)iswithin . limits. SR 3.1.9.3 Verify that each flow path is operable and 31 days thateachvalve(manual,poweroperatedor automatic, that is not locked, sealed, or otherwise secured) in the above required flow paths is in its correct position. SR 3.1.9.4 Verify.that each automatic valve in the 24 months above required flow paths actuates to its correct position on an SIAS test signal. SR 3.1.9.5 Verify each charging pump is OPERABLE. In accordance with the Inservice Testing Program (continued) SAN ONOFRE--UNIT 2 3.1 j

ECCS -Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE-FREQUENCY SR' 3.5.2.4 Verify ECCS piping is full of water. 31 days SR 3.5.2.5 Verify the following ECCS pumps develop the In accordance indicated developed head and/or flow rate. with the Inservice Full Flow Full Flow - Miniflow Testing Program Eyst GPM Head (Ft) Head (Ft) HPSI-P017 65') 'a 2142 HPSI-P018 650 a 2101 HPSI-P019 650 a 2103 LPSI-P015 a 406.1 LPSI-P016 a 406.1 i SR 3.5.2.6 Deleted l l SR 3.5.2.7 Verify each ECCS automatic valve in the 24 months i flow path actuates to the correct position i on an actual or simulated actuation signal. 4 SR 3.5.2.8 Verify each ECCS pump starts automatically 24 months on an actual or simulated actuation signal. SR 3.5.2.9 Verify each LPSI pump stops on an actual or 24 months simulated actuation signal. (continued)' (continued) SAN ONOFRE--UNIT 2 3.5-6

MSSVs 3.7.1 Table 3.7.1-2 (page 1 of 1) Main Steam Safety Valves (Lift Settings) VALVE NUMBER LIFT SETTING

Steam Generator Steam Generator (psig)

)

1
2

{ 2PSV-8401 2PSV-8410 1085 2PSV-8402 2PSV-8411 1092 2PSV-8403 2PSV-8412 1099 2PSV-8404 2PSV-8413 1106 2PSV-8405 2PSV-8414 1113 2PSV-8406 2PSV-8415 1120 2PSV-8407 2PSV-8416 1127 2PSV-8408 2PSV-8417 1134 2PSV-8409 2PSV-8418 1140 The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. Each MSSV has an as-foundtoleranceof+2%/-3%. Following testing according to LC0 5.5.2.10, MSSVs will be set within +/-1% of the specified lift setpoint. l (continued) SAN ON0FRE--UNIT 2 3.7.4

Reporting Requirements 5.7 5.7 Reporting Requirements (continued) i 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3.b.2 Letter, O. D. Parr (NRC) ff Review of the Proposedto A. E. Sc December 9, 1975 (NRC Sta Combustion Engineering ECCS Evaluation Model Changes) (Methodology for Specification 3.2.1 for Linear Heat Rate) 4.a.1 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137P, August 1974 4.a.2 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 1-P, January 1977 4.a.3 " Calculative Methods for the ABB C-E Small Break LOCA Evaluation Model," CEMPD-137, Supplement 2-P-A, April 1998 4.b.1 Letter, K. Kniel (NRC) to A. E. Scherer (CE), dated. September 27, 1977 (Evaluation of Topical Report CENPD-133,- Supplement, 3-P and CEhPD-137, Supplement 1-P) (Methodology for Specification 3.2.1 for Linear Heat Rate) 4.b.2 Letter, T..H. Essig (NRC) to I. C. Rickord (ABB), " Acceptance for Referencing of the Topical Report CENPD-137(P)iBreakLOCAEvaluationModel'(TAC Supplement, 2, ' Calculative Methods for the C-E Smal M95687)," December 16, 1997. 5. " Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A,May1988 (Methodology for Specifications 3.2.4 for Departure From Nucleate Boiling Ratio, and 3.2.5 for Axial Shape Index) The core operating (e.mits shall be determined such that allg., fuel therm li -c. a)plicable limits tiermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient anal sis limits, and accident analysis limits) of the safety anal sis are met. (continued) SAN ONOFRE--UNIT 2 5.0-28 I l

PCN-495 " Deletion of ECCS Credit for Charging Pumps" i 4 Attachment F PROPOSED TECNNICAL SPECIFICATIONS (Changes Incorporated) San Onofre Unit 3 'I i

ti, t Boration Systems - Operating 3.109 ' SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ~SR 3.1. 9.1' ' Verify the boron concentration in the BAMU 7 days tank (s)iswithin-limits. SR. 3.1.9.2 Verify.the volume of borated water 7 days contained in the BAMU tank (s) is within - limits. SR 3.1.9.3 - Verify that' each flow path is operable and 31 days automatic, that n(manualpower operated or that each valve s not Iocked, sealed, or otherwise secured).in the above required flow paths is in its correct position. SR 3.1.9.4 Verify that each automatic valve in the 24 months { above required flow paths actuates to its correct position on an SIAS test signal. SR 3.1.9.5 Verify each charging pump is OPERABLE. In accordance with the Inservice Testing Program 1 ) i (continued) SAN ONOFRE--UNIT 3 3.1-21 it j

m ECCS -Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued) . SURVEILLANCE' FREQUENCY SR 3.5.2.4 Verify ECCS piping is full of water. 31 days SR 3.5.2.5' Verify the following ECCS pumps develop the In accordance indicateddevelopedheadand/orflowrate. with the Inservice adhr5' HeaN[Fj Tesu ng Nogram Emut e HPSI P017 650 a 2093 l HPSI-P018 650 a 2132 HPSI-P019 650 a 2099 2 396 LPSI-P015 LPSI-P016 2 396 SR 3.5.2.6-Deleted l SR 3.5.2.7 Verify each ECCS automatic valve in the 24 months flow path actuates to the correct position on an actual or simulated actuation signal. 1 SR 3.5.2.8 Verify each ECCS pump starts automatically 24 months on an actual or simulated actuation signal..

SR 3.5.2.9 Verify each LPSI pump stops on an actual or 24 months simulated actuation signal.

(continued) (continued) SAN ONOFRE--UNIT 3 3.5-6

( l. ( l MSSVs l 3.7.1 Table 3.7.1-2 (page 1 of 1) Main Steam Safety Valves (Lift Settings) VALVE NUMBER LIFT SETTING

Steam Generator Steam Generator (psig)

1
2 3PSV-8401 3PSV-8410 1085 j

3PSV-8402 3PSV-8411 1092 3PSV-8403 3PSV-8412 1099 3PSV-8404 3PSV-8413 1106 3PSV-8405 3PSV-8414 1113 i 3PSV-8406 3PSV-8415 1120 3PSV-8407 3PSV-8416 1127 3PSV-8408 3PSV-8417 1134 3PSV-8409 3PSV-8418 1140 The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure. Each MSSV has an as-foundtoleranceof+2%/-3%. Following testing according to LC0 5.5.2.10, MSSVs will be set within +/-1% of the specified lift setpoint. l i SAN ON0FRE--UNIT 3 3.7.4

Reporting Requirements-5.7 5.7 Reporting Requirements (continued) 5.7.1.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3.b.2 Letter, O. D. Parr (NRC) to A. E. Scherer (CE), dated December 9, 1975 (NRC Staff Review of the Pro)osed { Combustion Engineering ECCS Evaluation Model ':hanges) (Methodology for Specification 3.2.1 for Linear Heat Rate) 4.a.1 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137P, August 1974 4.a.2 " Calculative Methods for the C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 1-P, January 1977 4.a.3 " Calculative Methods for the ABB C-E Small Break LOCA Evaluation Model," CENPD-137, Supplement 2-P-A, April 1998 September. Kniel (NRC) to A. E. Scherer (CE), dated 4.b.1 Letter K 27, 1977 (Evaluation of Topica Report CENPD-133, Supplement, 3-P and CENPD-137, Supplement 1-P) 4 (Methodology for Specification 3.2.1 for Linear Heat Rate) 4.b.2 Letter, T. H. Essig (NRC? to I. C. Rickord (ABB), " Acceptance for Referencin of the Topical Re) ort CENPD-137(P)i Break LOCA,E aluation M;Jel' (TAC Supplement ' Calculative Met lods for the C-E Smal M95687)," December 16, 1997. 5. " Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A,May1988 (Methodology for Specifications 3.2.4 for Departure From Nucleate Boiling Ratio, and 3.2.5 for Axial Shape Index) The core operating (e.mits shall be determined such that allg., fuel therm li c. a >plicable limits + tiermal hydraulic ~ imits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient anal sis limits, and accident analysis limits) of the safety anal sis are met. (continued) SAN ONOFRE--UNIT 3 5.0-28

Attachment G PROPOSED BASES CHANGES (for information only - strikeout for deletions andhighlightforadditions) San Onofre Unit 2 i

Boration Systems - Operating B 3.1.9 BASES (continued) SURVEILLANCE SR 3.1.9.1 and 3.1.9.2 REQUIREMENTS SR 3.1.9.1 verifies that the boron concentration of the available boric acid solution in the BAMU tanks is sufficient for reactivity control. SR 3.1.9.2 verifies that a sufficient volume of borated water is available for RCS makeup. The minimum required volume and concentration of stored boric acid in the BAMU tank (s) is dependent upon the RWST boron concentration and is specified in a Licensee Controlled Specification. The 7 day Surveillance Frequency ensures that an adequate initial water supply is available for boron injection. SR 3.1.9.3 and 3.1.9.4 These SRs demonstrate that each automatic boration system pump and valve is operable and actuates as required. In response to an tctual or simulated SIAS the charging pumps start, the VCT is isolated, and the charging pumps take suction from the OPERABLE BAMU tank (s)for manual, power and RWST. 1 Verification of the correct alignment operated, and automatic valves in the Boration System Flow paths )rovides assurance that proper boration flow paths are availa)1e. These SRs do not apply t'o valves that are locked, sealed, or otherwise secured in position, because these valves were previously verified to be in the correct position. WlEti4M j $3ach[de{lfy@i@ce?)nspectionsifis@ TiiiiiWidFlh "o']pra'bT thelinservJcek t htecMcomponestl[j....._._ja ajsjMeaMaM. gem serv j REFERENCES 1. 10 CFR 50, Appendix A, GDC 26. 2. 10 CFR 50, Appendix A, GDC 27. ( SAN ON0FRE--UNIT 2 B 3.1-58 Amendment No. 127 4/30/98

u ECCS-Operating B 3.5.2 .B 3.5' EMERGENCY CORE COOLING SYSTEMS (ECCS) B 3.5.2 ECCS-Operating BASES BACKGROUND The function of the ECCS is to )rovide core cooling and negative reactivity to ensure t1at the reactor core is protected after any of the following accidents: a. Lossofcoolantaccident(LOCA); b. Control Element Assembly (CEA) ejection accident; c. Loss of secondary coolant accident, including uncontrolled steam release; and. d. Steamgeneratortuberupture_(SGTR). The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power. There are two phases of ECCS operation: injection and recirculation. In the injection phase, all injection is initially added to the Reactor Coolant System (RCS)(via the cold legs. After the refueling water storage tank RWST) has been depleted, the ECCS recirculation phase is entered as the ECCS suction is automatically transferred to the containment sump. During the later portions of the recirculation phase, the injection flow is split approximately equally between the hot.and cold legs. Two redundant, 100% capacity trains are provided. In MODES 1, 2, and 3, with pressurizer pressure 2 400 psia, eachtrajn$.consistsofhighpressuresafety(LPSI)7and injection (HPSI)T ch;rgir.;gij low pressure safety injection

tt;y;;;;,;.

In MODES 1, 2, and 3, with pressurizer pressure 2 400 psia, both trains must be OPERABLE. This ensures that 100% of the core cooling requirements can be provided in the event of a single active failure. A' suction header supplies water from the RWST or the containment sump to the ECCS pumps. Separate piping supplies each train. The discharge headers from each HPSI pump divide into four supply lines. Both HPSI trains feed (continued) SAN ONOFRE--UNIT 2 B 3.5-11 Amendment No. 127 4/30/98 L

T ECCS-Operating B 3.5.2 BASES (continued) BACKGROUND into each of the four injection lines. The discharge header (continued) from LPSI pumps divides into two supply lines, each feeding the injection lino to two RCS cold legs. Orifices are set to balance the flow to the RCS. This flow balance directs sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legse P__JJA J. A.L E._ AL_ J_..__A_-.. ___.JJJ L.. AL- .L __2__ b5Lu5b IJ b u NEE B IVE b rIE 555VbubVig

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[ ECCS-Operating B 3.5.2 BASES (continued) l BACKGROUND I (continued) jp_; gj'fpg, e!'. j g',.vgj jp,t.y _g.g igir.p_. g a g.) tyg g., .. 3 n,y...g . 3 p. .,.. 7 ..r.....y, y....y m..,, g r,.n 'r" I'."': ' ', ' Y"T l L' "Et " ""Y " "Y! ' 5 ' ' "Y _ ' '2'f"' ' _ ^ M""

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Y' ? ' * "' ? % ;YY _Y.* * " Ji _... > 2. 'a". ' "1. ' " i ". V9 ' 1"" ?. _ ' " L * * ""' 3 ' "' ' " " ' " \\'"*/ "'"*'Y'""' , os ...m s...... Durino low temperature conditions in the RCS, limitations are placed on the maximum number of HPSI u ps that may be OPERABLE. Refer to the Bases for LCO 3.4 1 " Low Temperature Overpressure Protection (LTOP ystem," for the basis of these requirements. During a C^< ~" large break LOCA, RCS pressure will decrease to psia in < 20 seconds. The safety injection (SI)-systemsareactuateduponreceiptofanSIAS. The actuation of safeguard loads is accomplished in a programed time sequence. If offsite power is available the safeguard loads start imediately in the programed, I se If o fsite power is not available the En ineered Safuence.ety Feature ESF) buses shed normal operpling loa!s and are connected t the diesel generators (DGs? Safeguard loads are then actuated in the programed t'ime sequence. The time delay associated with diesel starting, sequenced j loading and pump starting determines the time required before p, umped flow is available to the core following a LOCA. The active ECCS gompopents, along with the passive safety SITS " Safety in; ection tanks (SITS? covered in LCO 3.5.1,d in LCO 3.5.4, In;ection Tanks (Stora e Tank ( WST),1) provide th J," and the RWST covere " Refueling Water water necessary to mee GDC 35 Ref. APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSES criteria established by 10 CFR 50.46 (Ref. 2) for ECCSs, will be m,et following a LOCA: a. Maximum fuel element cladding temperature is s 2200*F; b. Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation; c. Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount (continued) SAN ONOFRE--UNIT 2 B 3.5-13 Amendment No. 127 4/30/98

PI ) i j ECCS-Operating l B 3.5.2 BASES (continued) APPLICABLE generated if-all of the metal in the cladding i SAFETY ANALYSES cylinders surrounding the fuel, excluding the cladding (continued) surrounding the plenum volume, were to react; d. Core is maintained in a coolable geometry; and e. Adequate long term core cooling capability is maintained. The LC0 also limits the potential for a post trip return to power following a steam line break (SLB) and ensures that l containment temperature limits are met. Both HPSI and LPSI subsystems are assumed to be OPERABLE in the large break LOCA analysis at full power (Ref. 3. This analysis establishes a minimum required runout flow)for the HPSI and LPSI pumps, as well as the maximum required response time for their actuation. The HPSI pumps and ct.;r 'Yg F;.;;; are credited in the small break LOCA ana1Ns. Tiis analysis establishes the flow and discharge head requirements at the design point for the HPSI pump. The SGTR and SLB analyses also credit the HPSI pumps, but are not limiting in their design. The large break LOCA event with a-loss of offsite power and a single failure (disabling one ECCS train) establishes the OPERABILITY requirements for the ECCS. During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclesr reaction is terminated either by moderator voiding during large breaks or control element assembly (CEA)-insertion during small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core. On smaller breaks, RCS pressure will stabilize at a value dependent upon break size, heat load, and injection flow. The LCO ensures that an ECCS train will deliver sufficient water to match decay heat boiloff rates soon enough to minimize core uncovery for a large LOCA. It also ensures that the HPSI ;r.d ct.;rgir.i pumpt will de iv water during a small brea c LOCA, and......g s,ufficient ..s .... r -..r., will provide sufficient boron to maintain the core subcritical following an SLB. The SGs continue to serve as the heat sink providing core cooling during a small break LOCA. (continued) SAN ONOFRE--UNIT 2 B 3.5-14 Amendment No. 127 4/30/98 l

ECCS-Operating 1 B 3.5.2 BASES (continued) APPLICABLE ECCS-Operating satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES ' Statement. (continued) l LCO In MODES 1, 2, and 3, with pressurizer pressure a 400 psia, two inde>endent (and redundant) ECCS trains are required to ensure tiat sufficient ECCS flow is available, assuming there is a sin Additionally, gle failure affecting either train. individual components w l may be called upon to mitigate the consequences of other transients and accidents. InMODES1and2,andinMODE3withpressurizerpressurg$@ 2 400 psia, an ECCS train consists of a HPSI subsystem-a LPSI subsystem, and e charging ;.b;ystem. Each train includes the piping, instruments, and controls to ensure the availability of an OPERABLE flow path capable of taking suction from the RWST on a SIAS and automatically transferring suction to the containment sump upon a recirculation actuation signal (RAS). During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the RWST to the RCS, via the HPSI and LPSI pumps and their respective supply headers, to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply part of its flow to the RCS hot legs via the shutdown ggpljngp suctign ngz,ges J.[.cjjjg{gj{.g pggj}g(330. l ..rr.. I R C S v i e.......... ... 3...,..nas. The flow path for each train must maintain its designed l independence to ensure that no single failure can disable both ECCS trains. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure 2 400 psia, the ECCS OPERABILITY re uirements for the limitingDesignBasisAccident(DBA large break LOCA are based on full power operation. Alt ough reduced power would not require the same level of performance, the accident i (continued) SAN ONOFRE--UNIT 2 B 3.5-15 Amendment No. 127 4/30/98 l

= ECCS-Operating B 3.5.2 BASES (continued) APPLICABILITY analysis does not provide for reduced cooling requirements (continued) in the lower MODES. The HPS1 pump performance is based on the small break LOCA, which establishes the pum curve and has less dependence onper. ],[.gc..p performance ergjr.pgp ihe'rekUYremen E oI 5dDES I E E3 with il d prb E re ~ ~ 2 400 psia, are bounded by the MODE 1 analysis. The ECCS functional requirements of MODE 3, with RCS gressure < 400 psia, and MODE 4 are described in LCO 3.5.3, ECCS -Shutdown." In MODES 5 and 6, unit conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops MODE 5 Loops Filled," 'and LCO 3.4.8, "RCS Loops-MODE 5 Loops Not Filled." MODE 6 core cooling req)uirements are addressed by LC0 3.9.4, " Shutdown Cooling (SDC and Coolant Circulation-High Water Level " and LCO 3.9.5, " Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level." ACTIONS A.1 and B.1 An ECCS train is inoperable if it is not capable of delivering the design flow to the RCS. The individual components are inoperable if they are not capable of performing their design function, or if supporting systems are not available. The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of each of Condition A and Condition B is to maintain a combination of OPERABLE equipment such that 100% of the ECCS flow equivalent to 100% of a single OPERABLE train remains available. This allows increased flexibility in plant operations when components in opposite trains are inoperable. (continued) SAN ONOFRE--UNIT 2 B 3.5-16 Amendment No. 127 07 5 98 Re-issued 08 4 98

ECCS-Operating B 3.5.2 . BASES (continued) ACTIONS -A.1 and B.1-(continued) Each of Condition A and Condition B includes a combination of OPERABLE equipment such that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available.- Condition-A addresses the specific condition where the only affected ECCS subsystem is a single LPSI subtrain. The availability of.a least 100%.of the ECCS flow equivalent to a single OPERABLE-ECCS train is implicit in the definition of Condition A. If LCO 3.5.2 requirements are not met due only to the existence of Condition A, then the ino)erable LPSI subtrain components must be returned to OPERABLE status within 7 days of discovery of Condition A. A Configuration Risk Management Program (CRMP) defined in Administrative Controls section 5.5.2.14 is implemented in the event of Condition A. This 7-day Completion Time is based on the findings of the kterministic and probabilistic analysis that are discussed in Reference 6. Seven days.is a reasonable amount of time to perform many corrective and preventative maintenance items on the affected LPSI subtrain. Reference 6 concluded that the overall risk impact of this Com either risk-beneficial or risk-neutral. pletion Time was Condition B addresses other scenarios where the availability of at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS-train exists but the full requirements of LCO 3.5.2 are not met. If Condition B exists, then inoperable components must be restored such that Condition B does not exist within 72 hours of discovery. The 72 hour Completion Time is based on an NRC reliability study (Ref. 4) and is a reasonable amount of time to effect many repairs. 5 55E55 'I ^*I'i_fo95! 5 53IS" b i $'_$li!9 ), $'i'5!95 """'t ' 't"'1 "L':!'l > !'l', ""!'.!iT"" ' 'lit !"'le : ' !* t ""u _.._ _- 1""Vf ' !" ' ',X'"'i!! ','? _ ' 1 "' "11 L, '_'""t,' 2"!_ !'J".'.U ','" ' " " '{ ' ' '"' L"' "Y:" 2!!! :L":"' '""r;' !"L', "' "_""'" ','J.!: _!:" "' "r!' 7"LT _"~' '< '"!" '"". "I t 'i'r _'!r!_il"" !N"!!.!!!"' ' ' !!!!!""LTI _I"'t!!r' t. A," ':'l '"! !' "' ?l"? '""'3 "??'I'.". t, _ ? l'_'"r? ' ?"i! !"" _""l L!' _ l' "1"' 2"!t i"" ' !! ?"! _"! !""!" ' ' !. "s",'.r ' ' ' ' " " _ ' ! '"' _ ! r '.,_ "' '_'!.._ "..',' ' !. _" "m,, ' L'.".' " ...s. ...r s... (continued) SAN ON0FRE--UNIT 2 B 3.5-17 Amendment No. 127 07 15 98 Re-issued 08 14 98

i ECCS-Operating B 3.5.2 BASES (continued) 1 ACTIONS A.1 and B.1 (continued) An event accompanied by a loss of offsite power and the failure of an emergency DG can disable one ECCS train until power is restored. A reliability analysis (Ref. 4) has shown that the impact with one full ECCS train inoperable is I sufficiently small to justify continued operation for i 72 hours. Reference 5 describes situations in which one component, such as a shutdown cooling total. flow control valve, can disable both ECCS trains. With one or more components inoperable, such that 100% of the equivalent flow to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analyses. In such a situation, LCO 3.0.3 must be immediately entered. C.1 and C.2 l If the inoperable train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and pressurizer pressure reduced to < 400 psia within 12 hours. The allowed Completion Times l are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems. l SURVEILLANCE SR 3.5.2.1 and 3.5.2.2 REQUIREMENTS SR 3.5.2.1 verification of proper valve position ensures that the flow path from the ECCS' pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removing power or by key locking the control in the correct position ensures-that the valves cannot be inadvertently misaligned or change position as the result of an active failure. These valves are of the type described in Reference 5, which can disable the function of both ECCS (continued) SAN ONOFRE--UNIT 2 B 3.5-18 Amendment No. 127 07/15/98 u

r: I ECCS-Operating B 3.5.2 BASES (continued) SURVEILLANCE SR 3.5.2.1 and 3.5.2.2 (continued) REQUIREMENTS trains and invalidate the accident analysis. SR 3.5.2.2 verification of the proper positions of the Containment Emergency Sump isolation valves and ECCS pumps / containment spray pumps miniflow valves ensures that ECCS operability and containment integrity are maintained. Securing these valves in position with power available will provide additional assurance that these valves will operate on a l RAS. A 12 hour Frequency is considered reasonable in view of other administrative controls ensuring that a mispositioned valve is an unlikely possibility. -SR 3.5.2.3 Verifying the corret s ali nment for manual, power operated, and automatic valves in the-ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those i valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under procedural control and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience. (continued) SAN ONOFRE--UNIT 2 B 3.5-19 Amendment No. 127 07/15/98

p.,...._ i ECCS-Operating B 3.5.2 BASES (continued) SURVEILLANCE SR 3.5.2.4 REQUIREMENTS (continued) nonoperatingmode$heECCSpums '!ith thc ~ cxception of ; st in operation, 1 arenormallyinastandb"y,eT.; As such, f ow path piping has the of entrained gases. potential to develop voids and pockets Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e., air, nitrogen, or hydrogen) into the reactor vessel fo lowing an SIAS or during SDC. The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the adequacy of the procedural controls governing system operation. SR 3.5.2.5 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by Section XI of the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis. SRs are specified in the Inservice Testing Program, which encompasses Section XI of the ASME Code. Section XI of the ASME Code provides the activities and Frequencies necessary to satisfy the requirements. SR 3.5.2.6 bEGifDis-horst h ed at design fl;W is a ror el test of $'55i E5E t!'.'."'.'""N!l5![$.Sl fiEiN'_5 i!MC "I' $Y"' L _" _ r"' _"' if. ', ' !MN"!f ' "i ""!"_ _'!' . I' " _ !""!_ _ _. ' !r ' '.!'";"!

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ECCS-Operating B 3.5.2 ' BASES (continued) SURVEILLANCF SR 3.5.2.7. SR 3.5.2.8. and SR 3.5.2.9 REQUIREMENTS (continued)' These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SIAS and/or an actual or simulated RAS as appropriate to l each valve, that each ECCS pump starts on receipt of an actual or simulated SIAS, and that the LPSI pumps stop on receipt of an actual or simulated RAS. The 24 month Frequency is based on the need to perform these Surve111ances under the conditions that apply during a plant i outage and the potential for unplanned transients if-the Surve111ances were performed with the reactor at power. The 24 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the Engineered Safety Feature Actuation System (ESFAS) testing, and equipment performance is monitored as part of the Inservice Testing Program. SR 3.5.2.10 Periodic inspection of the containment sump ensures that it is unrestricted and stays in proper operating condition. The 24 month Frequency is based on the need to perform this i Surveillance under the conditions that apply during an outage, on the need to have access to the location. This Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience. i REFERENCES 1. 10 CFR 50, Appendix A, GDC 35. 2. 10 CFR 50.46. 3. UFSAR, Section 6.3 % $ Q (( M i @ M Q. 4. NRC Memorandum to V. Stello, Jr., from R. L. Baer, " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975. 5. IE Information Notice No. 87-01, January 6, 1987. 6. CE NPSD-995, "CE0G Joint Applications Report for Low Pressure Safety Injection System A0T Extension," May 1995. SAN ONOFRE--UNIT 2 B 3.5-20a 07/15/98

l i l l-l l l Attachment H PROPOSED BASES CHANGES \\ (for information only - strikeout for deletions and highlight for additions) San Onofre Unit 3 ] l

Boration Systems - Operating j B 3.1.9 BASES (continued) SURVEILLANCE SR 3.1.9.1 and 3.1.9.2 REQUIREMENTS SR 3.1.9.1 verifies that the boron concentration of the available boric acid solution in the BAMU tanks is sufficient for reactivity control. SR 3.1.9.2 verifies that a sufficient volume of borated water is available for RCS makeup. The minimum required volume and concentration of stored boric acid in the BAMU tank (s) is dependent upon the RWST boron concentration and is specified in a Licensee Controlled Specification. The 7 day Surveillance Frequency ensures that an adequate initial water supply is available for boron injection. SR 3.1.9.3 and 3.1.9.4 These SRs demonstrate that each automatic boration system pump and valve is operable and actuates as required. In response to an actual or simulated SIAS the charging ) umps start, the VCT is isolated, and the charging pumps ta ce suction from the OPERABLE BAMU tank (s) and RWST. Verification of the correct alignment for manual, power operated, and automatic valves in the Boration System Flow paths provides assurance that proper boration flow paths are available. These SRs do not apply to valves that are locked, sealed, or otherwise secured in position, because these valves were previously verified to be in the correct position. SEMI %Iii This755lVEE{ffEEfdhiS355{$iipIFri6i{ffyEiE{idifrMEidE!@l$ thesinserviceMestinggProgramn%SsEh3nserviceMaspecMons MetMtisomPOReuildegraditp[igh(11MMienifg}lutesi REFERENCES 1. 10 CFR 50, Appendix A, GDC 26. 2. 10 CFR 50, Appendix A, GDC 27. SAN ONOFRE--UNIT 3 8 3.1-58 Amendment No. 116 4/30/98

ECCS-Operating B 3.5.2 8 3.5 EMERGENCYCORECOOLINGSYSTEMS(ECCS) B 3.5.2 ECCS -Operating BASES i BACKGROUND. The function.of the ECCS is to )rovide core cooling and negative reactivity to ensure tiat the reactor core is ) protected after any of the following accidents: a. Lossofcoolantaccident(LOCA); b. Control Element Assembly (CEA) ejection accident; c. Loss of secondary coolant accident, including uncontrolled steam release; and d. Steam generator tube rupture (SGTR). The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power. There are two phases of ECCS operation: injection and recirculation. In the injection phase, all injection is cold legs.- After the refueling water storage tank)( initially added to the Reactor Coolant System (RCS via the has been depleted, the ECCS recirculation phase is entere as the ECCS suction is automatically transferred to the-containment sump. During the later portions of the recirculation phase the injection flow is split approximately equally between the hot and cold legs. Two redundant, 100% ca)acity trains are provided. In MODES 1, 2, and 3, wit 1 pressurizer pressure 2 400 psia, 3;rgn;;g'Nl consists of high pressure safety (LP eachtrafn. injection (HPSI)r low pressure safety injection

d;ys t;;;.;. iIn MODES 1, 2, and 3, with pressurizer pressure 2 400 psia, both trains must be OPERABLE. This ensures that'100% of the core cooling requirements can be i

provided in the event of a single active failure. A suction header supplies water from the RWST or the containment sump to the ECCS pumps. Separate piping supplies each train. The discharge headers from each HPSI pump divide into four supply lines. Both HPSI trains feed 4 SAN ONOFRE--UNIT.3 8 3.5-11 Amendment No. 116 4/30/98 i

ECCS-Operating B 3e5e2 BASES (continued) BACKGROUND inEo each of the four injection 11nese. The discharge header (continued) from LPSI pumps divides into two supply linesg each feeding i the injection line to two RCS cold legs. Orifices are set to balance the flow to the RCSe This flow balance directs sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legs. e_.JJA 2-A

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i Y, ",,' "23 ' " '.". r. ?,,',,' "; ?.. ' ". Y.. *. * * ""' v ' " ' ' " ' " ' " l \\"""I "'""*F'""' ..m m i During low temperature conditions in the RCS, limitations are placed on the maximum number of HPSI pu ps that may be OPERABLE. Refer to the Bases for LC0 3.5 1 " Low TemperatureOverpressureProtection(LTOPj stem," for the basis of these requirements. During a.~ilitf~(Ri large break LOCA, ds. pressure will t RCS decrease tb v zo0 psia in < 20 secon The safety 1 Theactuat'(SI)systemsareactuateduponreceiptofanSIAS. ~ injection on of safeguard loads is accomplished in a programmed time sequence. If offsite power is available the safeguard loads start immediately in the programmed, secuence. If o fsite power is not available the Engineered Safety Feature ESF) buses shed normal oper fing loads and are connected t the diesel generators (DGs. Safeguard loads are then actuated in the programmed t me sequence. The time delay associated with diesel starting, seauenced 4 before p,and pump starting determines the time requfredumped flow is a loading LOCA. The active ECCS qomponents, along with the passive safety in;ection tanks (SITS / covered in LCO 3.5.1,d in LCO 3.5.4 " Safety In ection Tanks LSITsJ," and the RWST covere "R;efueling Water Stora waternecessarytomeegeTank(WST),"providethecooling, GDC 35 Ref. 1) APPLICABLE The LCO helps to ensure that the following acceptance SAFETY ANALYSES criteria established by 10 CFR 50.46 (Ret. 2) for ECCSs, will be m,et following a LOCA: a. Maximum fuel element cladding temperature is s 2200"F; b. Maximum cladding oxidation is s 0.17 times the total cladding thickness before oxidation; c. Maximum hydrogen generation from a zirconium water reaction is s 0.01 times the hypothetical amount SAN ONOFRE--UNIT 3 B 3.5-13 Amendment No. 116 4/30/98

ECCS-Operating 8 3.5.2 ~ BASES (continued) APPLICABLE generated if all of the metal in the cladding SAFETY ANALYSES cylinders surrounding the fuel, excluding the cladding (continued) surrounding the plenum volume, were to react; d. Core is maintained in a coolable geometry; and e. Adequate long term core cooling capability is maintained. The LC0 also limits the potential for a post trip return to power following a steam line break (SLB) and ensures that containment temperature limits are met. Both HPSI and LPSI subsystems are assumed to be OPERABLE in the large break LOCA analysis at full power (Ref. 3). This analysis establishes a minimum required runout flow for the HPSI and LPSI pumps, as well as the maximum required response time for their actuation. The HPSI pumps and cher ' analysY;s.pu; ras are credited in the small break LOCA T11s analysis establishes the flow and discharge head requirements at the design point for the HPSI pum). The SGTR and SLB analyses also credit the HPSI pumps, )ut are not limiting in their design. The large break LOCA event with a loss of offsite power and a single failure (disabling one ECCS train) establishes the OPERABILITY requirements for the ECCS. During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control element assembly (CEA) insertion during small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core. On smaller breaks, RCS pressure will stabilize at a value dependent upon break size, heat load, and injection flow. The LC0 ensures that an ECCS train will deliver sufficient water to match decay heat boiloff rates soon enough to minimize core uncovery for a large LOCA. It also ensures that the HPSI and chargin:; >umpt will deliver sufficient water during a small breat.0CA, and that the ll PSI pu;rgs will provide sufficient boron to maintain the core subcritical following an SLB. The SGs continue to serve as the heat sink providing core cooling during a small break LOCA. SAN ON0FRE--UNIT 3 B 3.5-14 Amendment No. 116 4/30/98

ECCS-Operating B 3.5.2 BASES (continued) APPLICABLE ECCS-Operating satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES Statement. (continued) LCO In MODES 1, 2, and 3, with pressurizer pressure 2 400 psia, two inde)endent (and redundant) ECCS trains are required to ensure t1at sufficient ECCS flow is available, assuming there is a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents. In MODES 1 and 2, and in MODE 3 with pressurizer pressur 2 400 psia, an ECCS train consists of a HPSI subsystem g,d E a LPSI subsystem, and a darging 32 system. Each train includes the piping, instruments, and controls to ensure the availability of an OPERABLE flow path capable of taking suction from the RWST on a SIAS and automatically transferring suction to the containment sump upon a recirculation actuation signal (RAS). During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the RWST to the RCS, via the HPSI and LPSI pumps and their respective supply headers, to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply part of its flow to the RCS hot legs via the shutdown 992!!"9. p suctign ngzyes J{.2,g[g{gj..g pggj}g ; y, oons, o u s...... om..., n~. v. ens -m.. on. ,ory..s, ..s nsa ..u uns uv.mu. suu.yiny i.nu,. The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS pressure limiting Design Basis Accident (DBA)quirements for the 2 400 psia, the ECCS OPERABILITY re large break LOCA are based on full power operation. Although reduced power would not require the same level of performance, the accident SAN ON0FRE--UNIT 3 8 3.5-15 Amendment No. 116 4/30/98

ECCS - Operating-B 3.5.2 BASES (continued) APPLICABILITY-analysis does not provide for reduced cooling requirements (continued) in the lower MODES. The HPSI pump performance is based on the small break LOCA, which establishes the pum curve and has _less dependence onjower. ].gc..p performanc erg {r.py Ihelehir$ments'oI3idDES 2 anE3 wYth iE5 3ressbre ~ 2 400 psia, are bounded by the MODE 1 analysis. The ECCS functional requirements of MODE 3, with RCS gressure<400 psia,andMODE4aredescribedinLC03.5.3, ECCS - Shutdown. " In MODES 5 and 6, unit conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled." MODE 6 core cooling req)uirements are addressed by LC0 3.9.4, " Shutdown Cooling (SDC and Coolant Circulation-High Water Level " and LCO 3.9.5, " Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level." ACTIONS A.1 and B.1 An ECCS train is inoperable if it is not capable of delivering the design flow to the RCS. The individual components are inoperable if they are not capable of performing their design function, or if supporting systems are not available. The LC0 requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of each of Condition A and Condition B is to maintain a combination of OPERABLE equipment such that 100% of the ECCS flow equivalent to 100% of a single OPERABLE train remains available. This allows increased flexibility in plant operations when components in opposite trains are inoperable. (continued) SAN ONOFRE--UNIT 3-B 3.5-16 Amendment No. 116 07 15 98 Re-issued 08 14 98

f ECCS-Operating B 3.5.2 BASES (continued) ACTIONS A.1 and B.1 (continued) Each of Condition A and Condition B includes a combination of OPERABLE equipment such that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available. Condition A addresses the specific condition where the only affected ECCS subsystem is a single LPSI subtrain. The availability of a least 100% of the ECCS flow equivalent to t a single OPERABLE ECCS train is implicit in the definition of Condition A. If LCO 3.5.2 requirements are not met due only to the existence of Condition A, then the inoperable LPSI subtrain l components must be returned to OPERABLE status within 7 days of discovery of Condition A. A Configuration Risk Management Program (CRMP) defined in Administrative Controls section 5.5.2.14 is implemented in the event of Condition A. This 7-day Completion Time is based on the findings of the deterministic and probabilistic analysis that are discussed in Reference 6. Seven days is a reasonable amount of time to perform many corrective and preventative maintenance items on the affected LPSI subtrain. Reference 6 concluded that the overall risk impact of this Com either risk-beneficial or risk-neutral. pletion Time was Condition B addresses other scenarios where the availability of at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train exists but the full requirements of LC0 3.5.2 are not met. If Condition B exists, then inoperable components must be restored such that Condition B does not exist within 72 hours of discovery. The 72 hour Completion Time is based on an NRC reliability study (Ref. 4) and is a reasonable amount of time to effect many repairs. 95E9E3f_SI $I'i_E",5'[!iM59$iSE' ~$_^9E5'_$[i!9'i,E5'5'5!95 """"3 " 't'4'1 " L'! !" I"' ' ""!". 37'4"'"_3 '_'f'!a 1""< ! !"7 ' 1 ""... __ ' "!Pr i"" ' ),X"1!!"? _ ' ; "! _ "l ' L, '_ '"",',', 2 "t _ ! L'"l.X A "U "!""{ ' ""' e 1"' _"r '" !!!! ! L "!"', A'.'?t;' ""I';-r" ' 1" _ "":"..','.'.! ' _ !! " ; "' Y E!', ""L t '""' ;t :"""_ '!"~'. 31 t'"t_'!!!_i'r" !N"!?."!!"! "1"!!""L t! _1"'L!! r'." Ae" ' !" _ ;! !'_'"' ? L "? """"3 " t r"."._ t, _ l'_'Ytt ' ""1! """ _ "" " "Li ' _ l' "1"' 2"tt '_ "" ' !! t ','! _ "! !"""" ' ' !. "'"M l. ". 3""._ '."y"! _ i t. '. _ ' "!.. ".','. ' 7. ! ' ", o ' L' ".'. !..! ". " ' ' "' ' " "" m.. l i l (continued) SAN ONOFRE--UNIT 3 8 3.5-17 Amendment No. 116 07 5 98 l Re-issued 08 4 98

r ECCS-Operating B 3.5.2 BASES (continued) ACTIONS A.1 and B.1 (continued) An event accompanied by a loss of offsite power and the failure of an emergency DG can disable one ECCS train until power is restored. A reliability analysis (Ref. 4) has shown that the impact with one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours. Reference 5 describes situations in which one component, such as a shutdown cooling total flow control valve, can disable both ECCS trains. With one or more components inoperable, such that 100% of the equivalent flow to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analyses. In such a situation, LC0 3.0.3 must be imediately entered. C.1 and C.2 If the ino)erable train cannot be restored to OPERABLE status wit 11n the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and pressurizer pressure reduced to < 400 psia within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems. l l SURVEILLANCE SR 3.5.2.1 and 3.5.2.2 i REQUIREMENTS SR 3.5.2.1 verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removing power or by key locking the control in the correct position ensures that the valves cannot be inadvertently misaligned or change position as the result of an active failure. These valves are of the type described in Reference 5, which can disable the function of both ECCS l (continued) l SAN ON0FRE--UNIT 3 B 3.5-18 Amendment No. 116 07/15/98

r1~ ECCS-Operating B 3.5.2 BASES (continued) l SURVEILLANCE SR 3.5.2.1 and 3.5.2.2 (continued) REQUIREMENTS trains and invalidate the accident analysis. SR 3.5.2.2 verification of the proper positions of the Containm0M Emergency Sump isolation valves and ECCS pumps spray pumps miniflow valves ensures that ECCS o/ cont 6 Wnt perability and containment integrity are maintained. Securing these valves in position with power available will provide additional assurance that these valves will operate on a RAS. A 12 hour Frequency is considered reasonable in view of other administrative controls ensuring that a mispositioned valve is an unlikely possibility. SR 3.5.2.3 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are j locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those position.pable of being mispositioned are in the correct valves ca The 31 day Frequency is appropriate because the valves are operated under procedural control and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience. l I L (continued) l SAN 0N0FRE--UNIT 3 B 3.5-19 Amendment No. 116 07/15/98

l ECCS-Operating B 3.5.2 BASES (continued) SURVEILLANCE SR 3.5.2.4 REQUIREMENTS "iththeexcc-tier,ofs-stcrsir,ecretier.,$jheECCSpumfs (continued) are nomally in a standby, nonoperating mode. As such, f ow i of entrained gases. potential to develop voids and pockets path piping has the ) Maintaining the piping from the ECCS pumps to the RCS full of water ensures that the system will perfom properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, cavitation, and pumping of noncondensible gas (e.g. pump, air, nitrogen, or hydrogen) into the reactor vessel following an SIAS or during SDC. The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the adequacy of the procedural controls governing system operation. SR 3.5.2.5 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or otler hydraulic component problems is required by Section XI of the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This veriffes both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the unit safety analysis. SRs are specified in the Inservice Testing Program, which encompasses Section XI of the ASME Code. Section XI of the ASME Code provides the activities-and Frequencies necessary to satisfy the requirements. SR 3.5.2.6 hk b i!$5I55 M b_5$_k!!E9_ bed..I3_5.99F'5 $53!<5I.co, _":' ' ' "Y t'""t t1'.":.'"r"N!_ ' N!'_'". "I !!!!!"'1 "12 :"' !""N L _" _5""' *:.'. if ', ' !!"!!3 ' r i " "I',' _ '! ' '; " ! " " _ '""! _ _ _. ' !N" ' ' r'";"I

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2 ':!' ' ?"" ' ' "" """ _ I"71 t ' !.'.' ' _ l " ' ' "17 " v',"' d'Y'I' U Z, ,_a_ " M...t.' ",' r'."'." *. !", "..L Y.s "Y _ t,".T. t.,' !..!! *.' "s " >. 'v'n' I.'.?.".!_".,". '""' 1 n .s es (continued) SAN ON0FRE--UNIT 3 B 3.5-20 Amendment No. 116 07/15/98

T ECCS-Operating B 3.5.2 BASES (continued) SURVEILLANCE SR 3.5.2.7. SR 3.5.2.8. and SR 3.5.2.9 REQUIREMENTS (continued) These SRs demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SIAS and/or an actual or simulated RAS as appropriate to each valve, that each ECCS pump starts on receipt of an actual or simulated SIAS, and that the LPSI pumps stop.on receipt of an actual or simulated RAS. The 24 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power. The 24 month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the Engineered Safety Feature Actuation System (ESFAS) testing, and equipment performance is monitored as part of the Inservice Testing Program. SR 3.5.2.10 Periodic inspection of the containment sump ensures that it is unrestricted and stays in proper operating condition. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during an outage, on the need to have access to the location. This Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience. REFERENCES 1. 10 CFR 50, Appendix A, GDC 35. 2. 10 CFR 50.46. 3. UFSAR, Section 6.3 R $ $ { E {b6 M M[{.3 4. NRC Memorandum to V. Stello, Jr., from R. L. Baer, " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975. 5. IE Information Notice No. 87-01, January 6, 1987. 4 6. CE NPSD-995, "CEOG Joint Applications Report for Low Pressure Safety Injection System A0T Extension," May 1995. SAN ON0FRE--UNIT 3 B 3.5-20a 07/15/98

ECCS-Operating B 3.5.2 Attachment I PROPOSED UPDATED FINAL SAFETY ANALYSIS REPORT - CHAPTER 6 (strikeout for deletions and highlight for additions) San Onofre Units 2 and 3 1 i f

San Onofro 2&3 FSAR Updatsd EMERGENCY CORE COOLING SYSTEM 6.3.3.3 e-.11 ar==k an=1vef= 6.3.3.3.1-Safety Injection System Assumptions The major components of the safety injection system (SIS) include three high pressure pumps, two low pressure pumps and four safety injection tanks. Two HPSI and the LPSI pumps are automatically actuated by a SIAS, generated by either a-low pressuriser pressure signal or a high containment pressure signal. Flow from the v...Lle safety injection tanks is initiated when the -cold leg pressure drops below the SIT pressure. In performing the'small break estentet-sens idMNGubMym$g conservative -assumptions are made concerning the availability of safety injection flow. It is assumed that offsite power is lost upon reactor trip and, therefore, the . safety injection pumps must await diesel startup and load sequencing before they can start.' The delay time based on SIAS generated from a low pressuriser pressure signal is R$$jN 92-* seconds for the C .wi-, 0 HPSI systems and 4t-9 seconds-for-the LPSI systemis.. For breaks in a reactor coolant pung discharge leg, it is also assumed that all safety injection flow delivered to the broken leg spills out the break. An analysis of the possible single failures that can occur within the SIS has shown that the worst single failure for a small IOCA is the failure of one of the emergency diesels to start (Reference 7). This failure causes the loss of one high pressure pump, one low pressure pump and two LPSI pump header ' isolation valves. These losses result in a minimum of safety injection water being available to cool the core. The available safety injection water is supplied by one HPSI pump which injects into each cold leg, one LPSI pung which injects into two cold legs and four safety injection tanks each of which injects into one cold leg.

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San On: fro 2&3 FSAR-Updated EMERGENCY CORE COOLING SYSTEM Table 6.3-5 presents the high and-tow pressure safety injection pump flow rates $ge@Miji[llig@pu$47)ijdji$fM3jiigistihis4% L.. 0 - L,0..11,. 4.Le, Z.... . 1-. 11 =....Li. 1,.0j..a I.. .. lu ;.L. ' l L...k 1 l.iii.,ue. 6.3.3.3.2 Core and System Parameters he significant core and system parameters used in the small break hDC33 audi$$$iji catentations are presented in paragraph 15.6.3.3. X PLHGR of $$25 25-9 kW/ft was used in the calculations. This is conservative compared to the PLHGR limit of 13.0 kW/ft assumed for the large break LOCA calculations. 2e initial steady state fuel rod conditions were obtained from the FATES 3M computer program"' "' *d. The small break M ji$$W3E4 e1.le;.i...... performed at a jiiiiijlfjdiiWifii$i~E$37#isigl$ hot rod average burnup er ;7?, ;;;;/;.... g 6.3.3.3.3 Break Spectrum Calculations were performed for the following $ijiBR six break sizes, #6W t 9-1, 0.07 ;, 0.05, ($$$ 9 099 and 0.01 ft*. The breaks were conservatively postulated to occur in the reactor coolant pung discharge leg. .6.3.3.3.4 Results and Conclusions he results of this analysis are presented in paragraph 15.6.3.3. Table 15.6-19 lists the important results which demonstrate cospliance with the acceptance criteria of Reference 1. Table 15.6-12 presents the time sequence of important events for short term cooling for the Biliiii six small breaks analysed. 388biR3iiilit3ipiiiidsulttWdiltisiteE6WEiGMF2fdthBiENMKeiiGleiiDW2iiiiksuilinut paceTee#8stidausselsen@tsiiitmoosseamoseptuesmememuessiesme#Biirtosseslism gggymaWjggggg n ... 1;...L c_

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Sr.n Onofra 2fs3 FSAR c Updated EMERGENCY CORE COOLING SYSTEM TablW3N M 5 MIGHIP$tES$URC234FETCNIPl3EP m'enLtymanners<nGMince 3Asmanimiiff;mImmergener1Essis13eimsastuf! psi 14d1 hCSTPsniiiiiiNiiiThisiiiiiiT F10iiE%stis"?(5inf/siifE)l" ists -s372 1260 228344 1100 32*i2 1400 39511 M6 18476 566 5652% $66 86e75 540 kW171 l 406 Ve6YW 200 $$670 156 866?D 160 566YU Bd h66?d IN N667# 6 te6Te Mal M&Hiiifdifictiaiss*Buiilbreak!%MMiifsiiGiiiiid3ttiiiiG25% isCWeitliMaignaiIoiict#63brea@siisdithsc:ha reesisiefiinstschtiiilmunI%uitaliipisdN6%ttinifishree $stactEMNMDiiiiiff simultaneous hot and cold leg injection maintains core cooling and boric acid flushing. The plant operator initiates appropriate procedures to be used based on the ability to achieve shutdown cooling entry criteria. Figure 6.3-1 shows the basic sequence of events and the time schedule for operator actions for the San Onofre LTC plan. The operator's first action is to initiate cooldown no later than 1 hour post-LOCA by releasing steam from the steam generators. The steam is released through the turbine bypass system if ac power is available, or through the atmospheric dump system if power is unavailable. 1/98 6.3-51 (cont.) Revision 13

San Onnfra 2&3 FSAR Updated EMERGENCY CORE COOLING SYSTEM nErrnENCEs w u=+-u----+.= =. w m.=y.i- _ --*

==w-7 r-*,-ww+ y - -*=== -----.=== u w-,. rw_,= _ yww-_ _ _. _. ..m _m. m, 2. Pilgrim II PSAR, Section'6E, Amendment 21, October 30, 1975, Docket No. 50-471. 3. " Calculative Methods for the CE Large Break LOCA Evaluation Model," Combustion Engineering, CENPD-132 August 1974 (Proprietary). " Updated Calculative Methods for the CE Large Break LOCA Evaluation Model," Combustion Engineering, CENPD-132, Supplement 1, December 1974 (Proprietary). 4. "CE Fuel Evaluation Model," Combustion Engineering, fMEDE@MIA.M EENFB-t99, July 1974 (Proprietary). 5. System 80 CESSAR PSAR, Section 6.3.3, Docket No. STN-50-470. 6. Letter, D. C. Switzer (NNECO) to 0. D. Parr (NRC), Millstone Unit 2 ECCS Reevaluation, July 10, 1975, Docket No. 50-336. l 7. " Calculation Methods for the CE Small Break LOCA Evaluation Model," Combustion Engineering Proprietary Report, CENPD-137, August 1974 (Proprietary). " Calculative Methods for the CE Small Break LOCA Evaluation Model,a t'nwon-117. snemi -- -* 1-D, January 1977 (Proprietary). 5 8. " Post-LOCA Long Term Cooling Evaluation Model," Combustion Engineering, CENPD-254-P-A, June 1977 (Proprietary). 9. COMPERC-II, A Program for Emergency Refill-Reflood of the Core," Combustion Engineering, CENPD-134, April 1974 (Proprietary); with Supplement 1, December 1974 (Proprietary). 10. Letter, R. L. Baer (NRC) to A. E. Scherer (CE), July 30, 1979. 11. CEN-161(B)-PM (Proprietary), " Improvements to Fuel Evaluation Model" ^ ~ ^ 3) '. 4.w,.i.L..y)- Combustion Engineering, duty j t9tt $%HN. 12. ~ CEN4511mFWFT TT TNERMi!DiiFasaniliisin7sniipiia%Emialnae;4an W$@?*ii'**YW) ^ - W. ^'....;.. 6..=.--2...,;) L.w ?. ~: ".+ = .....,......w.. 1999-13. Memorandum for File dated September 9, 1993;

Subject:

SIT Operability Assessment, NCR 93080012 Disposition Step 1, San Onofre Nuclear. Generating Station, Units 2 and 3. 14. Memorandum for File dated August 13, 1992;

Subject:

Active Failure Exemption for ECCS Check Valves.

15. -P-A to LD-81-095, "CE ECCS Evaluation Model Flow Blockage Analysis", December 1981.

16. Memorandum for file dated December 23, 1996;

Subject:

NEDOTRAK Action DBD-92-146, " Safety Injection FMEA Table 6.3.1, Chapter 15 Accident Analyses. HPSI Header Isolation Valves: 2/3HV9323, 2/3HV9324, 2/3HV9326, 2/3HV9327, 2/3HV9329. 2/3HV9330, 2/3HV9332, 2/3HV9333." 1/98 6.3-67 Revision 13

l i ATTACNMENT J PROPOSED UPDATED FINAL SAFETY ANALYSIS REPORT - CNAPTER 15 SAFETY ANALYSIS (strikeoutfordeletionsandhighlightforadditions) San Onofre Units 2 and 3 4 I

1 San Onofra 2&3 FSAR Updated DECREASE IN REACTOR COOLAFf INVENTORY . ~...... ---w. v, .- vo. -. so.- - - -.. - -, Broek . "._.wi Oywi F.J Sr,se y... ..e -fet't

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San Onofra 2&3 FSAR Updated DECREASE IN REACICR COOLANT INVENIVRY 15.6.3.3.3 Core and System Performance 15.6.3.3.3.1 Large Break IDCA A. Mathematical Model The large break LOCA analysis was performed using the NRC-approved June 1985 version of the ABB CE large break LOCA evaluation model (reference 4, Supplement 3). In the ABB CE evaluation model, the CEFLASH-4A computer program (reference 1) is used to determine RCS behavior during the blowdowc phase, and the COMPERC-II computer program (reference 6) is used to determine RCS behavior during the refill and reflood phases of the large break LOCA. The core flow and thermodynamic parameters from these two computer programs are input to the STRIKIN-II computer program (reference 7) which is used to calculate the hot rod cladding temperature transient. Also input into STRIKIN-II are steam cooling heat transfer coefficients which are calculated using the HCROSS (reference 25) and PARCH (references 25 and 8) computer-programs. The peak cladding temperature and maximum cladding oxidation are obtained frcm the STRIKIN-II calculation. The maximum core-wide cladding oxidation is obtained from the results of both the STRIKIN-II and COMZIRC (reference 6, supplement 1) computer programs. The initial steady state fuel rod conditions used in STRIKIN-II are determined using the FATES 3B computer program (reference 24). 1/98 15.6-30 (cont.) Revision 13

l i 1 l San Onofro 2&3 FSAR Updated DECREASE IN REACTOR COOLANT INVENTORY 15.6.3.3.3.2 Small Break IOCA A. Mathematical Model Stei!6sallmesamecuRaccs!iportossianoiEMos19iii1NIWiissibeireenneGittEttia steelemasC2Evminstominissezio4ItssE6ii!Dialasciif7AsaptessmC2!asedhbn isDeiiipaticsiamm1KZbsineRzecGn==h=m=Immen1M#eserwinocaameanc assuppmT!4tabiliIsas@diciasiiiDisnabensimi!iiiiisitantibsliii!MDanica asis 161iisbessedirkii&OnterdesactersB =a1=MaiDiefeconoiClifistinis

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wua wan a-ww-wa+www au.. wawuwww aa auw 6. au waw meer _ _ _a n nos., m1997"'1FWTW ........s .n a, ._y_. ___m . - - - -. u s. w w... s, .uw.= wwww.awww usw== - -- -- w a ws wumwww wu w an e _a u won w w usa wen wavu auwwwa. au aw-ywuww ww usw.w v, a,, a 6.wma ...,..on.gerr, s u. a.uw s wa sna as awus s.awu muu n ,, s i n..,.. anas wwww. .a i u-w a s a win i. a wu ws uns +4 me9r ,,w. __mt m_ e2_m 2_ t____t. g....-- w. w-ww..ai_. _ - _ _ -..2-u. __ - - w a = s.w w w us,4uy .a = a usw a .wawu 4 6.. s _ ._3 au wwuww.vmwaww y payu wamwwauy i. ywsawuswe. nwwawawum ww wa w .. aJMA"EL_ ww. ya w-wa4w -a ua wanes, a uns w waaw s wa vs w maaaaawww up erms y. as4 waw n.o,. no., s.nitt - =. - ___g __t __2_s mt_ ww=.yu ww s y vy s en. as .u ww 2_ e i._.__. _ m ww ww sana.aw waw ye wa asseauena y ang e wwma asy w+ =ua a w weaamvaws s.us asay i.a sw u.,s._., -_ m...,... _. -.. m.e.._ n.,.m..--,...... 1er _t___ __2 wwuyu ww a ya vy s mau an u nrw u ww .t n wu ww sea.euw wav spy a wvsani wwumvavs wuaauy iuw a w a. a www yumew. rwwa aww . s wwe.ywa ss ww ww smaaw w a suu was+ was w a wu ywawwuw ywar ums w waa wu.us www uimauy waaw efREMEI5-54""--end-PARGH" ww ywie. y.wy. -- Lw autw.f-iIuy wf Caw-w y.w3 w..ww. w .u www.a4 u w.woww... B. Input Parameters and Initial Conditions The important input parameters and initial conditions used if to u ly w the small break LOCA jiisiialysis are listed in table 15.6-18. l 1/98 15.6-35 Revision 13

I' l-l San On: fro 2&3 FSAR Updated DECREASE IN REACTOR COOLANT INVElf!ORY C. Results The important results of the small break calculations are summarized j in table 15.6-19. Table 15.6-20 lists the variables plotted versus time. The plots are prisented in figures 15.6-82 through 15.6-ifi j

199, A plot of peak candding temperature (PCT) versus brcak size is presented in figure 15.6-11( tSt.

] l For none of the break sizes analyzed does the maximum claddiing temperature, maximum claddbag oxidation, or coreiwide cladding oxidation exceed the limits established by the acceptance criteria for ECCS performance listed in Section 6.3.3.1. Coolable geometry and long term cooling are maintained as discussed in Section 6.3.3.1. BasedlisiiQ$sepsliefissiii4MEiMililiymm.swa9qggijgglij((giji[ilOnofsil%2&ilECCS l il6diidEWidt:Kaf4EiiEttiliEWEislidettuihiEMMENEAC464 Dis iipoo^tiiiiitfutfMe1RbsessecasyntsiEpnanETJimiliiWDsiiiEdismeinstihOuisti M213Darls4 The highest PCT of the $iiiiiiff six breaks analyzed is 1846199t*F. This is more than $59 te*F lower than the PCT of the 0.8 DEG/PD break which is the limiting LOCA (see paragraph 6.3.3.1). 1/98 15.6-35 (cont.) Revision 13

San Onofra 2&3 FSAR Updated DECREASE IN REACTOR COOLANT INVENTORY ..~... .___.,.w ,,mm._.._._m., .m.., ..m .,,.e Units v -u..1 u. .._, m 9499 Mwe um s. _., v 5.6 IcW/ft .,.1..w. u.u... -o. v. .u .u..., u tS-e u... 1999 ...,, m..,_,- 4 ,,.y vuo . w u. om.- u.u. Rat e Poet-ter.i..liu. T y... L...i 7.1 te9t-4 u.u.. 299t 4 1 -v...y. y.. 6u.. u.u... Hest-9 tate .m, um, m, e, wv. ..u m.. _.y.... w. w... ..w .u 1 of.. u. ,m, _-_..o . u.-, u. Iu.LI.1 3.i 7....... 2256 pria 1 S99.9 4 .u... _.y.... 6tt.2 4 v.. ww.. vom.... so.. 1969 psia ....w .~............... o w. -. uw 1960 pri1r ..vu .y .us. .vu w vu my.u m ft5 psia 1 u;. ...m peta .us.. ,o --y .a u.vu.... Shutoff-Head 1 199-6 pria .~......... y.us m wu ,-y ou. v. w ____ .w.u ... u .u sw v.. u.a.a.a. owws y ue.uy .uw .v. ..gu 1 1/98 15,6-36 Revision 13

f San Onofr3 2&3 FSAR Updated DECREASE IN REACTOR COOLANT INVENTORY $lalblM7AO354 Issentanggemmenenne?mmuEmssssasTemeersemi ygNTMI M

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%$$d giijsiis j meepoint taineressurisiiffweesseemirszasvestootag teso pesa fiss3Ds135%RMainsiiF4isisisff8PS13Pisiiii )(%g sinii f(grit:ta?&eseWds'f0ff5stilIBigeei$ 5mesEiMiiiilsielsiiiiReiiiiiFNis Die siiiiiil M acesIisiisiEIsiiill!EasiiF!sniedLHiiiiiimesiEsiniNdentirliiaIIiiEsdatSai[I4%JIi! M Sheenigueistitie#IEciiNhtspoiGiGiliiYsbdihveragiiiDnineniEji@dWiltiotGBjid MadiiinAmidisatsiniWIiiec14EthiintE9E41esifitteiirsiiensteuiEEliiittf411l1Trila154sE6eed emersiQ Sliiiiiialis%WililHilfsis7JiiiiiticabliiiItsiiEWfuii4ElsidisiliimpessiEssiisiDiiiiingliGRus3FFi#6 sco47 1/98 15.6-36 (cont.) Revision 23

San Onofra 2&3 FSAR Updated DECREASE IN REACTOR COOLANT INVENTORY u.. m,,, m., m.,.

m..om.,,. _.,,,,,_

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..&.n asww.y m...w. n....h. .. us F 1 m w... asww.y m .w. si w w........- .y .w. m,.-.. .m._. m. m... w u.. m.. y.ovv.m m.. .. y.m um.... .-. y..__n... y.. ...n... .y... m. m. m... TdbleI1!C(Mt PR$CCLAD.< TEMPERATURE 8IMIDEIDATIOillflPRRCEgwa n FORTNSMh1FBREAF9PECTRDW BFeak PeelEECIN9 We^fE^MC2'inidiiiid WEE?c6ME)ltidii TempeFatsfiTF}1 DiiiidatilhiiWM 0xidatidaTQn)" 0! M 268/PD 113# iTil EDT21 O!b57.fts,/PD 1884 1781 0748 l OYO4?ft'/PO 1727 2719 IE0736 DT0iYfh8/Pb 113 Di6s Ed?66 ~.,.__.n.,,1,~da. i., nfy, hn.ho,7. d7i.-~_ive..,.. .,la,gconservative,,. t onLo it et.tkro g sag ntas enl da Rod %veragelca: NE550I25EfiUd2bfEtlNIMiC65f#5fd83DIEidEEidis$ 1/98 15,6-37 Revision 13

San Onofr3 2&3 FSAR Updated DECREASE IN REACTOR COOIANT INVENTORY Table 15.6-20 VARIABLES PLOTTED AS A FUNCTION OF TIME FOR EACH SMALL BREAK Variable Normalised Total Core Power Inner Vessel Pressure-Break Flow Rate l Inner Vessel Inlet Flow Rate Inner Vessel Two-Phase Mixture M ;;.1,... Heat Transfer Coefficient at Hot Spot Coolant Teeperature at Hot Spot M

N isiEE"M r_;.2,~t n.:

_:.c.T ..i_. 1 ) l l l l l 1/98 15.6-37 (cont.) Revision 13

San On:fra 2&3 FSAR Updated DECREASE IN REACTOR COOLANT INVENTORY REFERENCES 1 "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis,* CENPD-133, April 1974 (Proprietary). "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident", N WD-111. sunn1===nt 1, August 1974 (Proprietary). "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis (Modification), " NwDD-113. Sunn1===nt 2, December 1974 (Proprietary). "CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss of Coolant Accident", NMDD-111. surel===nt 3, January 1977 (Proprietary). "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis" N wDD-113. sunn1===nt 4, April 1977 (Proprietary). "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis" N wpD.123. sunn1===nt 5, April 1977 (Proprietary).

2. Letter, K. P. Baskin, Manager of Nuclear Engineering, Safety and Licensing, Southern California Edison, to Office of Nuclear Reactor Regulation, NRC, April 28, 1982.

3. ChdeZififWedsidi1DegulsEHishifi$isT5MD%WL9GM6Hl$634E*Acceptanod CEinirIEf6idasesseneEcomeGF=leiFitiiimattuiGLiitiENstiuEundiaWMinE peactoscr u e.yL-~. c.it. 1.

v. 1.....~, tv.. cvvuc., 3,.L 20. Lishi-W. Lei Cvv1.0 L.1... iv.. Io.uiv..,- T L 1 2-I Lm. El 20. L 1, Am i, 4, 1;74. i 4. " Calculative Met: 7ds for the C-E Large Break LOCA Evaluation Model," NNDD-112, August 1974 (Proprietary). " Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model," N NPD-132. sunn1===nt 1, December 1974 (Proprietary). " Calculational Methods for the C-E Large Break LOCA Evaluation Model," NNpD-132. sunnia - t 2, July 1975 (Proprietary). " Calculational Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and E Designed NSSS," CENPD-132. Sunnl===nt 3, June 1985 (Proprietary). 5. "Reflood Heat Transfer: Application of FLECHT Reflood Heat Transfer Coefficients to C-E's 16 x 16 Fuel Bundles," CENPD-213, January 1976 (Proprietary).

6. "COMPERC-II, A Program for Emergency Refill-Reflood of the Core,"

CENPD-134, April 1974 (Proprietary). 1/98 15.6-62 Revision 13

i 1 San Ontfra 2&3 FSAR Updated DECREASE IN PEACIOR COOLANT INV.:NTORY 17. Letter LD-79-070, from A. E. Schcrer, C-E, to R. P. Denise, NRC, dated December 11, 1979.

18. -P of Letter LD-78-069, from A. E. Scherer, C-E, to Dr.

Denwood F. Ross, NRC, dated September 18, 1978. 19. Letter from A. E. Lundvall, Baltimore Gas and Electric Company, to D. G. Eisenhut, NRC, dated January 31, 1980. 20. J. J. DiNunno, at. al., " Calculation of Distance Factors for Power and Test Reactor Sites," TID-14844, Division of Licensing and Regulation, ABC, Washington, D.C., 1962. 21. ANSI N18.2, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," 1973. 22. " Calculative Methods for the C-E Small Break IOCA Evaluation Model", CENPD-137, August 1974, (Proprietary). I ""alculative Methods for the C-E Small Break LOCA Evaluation Model", c'wumD-127. niermi- =t 1, January 1977, (Proprietary). j Eli1MllR& Sit $ge0Nat$iiERF;j$aCR$iiMIM$hsRE9piR Vin!rfMiinimamFM nessarammmm,:amersemanimisirmiismaxsemasidsgesammassanese.Mw 23. Pilgrim Unit 2 PSAR, Section 6E, Amendment 21, October 30, 1975, Docket No. 50-471. 24. "C-E Fuel Evaluation Model Topical Report," CENPD-139, July 1974, -l (Proprietary). "Isuprovements to Fuel Evaluation Model," CENPD-161(B), August 1989, (Proprietary). "Inprovements to Fuel Evaluation Model," CENPD-161 (R). Surmi t 1, January 1992, (Proprietary). 25. NDMS Document Type SS, Document Number 62498. to LD-81-095, "C-E ECCS Evaluation Mode Flow Blockage . Analysis," December 1981 (Proprietary). $42 MlIMMMBESWINCWIMC$IlitligadBAaBCEf4008ptenOEMdid l -Mf4EnticmandanummeeroesssadatWEeuspiamentDI tcahnliaschuCanidliciieh?2ifitIthitTWgesiittEmiifemkumm1Matmatiuiiittsedun 3sMiseWI* thsiifessunE347D9972 if582: J 1/98 15.6-64 Revision 13

e \\ CHAPTER 15 SAFETY ANALYSIS FIGURES 15.6-82 THROUGH 15.6-130 ARE REPLACED BY THE FOLLOWING NEW FIGURES 15.6-82 THROUGH 15.6-114 ) o i \\ i i 4 9

1.50 1.25 g W 3 ys g 1.00 s m g O O a 4 H 0.75 0 o g N .~ ~ a ( 2 0.50 m O Z 0.25 0.00 O 200 400 600 800 1000 TIME,6EC SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Updateo Final Safety Analysis Report NORMALIZED TOTAL CORE POWER (0.06 FT8/PD SBLOCA) Figure 13.6-82 3

i i 2400 2000 i 1600 g a. g .(M (3- \\ 3 1200 u) q u) g cc Q. 800 I V 400 3 0 600 1200 1800 2400 3000 TIME, SEC 5AN ON0FRE NUCLEAR GENERATING STATION Units 2 L 3 Updated Final Safety Analysis Report INNER VESSEL PRESSURE (0.06 FT'/PD $8LOCA) Figure 15.6 83 i ) 1

1200 \\ i 1 800 -O = ~ ,,J O g 3 o a LL. ~ 400 ~ 1 ~ ~ 200 N ~.. w w E I L g i p f f A i 1 8 t I 1 1 4 i f 0 600 1200 1800 2400 3000 TIME, SEC ~ 5AN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report BREAK FLOW RATE (0.06 FT'/PD SBLOCA) Figure 15.6 84 3 I

50000 O 00U o %< 20000 g 3 o J lJ-10000 0 0 600 1200 1800 2400 3000 TIME, SEC SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER YESSEL INLET Floh RATE (0.06 FT'/PD $8LOCA) Figure 15.6 85

I 3 I I 3 I I I I g I 5 g 3 5 g I s q 3 I E I B F 5 5 I E B E 5 5 4 6 ~ ~ b 40 ~ j \\ 32 4 j g a ( = W 24 W< " TCP.O.F C.ent... k g I 16 . DOT.T.OM OF ComE. ...........s-m.. -8 0 0 600 1200 1800 2400 3000 TIME, SEC SAN ON0FRE NNCLEAR GENERATING STATION Units 2 & 3 i Updated Final Safety Analysis Report INNER VESSEL TWO. PHASE MIXTURE LEVEL j (0.06 FT'/PD $8LOCA) Figure 15.6 86

5 10 4 10 4 LL. 3 C. 10 m .g-E o 2 cm 10 g I y 1 10 E 1 P t 1 I f f I t t q 9 g g g g g 6 3 3 f f f 8 1 1 I P I I i 1 f f 0 600 1200 1800 2400 3000 TIME, SEC 5AN ONOFRE NNCLEAR SENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report HEAT TRANSFER COEFFICIENT AT HOT SPOT (0.06 FT'/PD $BLOCA) Floure 15.6-87 .-+m,

F.-. 1800. l ~ 1 ~ \\ l I l-1 / 1500 1 1200 W o l J \\ x H. 900 4 m W t -E L L., 600 .t / i' i ~ C- ~ 300 j 0 600 1200 1800 2400 3000 TIME, SEC 1AN ONOFRE HCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysts Report COOLANT TEMPERATURE AT HOT $P0T (0.06 FT /PD SBLOCA) 8 Figure 15.6-88

y$ 'Soo 1600 l W g h< 1300 g ~ W n. 2 ~ 1000 5 i 700 f ~ ' ~- g 0 600 1200 1800 2400 3000 TIME, S.EC SAN ONOFRE NNCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report CLADDING TEMPERATURE AT HOT $ POT (0.06 FT'/PD SBLOCA) Figure 15.6 89

1.50 = = 1.25. T A 1.00 i,y H .0.75 O F-Q W g E f'9 ~ ~ 0.25 ~ ~ ~ , i ,.i - 0.00 0 200 400 600 800 1000 TIME, SEC SAN ON0Ftt NNCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety' Analysis Report NORMALIZED TOTAL CORE POWER (0.05 FT'/PD SBLOCA) Floure 15.6 90

2400.......... 1 1 2000 1600 g g i t $g 1200 \\A g w ( 800 w 400 i i \\ O 0 600 1200 1800 2400 3000 TIME, SEP SAN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER VESSEL PRESSURE j (0.05 FT'/PD SBLOCA) i Figure 15.6 91 J

1 1 1200 ~ ~ 1000 1 l 1 1 ~ ~ ~ 800 Q, W 1 600 I I 4 I G: 3 O y 1.L. O 200 k N .I i 1 f I l l b t f 1 4 I f I i E .t f I f I 0 600 1200 1800 2400 3000 h TIME, SEC SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report BREAK FLOW RATE (0.05 FT'/PD $8LOCA) Figure 15.6-92 1 l l i 1 h

e 50000 ~ ~ ~ ~ ~ ~ 40000 ~ O w .,J 20000 ( g g o J LL 10000 0 E 9 t l l l t t I g g f I t f f I I I I I I I I O I I O 600 1200 1800 2400 3000 TIME, SEC 5AN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER VESSEL INLET FLOW RATE (0.05 FT'/PD SBLOCA) Ffgure 15.6 93 1 )

48 40 tt r w y W 24 m et . Tc, C.Nr ee.ns I f g .o / ) 16 ~ o .DOTTCedCNrC n........N 8 0 0 600 1200 1800 2400 3000 TIME, SEC 1AN 040FRE NBCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER VESSEL TWO. PHASE MIXTURE LEVEL (0.05 FT'/PD 58LOCA) Figure 16.6 94

s 10 i 4 10 l J u.- 3 j o, 10-j cv [ x E o 2 cm 10 O s. 1 10 I I I I I I I i i I f e I I I e i i t E I f 1 1 1 1 I f I f f p t 0 600 1200 1800 2400 3000 TIME, SEC. SAN ON0FRE NNCLEAR GENERATING STATION Units 2 & 3 l 1 Updated Final Safety Analysis Report HEAT TRANSFER COEFFICIENT AT H0T SPOT (0.05 FT'/PD $8LOCA) Figure 15.6-95 J

i \\ 1 1800 1500 { 1200 E b i W E 900 g m - t i lif w ( .L_ J = = E ~ = ~ 1 ~ 1 l l 0 0 600 1200 1800 2400 3000 TIME, SEC l 1 $AN ONOFRE NOCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report 1 COOLANT TEMPERATURE AT HOT SPOT (0.05 FT'/PO SBLOCA) Figure 15.6 96

l 2200 l i l i ~ ] 1800 i 1 l 1600 i t u. o W E / D H< 1300 1 E I W l a. 2 W 1000 700 f 1* (: i,,, g O 600 1200 1800 2400 3000 TIME, SEC i $AN ON0FRE NBCLEAR GENERATING STATION 1 i Units 2 & 3 Updated Final Safety Analysis Report CLADDING TEMPERATURE AT HOT SPOT l l (0.05 FT'/PD $8LOCA) Figure 15.6 97 1 l t i

1.50 ~ ~ ~ 1.25 g "A Q 1.00 W 8 o W H 0.75 Q o m N ~ .J 2 0.50 g O z 0.25 0.00 O 200 400 600 800 1000 TIME, SEC SAN ON0FRE NGCLEAR GENERATING STATION Units 2 & 3 Uccated Final Safety Analysis Report NORMALIZED TOTAL CORE POWER (0.04 FT'/PD $8LOCA) Figure 15.6-98 i l I ]

1 l l l 2400 l i l 1 l l 2000 1 k 1 I j l 1600 ) 1 g a.

(n g'

l 8 X l m I w g n. 800 s e 1 400 i 0 O 600 1200 1800 2400 3000 TIME, SEC SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysts Report INNER VESSEL PRESSURE (0.04 FT'/PD SBLOCA) Figure 15.6 99 I I l i t. s

1 i 4 1200 4 1000 ~ 800-O w a 600 ~ m 3 o a 1 400 'l 3 N_ ~ 0 0 .600 1200 1800 2400 3000 TIME, SEC 5AN ON0FRE NNCLEAR GENERATING $TATION Units 2 & 3 Updated Final Safety Analysis Report BREAK FLOW RATE (0.04 FT'/PD SBLOCA) f Figure 15.6 100 Li

$QQQQ O 30000 ~ c w J <m 3 O ~ a LL. 10000 ~ 0 ~ i,,... i qg O 600 1200 1800 2400 3000 ) TIME, SEC SAN ONPFRE N; CLEAR GENERATING STAT!0N I Units 2 & 7 Updated Final Safety Analysis Report INNER VESSEL INLET FLOW RATE (0.04 FT'/PD 58LOCA) Figure 15.6-101 i \\ l a 9 I l

48.'.' y, .). 40 j N r . q w W a W 24 M aC TC.P OF COM.... o / 16 . DCr. TOM OF com 8. 0 0 600 1200 1800 2400 3000 TIME, SEC 5AN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER YESSEL Two. PHASE MIXTURE LEVEL (0.04 FT'/PD $8LOCA) Figure 15.6 102 i 1 a

5 10 ..........................I.......... H 4 10 ~ ~ Lt. 3 c. 10 N ~ H ~ u. W 2 H 2 cn 10 O H I [ u# i 1 10 10 0 600 1200 1800 2400 3000 TIME, SEC $AN ON0FRE NUCLEAR GENERATING STAT!0N Units 2 & 3 Updated Final Safety Analysis Report HEAT TRANSFER COEFFICIENT AT HOT SPOT (0.C4 FT'/PD $8LOCA) j Figure 15.6-103 t

i-1800 / 1500 12m i ur m 3 H 900 E W k i wH ~ O L / ~ ~ ~ 300 ~ ~ 0 0 600 1200 1800 2400 3000 TIME, SEC SAN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report COOLANT TEMPERATURE AT HOT SPOT (0.04 FT'/PD SBLOCA) Figure 15.6 104 f a

2200 1900 1600 LA. O Lij 12" D H< 1300 cc w ~ G. 2 N 1000 ^ 700 f ~ 400 0 600 1200 1800 2400 3000 TIME, SEC 5AN ON0FRE HCLEAR GENERATING STAT!0N Units 2 & 3 Updated Final Safety Analysis Report CLADDING TEMPERATURE AT HOT SPOT (0.04 FT'/PD SBLOCA) Figure 15.6 105 i_

r-1.50 b4 1.25 g y 3: p-O 1.00 Q-W g O o a H 0.75 0 H a Lu N a< 2 0.50 g O z 0.25. 0.00 0 200 400 600 800 1000 TIME, SED SAN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report NORMALIZED TOTAL CORE POWER ('0.01 FT'/PD SBLOCA) Figure 15.6 106 I I i l __ l

) 2400 ~ ) 2000 1600 ) U) G. td ~ D% \\ 1 W --) 1200 ~ U3 U) Lu 1 Q. ~ 800 ~ ~ 400-i 0 0 1000 2000 3000 4000 5000 TIME, SEC SAR DuoFRE NNCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER VESSEL PRES $URE (0.01 FT'/PD SBLOCA) Figure 15.6 107

) 300 250 1 G 200 O m i m m i s 2 a 150 m 3: O au. 1 100 b L 50 b . m ._ I j 0 O 1000 2000 3000 4000 5000 TIME, SEC $4N ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report BREAK FLOW RATE (0.01 FT'/PD SBLOCA) Figure 15.6-108 1

75000 60000.' 45000 o Lu t,1) 3 i I $ 30000 4' cc g o ~ J 11. 15000 ~ ~ e 0. -15000 0 1000 20G0 3000 4000 5000 TIME, SEC $AR ONOFRE NOCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER VESSEL INLET FLOW RATE (0.01 FT'/PD SBLOCA) Figure 15.6 109

48 a e 40 ~ '32-tt n ~_ m x w w 24 m< - ro.,.e.,.e.nne... ........................... g c a O 3 16 .sorTou o.r ecas 8 1 0 1 0 1000 2000 3000 4000 5000 l TIME, SEC SAN ON0FRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report INNER VESSEL TWO. PHASE MIXTURE LEVEL (0.01 FT'/PD SBLOCA) Figure 15.6 110 i

5 10 4 10 w LL 3 C. 10 N m k m 2 a3 10 ) 1 d p I 1 10 0 10 0 1000 2000 3000 4000 5000 TIME, SEC i $AN ON0FRE NUCLEAR GENESATING STATION Units 2 & 3 Updated Final Safety Analysis Report HEAT TRANSFER COEFFICIENT AT HOT $ POT (0.01 FT'/PD SBLOCA) Figure 15.6-111 l t. -. }

p 1200 ~ u. o ui xD s00 T L w q 0-i 2 w I 400 j 200 0 0 1000 2000 3000 4000 5000 TIME,,SEC 5AN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report COOLANT TEMPERATURE AT H0T SPOT (0.01 FT'/PD $6LOCA) Figure 15.6 112 4 i

y 2200 1900 1600 ~ u. O m T --) H< 1300 I ~ E W 4 G.2 m H 1000 o 700 400 ~ 0 1000 2000 3000 4000 5000 TIME, SEC $AN ON0 Fat NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report CLADDING TEMPERATURE AT HOT SPOT (0.01 FT'/PD $8LOCA) Figure 15.6 113 1 +

2200 i 'S j ) i 1 1600 LA. O W T 3 ~ H 1300 T W a. 2 m H / 1000 i l ~ ~ 700 400 0.00 0.02 0.04 0.06 0.08 0.10 BREAK AREA, FT2 SAR ONOFNE NUCLEAR GENERATING STATION Units 2 & 3 Updated Final Safety Analysis Report PEAK CLAODING TEMPERATURE VER$US BREAK SIZE FOR THE SMALL BREAK LOCA ECCS PERFORMANCE ANALY$!$ F1gure 15.6 114 U.}}

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	\|	}}[[Accession number::{{#if:ML20195E517\|ML20195E517\|{{#if:\|}}}}]]
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Site:	\|	}}{{#show:ML20195E517\|?site}}{{#arraymap:05000361, 05000362\|,\|x\| \|\| }}{{#if:05000361, 05000362\|{{#if:{{#show:{{#explode:05000361, 05000362\|,\|0}}\|?company\|link=none}}\|{{#if:{{#show:{{#show:{{#explode:05000361, 05000362\|,\|0}}\|?company\|link=none}}\|?icon\|link=none}}\| [[{{#show:{{#show:{{#explode:05000361, 05000362\|,\|0}}\|?company\|link=none}}\|?icon\|link=none}}\|link={{#show:{{#explode:05000361, 05000362\|,\|0}}\|?company\|link=none}}]]}}}}}}

Issue date:	\|	}}[[Start date::Issue date::{{#if:06/08/1999\|06/08/1999}}]]

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From:	\|	}}{{#arraymap:\|,\|x\|[[from::{{#if:{{#show:x\|?name\|link=none}}\|{{#show:x\|?name\|link=none}}\|x}}]]}}{{#if:\|{{#if:{{#show:\|?name\|link=none}}\|{{#show:\|?name\|link=none}}\|}}}}{{#if:SOUTHERN CALIFORNIA EDISON CO.\| {{#arraymap:{{#replace:{{#replace:SOUTHERN CALIFORNIA EDISON CO.\|, LLC\| }}\|, Inc\| }}\|,\|x\|[[from::{{#if:{{#show:x\|?name\|link=none}}\|{{#show:x\|?name\|link=none}}\|x}}]] }} }}

To:	\|	}}{{#arraymap:\|,\|x\|[[to::{{#if:{{#show:x\|?name\|link=none}}\|{{#show:x\|?name\|link=none}}\|x}}]]}}{{#if:\|{{#arraymap:\|,\|x\|to::x}}}} {{#arraymap:{{#replace:{{#replace:\|, LLC\| }}\|, Inc\| }}\|,\|x\|[[to::{{#if:{{#show:x\|?name\|link=none}}\|{{#show:x\|?name\|link=none}}\|x}}]]}}

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