Text

Proposed Tech Specs to Support Refueling W/Vantage 5 Improved Fuel Design
	ML18022A714
Person / Time
Site:	Harris
Issue date:	04/17/1989
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML18022A713	List: ML18022A712; ML18022A714;
References
	NUDOCS 8904250293
	Download: ML18022A714 (390)
	v • d • e

ATTACHMENT 2 TECHNICAL SPECIFICATIONS CHANGE PAGES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT TRANSITION TO 17 x 17 VANTAGE 5 FUEL 85'04 ~0>>8 POP P AOOcg 0890 4I 7 000400 (277CRS/lah) toe

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l. 0 DEFINITIONS SECTION PAGE ACTIONo ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ o o ~ o ~ ~ ~ ~ ~ o 1.2 ACTUATION LOGIC TESi............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~

1.3 ANALOG CHANNEL OPERATIONAL TEST....... ~ "" ~ ~ ~ ~ .-~-~~~--~-~~~ ~

1e 4 AXIAL FLUX D IFFERKNCEo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1. 5 CHANNEL CALIBRATION...........................................

1.6 CHANNEL CHKCKo ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

l. 7 CONTAINMENT INTEGRITY........ -... - " -"-- ~ -- ~ ~ ~ - " ~ --- ~ ~ ~ ~ ~ -- ~ 12 1.8 I CAVA/ eo C ONTROLLED LPKAGCo ~ 1-2

~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ e ~ o ~ ~ ~ ~ ~ o ~ ~ ~

1. 9 CORE ALTERATION............. ~ ~ ~ -~- ~ ~ ~ -- ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ -~ ~ -~ ~ 1-2

l. 0 DIGITAL CHANNEL OPERATIONAL TEST......... ~ .~. - -.... ~ ~ ~ ~ ~ ~ ~ - ~ ~ 1-2 DOSE EQUIVALENT I-131............... 1-2

l. 12 E - AVERAGE DISINTEGRATION ENERGY... 1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME..................... 1-3 1.14 EXCLUSION AREA BQUNDARY........................... ~ - ~ ~ ~ ----- ~

1a3

1. 15 FRE)UENCY NOTATIONo ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ e ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ 1-3

1. 16 GASEOUS RAQMASTE TREATMENT SYSTEM................. ~ ~ - ~ ~ - ~ ~ ~ ~ ~ 1-3

l. 17 IDKNTIFIED LEAKAGEe ~ ~ ~ ~ ~ ~ ~ o o ~ o ~ ~ ~ o ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ 1-.3

1. 18 MASTER RELAY TESTo ~ ~ ~ oo ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ eooooooo ~ ~ ~ o ~ ~ ~ ~ oo ~ o ~ ~ ~ ~ 1-4 1.19 MEMBER(S) OF THE PUBLIC...................................... 1-4 1.20 OFFSITE DOSE CALCULATION MANUAL....-~ ~ --- ~ - ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - ~ 1-4 1.21 OPERABLE - OPERABILITY......,.......,........................ 1-4 lo22 OPERATIONAL HOOK - MODEe ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1-,4

l. 23 P HYHCS TESTS o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o 1-4 1.24 PRESSURE BOUNDARY LEAKAGE.........-..-.-..-...-..-.-..- 1-4 1.25 PROCESS CONTROL PRQGRAM........... ~ - ~ ~ ~ -~ -- ~ ~ ~ ~ - ~ ~ ~ - ~ ~ ~ ~ ~ ~ - ~ ~ 1-5

1. 26 PUnr E PURG INGo ~ o ~ o ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~

1-5

l. 27 QUADRANT P(MER TILT RATIO........... 1-5

l. 28 RATED THERMAL P(%ER.. 1-5

l. 29 REACTOR TRIP SYSTEM RESPONSE TIME........... -- ~ ~ ~ ~ ~ -~- ~ ~ --- ~ ~ 1-5

l. 30 REPORTABLE EVENT......... ~ ~ ~ ~ ~

1. 31 HUTDQO MARG N o I ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

1-5 l.lo. CORE OPERATlN6 LtMI78 REPORT...... ~ ~ ~.... ... ~

I-2 SHEARQN HARRIS - UNIT 1

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3.0/4 0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UZRlKENTS SEGTIOM 3/4 ' APPLICABILITYe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 0-1 3/4 1 REACTIVITY CONTROL SYSTEMS 3/4el ~ 1 BORATIOM CONTROL Shutdown Margin - MODES 1 and 2. " .. ~ ~ ~ ............ ..... ~ 3/4 1-1 Shutdown Margin - MODES 3i 4, and S ~ . ~ ~ ~ .~ .... ~ ~ . ~ .. ". " 3/4 1-3 FIGURE 3el 1 SHUTDOWN MARGIN VERSUS RCS BORON CONCENTRATION MODES 3y 4y AMD 5 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-3a Moderator Temperature Coef f 1 cia'n't ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-4 Mx.namum Temperature for Crxtxcalxtye ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ 3/4 1-6

~ ~ ~ ~

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3/4 ~ 1 e2 BORATIOM SYSTEMS Plov Pa'th Shutdol%e ~ ~ ~ ~ ~ ~ m ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ e ~ ~ ~ 3/4 1-7 Plov Paths - Operatxnge ~ e ~ e ~ e ~ ~ ee ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ m ~ ~ ~ ~ m 3/4 1-8

~aarg1ng Pump - Shutdovneeee ~ ,~,~~~~~~~~ .~~~~~~~~~~~~~~~ . 3/4 1-9

~aargxng Pumps - Operating...."....".e.... ~ . "......." 3/4 1-10 Borated Water Source - Shutdown....... ~ ............ ". " . 3/4 1-11 Borated Water Sources Operating........................ 3/4 1-12 3/4.1 ' MOVABLE CONTROL ASSEMBLIES Group Height ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-14 TABLE 3 ~ 1-1 hCCIDENT ANALYSES REQUIRING REEVALUATION IM THE EVENT OP AM IMOPERABLE ROD e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ + ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-16 Position Indication Systems - Operating.... .. ~ ~ " "" "

~ ~ 3/4 1-17 Position Indication System - Shutdown..."" ~ . ~ " " ""

~ ~ 3/4 1-18 Rod Drop Time ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-19 Shutdom Rod Insertion Lxmxte ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-20 Control Rod Insert@on Lama, ts..... ~ ~ ~ . ~"

~ ~

" " ". " " "". 3/4 1-21 FIGURE 3.1-2

~ ~ e t'DCLBTEb (e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 1-22 SHEARON HARRIS - UNIT 1 LV Amendment No. P

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LIHITIMG CONDITIONS POR OPERATIOH AND SURVEILLANCE RE UIREMEliTS SEcTzom PACE 3/4.2 POMER DISTRIBUTION LIMITS 3/4 ~ 2~1 AXIAL PLUX DIFPERENCEo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2 1 FICURE 3 '-1 CL Kl g'PQP (Pr 0 Po ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-4 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fq(Z)......o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ 3/4 2-5 FIGURE 3.2-2 K(Z) - LOCAL AXIAL PENALTY FUHCTION FOR Fq(Z) ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-8 3/4.2.3 RCS FIAN RATE AND NUCLEAR ENTfihLPY RISE HOT CHANNEL PACTORo ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-9 3/4 ' 4 QUADRANT PNER TILT RATIOo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-l.l 3/4o2 ~ 5 DHB PARAMETERS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-14 3/4. 3 INSTRUMENTATION 3/4 ~ 3 ~ 1 REACTOR TRIP SYSTEM INSTRUMENTATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-1 TABLE 3 ' 1 REACTOR TRIP SYSTEM 1HSTRUHENTATION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEH INSTRUMENTATIOM RESPONSE TIHES ~ ~ ~ ~ 3/4 3-9 TABLE 4o3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQlgREMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-11 3/4o3o2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATIONo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUHENTATIONo ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-18 TABLE 3.3 4 ENGINEERED SAPETY PEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-28 TABLE 3 '-5 ENGINEERED SAFETY PEATURES RESPONSE.TIHES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3"37, TABLE 4o3 2 ENGINEERED SAFETY FEATURES ACTUATIOH SYSTEH INSTRUMENTATION SURVEILLANCE REQUIREMEHTS ~ oo ~o ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-41 3/4.3. 3 HONITORIHC IHSTRUMENThTION Radiation Monitoring for Plant Operations." " "" " "

~ ~ ~ 3/4 3-50 SHEAROH BARRIS - UNIT 1 Amendment No. g

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INOEN ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.3 CORPORATE NUCLEAR SAFETY SECTION F unctfon.................................................. 6-11 O~anf

~ zest f On Qanf Zaotf on ~ ~ ~ ~ ~ ~ ~ ~

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o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~

R eVf 8 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~

Records o o ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~

6.5.4 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Audf ts.................................................... 6-14 R 8 co rds ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ 6-15 AuthOrf tyo ~ ~ ~ ~ ~ ~ ~ o o o ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ o o ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ 6-15 6.5.5 OUTSIDE AGENCY INSPECTION AND AUDIT PROGRAM............... 6-15

6. 5 REPORTABLE EVENT ACTION... 6-15 6.7 SAFETY LIMIT VIOLATION.... 6-16 6.8 PROCEDURES AND PROGRAMS................................. "-- 6-16

6. 9 REPORTING RE UIRBIENTS 6.9.1 ROUTINE REPORTSo ~ o o ~ ~ o o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-20 S t2Lrtup Reporto o ~ ~ ~ o o ~ ~ o ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o 6-20 Annual Reports............................................ 6-20 Annual Radfologfcal Envfronaental Operatfng Report........ 6-21 Seaiannual Radioactive Ettluant Release Report............ 6-22

~RC OPEIM rhV$

Lib!(TS REPOR7 Monthly Operatf ng Reports.................... -. - ~ - " -""

- 6-23

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6-24 6.9.2 SPECIAL REPORTSo ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~

6-24

6. 10 RECORD RETENTION.............'.............................. 6-24 6.11 RADIATION PROTECTION PROGRAM...................'. "--. - ~ ~ ~- 6-26,

6. 12 HIGH RADIATION AREA........................................ 6-26 6.13 PROCESS CONTROL PROGRAM PCP .............................. 6-27 SHEARON HARRIS - UNIT 1 xfx

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DEFINITIONS CONTAINMENT INTEGRITY L.7 CONTAINHENT INTEGRITY shall exist when:

a. All penetrations requi.red to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE 'ALTERATION 1-9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe'onservative position.

~SEA'<><~+

'DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digi-tal computer hardware using data base manipulation to verify OPERABILITY of alarm and/or trip functions.

DOSE E UIVALENT I-131 1.LX DOSE EQUIVALENT I-Ul shall be that concentration of I-131 (microCurie/gram) which. alone ~ould produce the same thyroid dose as the quantity and isotopic mixture of I-331, I-U2, I-i33, I-U4, and I-?35 actually present. The thyroid dose. conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Tes Reactor Sites."

SHEARON HARRIS - UNIT 1 1-2

INSERT TO PAGE 1-2:

CORE OPERATING LIMITS REPORT 1.9.a The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle"specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Plant operation within these core operating limits is addressed within the individual specifications.

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0.5 O.d 0.8 0.9 1.0 1.1 1Z OF RATED TH POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS THREE LOOPS IN OPERATION SHEARON Y~IS - UNIT 2-2 Amendment No.

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8. .5 .i .5 .6 .7 .B -ci 1 ~ 1.1. 1 2 POMMER (f'rection ot noaineli XP SGR7 FOR: FZggRE 2 2 REACTOR CORE SAFETY LIMITS THREE LOOPS IM OPERATION 2-2

TABLE 2.2-1 REACTOR TRIP SYSTEH INSTRUHENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLRNANCE TA 2 LSS TRIP SETPRINT ALLOWABLE VALUE

1. Hanual Reactor Trip N.A. N.A. N.A. N.A. N.A.

2. Power Range, Hautron Flux

a. High Setpoint 7.5 4.56 0 <109K of RTP"* <ill.liof RTP>>>>

b. Low Setpoint 8.3 4.56 0 <251'f RTP>>>> <27.lX of RTP>>>>

3. Power Range, Hautron Flux, 1.6 0.5 0 <5X of RTP">> with <6.3X of RTP>>* with High Positive Rate a time constant a time. constant

>2 seconds >2 seconds

h. Power Range, Neutron Flux, 1.6 0.5 <5X of RTP">> with <6.3X of RTP*" with High Negative Rate a time constant a time constant

>2 seconds >2 seconds

5. Intermediate Range, 17. 0 8.41 0 <25K of RTP>>>> <30.9X of RTP*"

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 <10s cps <1.4 x 10s cps 6 OZ

7. Overtemperature hT Note 5 Sea Note 1 See Note 2 t.so

8. Overpower AT 1.9 See Note 3 See Note 4

9. Pressurizer Pressure-Low 5.0 2. 21 1.5 >1960 psig >1946 psig

10. Pressurizer Pressure-High 7.5 5. Ol 0.5 <2385 psig <2399 psig ll. Pressurizer Mater level-High 8.0 2. 18 1.5 <92K of instru- <93.8X of instru-ient span ment span

""RTP = RATED THERNAL POWER

I 61

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TABLE 2.2-1 Continued}

REACTOR TRIP SYSTEH INSTRUHENTAT ION TRIP SETPOINTS Cl SENSOR q0,5 '/. gq.S%

TOTAL ERROR FUHCTIOHAL UNIT ALLOWANCE 'CA Z TRIP SETPOINT ALLOWABLE VALUE LAO

12. Reactor Coolant Flow-Low 0.6 of loop of loop I

R.R i.'t8 f t;ndiaded ROEAA /~i indi+Cd AOW

13. Sleaa Generator Mater 19. 2 18. 18 I. 5 >38.5 of narrow >38. o narrow Level Low"Low range instrulent range instrument span span

14. Steal Generator Mater 19. 2 6.68 1.5 >38.5X qf narrow >36.8X of narrow Level-Low range instrument range instrument Coincident Mith span span Steam/Feedwater Flow 20. 0 3.4l Note 6 <40K of full <43.lX of full Hisaatch steam flow at RTP"" steam flow at RTP**

15. Undervol tage - Reactor 14. 0 1. 3 0.0 >5148 vol ts >4920 volts Coolant Pumps

16. Underfrequency - Reactor 5.0 3.0 0.0 >57.5 llz >57.3 llz Coolant Peaps

17. Turbine Trip

~

a. Low Fluid Oil Pressure N.A. N.A. N.A. >1000 psig >950 psig

b. Turbine Throttle Valve N.A. N.A. N.A. >)X open >lX open Closure

18. Safety Injection Input N.A. N.A. N.A. N.A. N.A.

from ESF

~~RIP = RATEO lllERHAL POWER

TABLE 2.2-1 Continued TABLE NOTATIONS NTE 1: OVERTEHPERATURE h7 KT

(~+)5) (~~)

(1 + z 1

< KT (K) - (1 KL (~)~

+ t S)

)

T (~q) - T' 1

Kl(P - P') - f,(KT))

Mhere: hT = -Heasured AT by RTO Hanifold lnstruaentation;

~1~ t 1 Lead-lag coapensator on measured AT;

+ qe Tl ~ l2 Tiae constants utilized in lead-lag compensator for hT, ri = a s, zg=35l Lag compensator on Neasured AT; Tlae constants utilized in the lag coI)pensator for AT,.r3 = 0 s; aT ln ted AT at RATEO TIIERHAL I'OWER; I.ll x 4 o.oz2,4

'F;

~1+ x 5 The function generated by the lead-lag coipensator for T dynaaic compensation; zs Tiay constants utilized in the lead-lag compensator for T

, x~ = 20 s, cs = 4 s~

Average teiperature, F; Lag compensator on measured T Tiae constant utilized In the aeasured T avg lag compensator, rq = 0 s;

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1: (Continued)

M

<'88. 'F (Nominal T at RATED THERHAL POMER);

I 0,00lolZ Q K3 /psxg3 M

~ Pressurixer pressure, psig; pl ~ 2235 psig (Nominal RCS operating pressure)>

~ Laplace transform operator, s and fl (hl) is a function of the indicated difference between top and bottom detectors of the pover-range neutron ion chaabers> with gains to be selected based on measured instrument response I during plant startup tests su h that!

CO

-2t.6 5 ly.0%

(1) For qt qb between and + > fl (hl) ~ 0, &ere qt and qb are percent RATED THERHAL POMER in the top and bottom halves of the core respectively, and q + qb is total THERHAL POMER in percent of RATED THERHAL POMEROY

->1.4 %

(2) For each percent that the magnitude of qt - qb exceeds , the hT Trip Setpoint shall be autoaatically reduced by 2.36Z of its value at RATED TH HAL POMER; and

+a.o%

(3) For each percent that the magnitude of qt - qb exceeds ~ , the hT Trip Setpoint shall be autoaaticaliy reduced by of its value at RATED THERHAL POMER.

l.57 .

NOTE 2: The channel's maximum Trip Setpoxnt shall not exceed its computed Trip Setpoint by more O than Lr9X hT span.

2.'

TABLE 2.2-1 Continue~(l TABLE NOTATlONS NOTE 3:

ST ~

.OVERPOMER AT (1 + rrS) (1 + rrS)--< ST o (K~

" Kr'1 ~ + rrS) (1 > rS)

T - Krr T 1 + rrS)

- T" - fr(SI))

Mhere: hT = hs defined In Hote 1, 1 t r 5 hs defined in Hote 1,

+ qa As defined in Note 1, As defined in Hote 1, hs defined in Note 1, AT As fined in Hote 1, 1.079 Ki Ks 0.02/4f for increasing average temperature and 0 for decreasing average temperature, t The function generated by 'the rate lag colllpensator for Tav dynamic 1 rq compensation, Tg Tiae constants utilized in the rate-lag compensator for T, tz avg's

= 10 s, 1

defined in Hote 1, hs defined in Note 1,

TNLE 2. 2-1 Cont inuegl TADLE NOTATIONS fll g NOTE 3: ~ (Continued) o.OOg, Ka /4F for 7 > T" and Kq = 0 for T < T4, As defined in Note 1, Indicated T avg at BATED TllfltNAL POMFA (Calibration teaperature for AT instruaentation, < 588.a'F),

S As defined in Note 1, and fq(hl) = 0 for all AI.

NOTE 4: The channe'I's aaxiauw Trip Setpoint shall not exceed its computed Trip Setpoint by aore than T span.

2.S/ I.I NOTE 5: he sensor error for tewperature is 1.9 and . for pressure.

Cl NOTE 6: The sensor error for steam floe is 0.9, for feed floe is I.5, and for steaa pressure is 0.75.

gi L-'q u.b>>

2. 1 SAFETY LIHITS BASES 2~1~1 REACTOR CORE ~8KPJQQE W I rH 4 ITACHFO INSERT trictions oE this afety 1m1 over eating of th e fue nd poa-sib dding pe1foration which vou lt in the release of E prod-ucts t re actor coolant. Ove 1n h e Euel cLadding cvcntcd by rcstr1ct1 1 operation to n the nu e boiling r erc the heat tranafe ficicnt ge and thc cl aurfa peraturc is slightly above 'ol uration temperatu Operation above t boundary of the nuclcat regime could result 1 czccas1ve g t tures because of set eparturc from nu-e boi HB) and ultant shar C't 1 on 1n tranafer coefEi-cle a not a direc tl urable t er during o p 'on and there-r-

fore POWER and rcac tor ratu re and pres sur e been 1 hrough the W-3 corre The M-3 DHB correlati oped di ct the DHB flu ocation of DHB for axial and nonunifo flux diatz ns. local DHB heat flux r R}

is defined aa t o of culated lux that vould a particular core lo c actual local lux and i at Ive 0 the margin to DNA The minimum val ue c ng steady"state o normal operationa transients, an cipated nts ia limite is value corre" sponds to a obability at confiden all that ll not occur itions.

and is c a an appropriate ma 0 operatin Thc of Figure 2 '-1 show the inta of THERMAL POWER, tor

's C System prcssure and avera e for vhich the minimum ss than 1 ~ 30, or the aver l.py a e vessel exit is equal to nthaLpy of mated liquid.

H

l. 42. for These curves are based on an enthalpy hot channel factor, F4H, of and a j Opj}g. /((el reference cosine vith a peak of L.55 for axial pover shape. An allovance ia ~c} f fe5 Vol yp}4TAG< ~

included for an increase in calculated F4H at reduced pover baaed on the expressions l,4,2, F

[1 + 0.3 (1-P)[ /gal l.OPARPueli And Where P ia the fraction of RATED THERMAL pOMER.

These limiting heat flux conditions are higher than those cal.cuLated for the range of all control rods fully vithdaavn to the maximum allovable control rod insertion assuming the axial pover imbalance is vithin the limits of the fl (41) function of the Overtcmperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on'the Over-tempcraturc 4T trips vill reduce the Setpointa to provide protection consistent vith cor Safct imi p f.g5t y 0,35 (f'p)]Nor YAHTA6<~4uai J

SHEAROH HARRIS - UHIT 1 B 2-1 Amendment No.

lP I'0

> protection to prevent ONB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps. On fndreasfng THERMAL POWER full power above P-7 (a power level of approximately ICE of or a turbine impulse chamber pressure at approximately power equivalent), an automatic Reactor ~ of RATED trip will occur ff'the flow in more than one loop drops below of nominal full loop flow. Above P-8 qo.5% SHEARON HARRIS - UNIT 1 B 2-5 sl rt P 'I 'P / LIMITING SAnyy SV~ SETTINGS BASES Reactor Coolant Flow Continued eo.<< (a power level of approximately 49~ of RATED THERMAL POWER) an automatic Reactor if trip will occur the flow in any single loop drops below . ~ of nominal full loop flow. Conversely, on decreasing power between P-B and the P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked. Steam Generator Water Level The Steam Generator Water Level Low Low trip protects the reactor from loss of heat sink in the event of a sustained steam/feedwater flow mismatch resulting from loss of normal feedwater. The specified Setpoint provides allowances for starting delays of the Auxiliary Feedwater System. Steam/Feedwatet Flow Mismatch and Low Steam Generator Wate~ Level The Steam/Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Trip System. 'his trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam/Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greate~ than or equal to 1.627 x 104 lbs/hour. The Steam Generator Low Water level portion of the trip is activated when the water level drops below 38.%, as indicated by the narrow range instrument. These trip values include sufficient 'allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam gener ators is minimized. Undervolta e and Underfreouen - Reactor Coolant Pump Buses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated ln the Underfrequency and.Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after. the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second. On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by the loss of P-7 (a power level of approximately LGX of RATED THERMAL POWER or a turbine impulse chamber pressure SHEARON HARRIS - UNIT 1 8 2-6 'l < g( I "1 l~ n %14 REACTIVITY CONTROL SYSTEMS 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3 '.3.1 All shutdown and controL rods shall be OPERABLE and .positioned within + 12 steps (indicated position) of their group step counter demand position. APPLICABILITY: MODES 1+ and 2*. ACTION:

a. With one or more rods inoperable duc to being immovable as a result of excessive friction bourse'.

or mechanicaL interference or known to bc untrippable, determine that thc SHUTDOWN MARGIN requirement of Specification 3.1 ~ 1 1 ~ is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 With more than one rod misaligncd from the group step counter demand position by more than + 12 steps (indicated position), be in HOT STANDBY within 6 hours' ~ With more than one rod inoperable, due to a rod control urgent failure alarm or obvious electrical problem in the rod control system existing for greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , be in HOT STANDBY within the foLLowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one rod trippable but inoperable duc to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than + 12 steps (indicated position),

POWER OPERATION may continue provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months '. The rod is restored to OPERABLE status within the above alignment requirements, or 2 ~ The rod is declared inoperable and the remainder of the rods in the group with the inoperabLe rod are aligned to within + 12 steps of the inoperable rod while maintaining the rod sequence Spsc+coHDh 3.l.3,b and insertion Limits of The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3~ The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then'ontinue provided that: a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days', this reevaluation shall confirm that the previously analyzed results of these accidents

See Special Test Exceptions Specifications 3.10.2

~ ~ ~ ~ and 3.10.3. ~ ~ ~ SHEARON HARRIS - UNIT 1 3/4 1-14 Amendment No. I ,4 at% E $ = 4 l % 'I s L REACTIVITY CONTROL SYSTEMS ROO DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to . seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: 2'1

a. T greater than or equal to 5514F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and Z. ACTION:

a. With the drop time of any rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceed-ing to MODE 1 or 2.

b. With the rod drop times within limits but determined with two reac.or coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 6/i" of RATED THERMAL POWER.

SURVEILLANCE RE UIREMENTS

4. X.3.4 The rod drop time of shutdown and control rods shall be demonstrated through measurement prior to reactor criticality:

For all rods following each removal of'he reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and C. At least once per 18 months.

SHEARON HARRIS - UNIT 1 3/4 I-5 REACTIVITY CONTROL SYSTEMS SHUTQOWN ROQ INSERTION LIMIT LIMITING CONQITION FOR OPERATION tahar, 3.1.3.5 All s oporo+ing imi+s R f ly withdraw as speciFied i'~ +e Cone APPLICABILITY: 0 S an 2 ACTION: Mith a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a. Fully withdraw the rod, or

~

b. Qeclare the rod to be inoperable and apply Specification

3. 1.3. 1.

SURVEILLANCE RE UIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:

a. Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or 0 during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

See Special Test Exceptions Specifications 3.10.2 and 3.10.3. Wth Keff ff, greater than or equal to 1. SHEARON HARRIS - UNIT 1 3/4 1-20 REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITINC CONDITION FOR OPERATION s eciF>ed In+e 3.1.3.6 The control ban shall b limited in physical insertion as 1! Op 'qu APPLICABILITY: MODES 1* and 2* ~. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal. to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using Ae. inser+ion i~i'pecifi'ed in +e Core Operah'g Limi+s Report C ~ Be in at leas STANDBY wxt in 6 ours.

SURVEILLANCE RE UIREMENTS 4.1.3.6 Thc positron of each control bank shall be determined to be within ~ ~ the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . <See Special Test Exceptions Specifications 3.1Q.2 and 3.10.3. ~With K ff greater than or equal to l. SHEARON HARRIS - UNIT 1 3/4 1-21 Amendment No. ~ I ~t~ 71' I l,)q aP <~t 7p ~ ,I L 220 (1 >86) 5 120 100 8 80 20 (O. ) 0 0.00 0.10 0 OM OAO 0.50 0. 0.70 0.80 0.90 1.00 RA'lED DERMAL FIGURE 3.1-2 3NLETEb) SHEARON HARRIS - UNIT 1 3/4 1-22 Amendment No. V) 'f 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITINC CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within.'. the acceptable operational space fo e Axial Offset Control (RAOC) operation, or sssfsoiNeii sseeQr saOpsaesgi me'r Raper>

b. within a ban about t a e AFD during Base Load operation ~

o~Specff~ ir +c Ccrc Oper&ig Litvaks Hepor+. APPLICABILITY: MODE 1 above 50Z of RATED THERMAL POWER~, ACTION: For RAOC operation with the indicated AFD outside of the limits, either. s ascii a ected

1. Restore the indicated AFD to within the limits within 15 minutes, or

2. Reduce THERMAL POWER to less than 50Z of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55Z of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

be For Base Load operation above APL with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target AFD, either:

1. Restore the indicated AFD to within the target band limits within 15 minutes, or

2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes.

C~ THERMAL POWER shall not be increased above 50Z of RATED THERMAL POWER unless the indicated AFD is withi the spec'f fled lifffftskw @ROC opera'hOn, NpSee Special Test Exception 3.10.2 ~APL is the minimum allowable power Level for Base Load operation and will be provided in th per Specification 6 '.1.6 ~ gore Qperflkirlfj JimHs Rep~ SHEARON HARRIS - UNIT 1 3/4 2-1 Amendment No. 4' V L I, ~4 .e lpga '~ ~ < ~ 'J V fO cl foe POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its li.mits during POWER OPERATION above 50Z of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE cxcore channel/

J(t least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

b. Monitoring and Logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at Least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shaLL be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when two or more OPERABLE excore channels are indicating the AFD to be outside the Limits. 4.2.1.3 When in Base Load operation, the target AFD of each OPERABLE excore channel shall be determined by measurement at Least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 When in Base Load operation, the target AFD shall be updated at Least once per 31 Effective Full Power Days by either determining thc target AFD in conjunction with the surveillance requirements of Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable. SHEARON HARRIS - UNIT 1 3/4 2-2 Amendment No. I 0 ga, I ~ ' 0 gent I ~ 'I CI FIC @RE B.a-l DELETED 120 110 (- 3> ) , 00 QO 20 10 10 0 20 30 40 FLUx mmamcz Q) FIGURE 3.2-'f (D<L<~E+ 3/4 2-4 Amendment No. I II 4e J w' hq F 1 '" ~ 4i l 5~ POWER DISTRIBUTION LIHITS 3/4 ~ 2 ~ 2 HEAT FLUX HOT CHANNEL PhCTOR - P (Z) LIHITINC CONDITION POR OPERATION 3 '.2 FQ(Z) shall be limited by the following relationships: a.w5 PQ(Z) C [K(Z)) FOR P ~ 0 ' P 4,90 PQ(Z) <,( [K(Z)j FOR P < 0 ~ 5 Where: P ~ THERNAL POWER, and RATED THERMAL POWER K(Z) > the function obtained from Pigure 3.2-2 for a given core height location. APPLICABILITY! NODE 1. ACTION t With FQ(Z) exceeding its limit: a~ Reduce THERMAL POWER at least 1Z for each 1Z FQ(Z) exceeds the .limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within thc next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) subsequent POWER OPERATION may proceed provided the Overpower hT Trip Setpoints have been reduced at least 1Z for each 1Z FQ(Z) exceeds the limit.

b. Identify and correct the cause of th>> out-of-limit condition prior to increasing THEBHAL POWER above the reduced limit re-quired by ACTION a., above) THERHAL POWER may then be increased provided FQ(Z) is demonstrated through incore mapping to be within its limit.

SHEABOM HARRIS - UNIT 1 3/4 2-5 Amendment No. l 'I W 0 POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not appLicable. 4.2.2.2 For RAOC operation, FQ(Z) shall be evaluated to determine if it is within its Limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5Z of RATED THERMAL POWER.

b. Increasing the measured F>(Z) component of the power distribution map by 3Z to account for inufacturing tolerances and further increasing thc value by 5X to account for measurement uncertainties'erify the requirements of Specification 3.2.2 arc satisfied.

C ~ Satisfying the following relationship.'Ã5 F M(Z) < x K(Z) for P > 0.5 PxWZ) s,WS F "(Z) < x K(Z) Eor P < 0 5 Q where F (Z) is the measured Fz(Z) increased by the allowances for manufac ring tolerances and 2ieasurement uncertainty, is the. FQ limit, K(Z) is given in Figure 3.2-2, P is the Eraction of RATED THERMAL POWER, and W(Z) is the cycle dcpcndcnt function that accounts Eor power distribution transients encountered during normal operation. This function is given in the as per Specification 6.9.1.6. Core.operabng Limni* Rcpo<4

a. Measuring F (Z) according to the following schedule:

Q 1 ~ Upon achieving equilibrium conditions after exceeding by LOX or more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined,* or

2. At least once per 31 Effective Full Power Days, whichever occurs first.

During power escalation at the beginning of each cycLe, power level may be increased untiL a power level for extended operation has been achieved and a power distribution map obtained.

SHEARON HARRIS - UNIT 1 3I4 2-6 Amendment halo. g+ g 1 I I I g4 0 POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued)

e. With measurements indicating maximum has increased since the previous determination of F~ (Z) either of the following actions shall be taken.')

F (Z) shalL be increased by 2X over that specified in S ecification 4.2.2.2c. or

2) F (Z) shall be measured at least once pcr 7 EEfective Full Plier Days until two successive maps indicate that FM maximum (Z) is not increasing.

E. With the relationships specified in Specification 4.2.2.2c above not being satisfied:

1) Calculate the percent F~(Z) exceeds its limit by the following expression'.

F(Z) illax imum g q5 P ~ x W(Z) x K(Z) x 100 for P h 0.5 F"(Z) x W(Z) maximum x 100 Eor P < 0.5 ede e e K(Z)

2) Onc of the foLlowing actions shaLl, be taken:

a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits ef in+6 by 1X AFD Eor each percent F~(Z) exceeds its limits specH)cd ~rg gporndng as determined in Specification 4 '.2.2f.l). Within 8 girnits Repor+ hours, reset the AFD alarm sctpoints to these modified LimLts l or b) Comply with the requirements of Specification 3.2.2 for F<(Z) exceeding its limit by the percent caLculated above, c) Verify that the requirements of Specification 4.2.2.3 Eor Base. Load operation are satisfied and enter Base Load operation. SHEARON HARRIS - UNIT 1 3/4 2-7a Amendment Ke.g r,$ is ~ 1 ~ ~ ( I 'll" ~1 ~v POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued)

g. The limits specified in SpeciEications 4.2.2.2c, 4.2.2.2e, and 4.2.2.2E above are not applicable in the EoLLowing core plane regions:

1. Lower core region from 0 to 15X, inclusive.

2. Upper core region from 85 to 100Z, inclusive.

4.2.2.3 Base Load operation is permitted at powers above APL if the foLlowing conditions are sati.sfied:

a. Prio~ to entering Base Load operation, maintain THERMAL PNER above APL" and Less than or equal to that aLLowed by Specification 4 '.2.2 Eor at Least thc previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain Base Load operation surveillance (AFD within

~g, iif7fl+s ape C H"led 4i4&s~) during this time period. Base Load operation is then in ~s 5'pc permttted providing 2(ERHAL PORER is maintained between APL" and QPsg o.fin/ APL or between APL and lOOX (wbitbever is most limiting) and L mHs Rc~> surveillanne is maintained pursuant to gpeoifioation 4.2.2.4. APLPii is defined as: g y5 'nimum ~ I ~ 100X F (Z) x V(Z) where: F (Z) is the measured F (Z) increased by the allowances for man factuling tolerances and measurement uncertainty. The F limit <qr, is . . K(Z) is given in Figure 3.2<<2. M(Z)BL is the cycl dependent function that accounts for limited power distribution transients encountered during Base Load operation. The function is given in thc Specification 6.9.1.6 ~ gow Qpsr&in9 LtrnitS Repot+

b. During Base Load operation, if the THERMAL POSER is decreased below APL then the conditions oE 4.2.2.3.a shall be satisEied before re-entering Base Load operation.

4.2.2.4 During Base Load operation F~(Z) shall be evaluated to determine if it is within its limit by!

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APL

b. Increasing thc measured F<(Z) component oE the power distribution map by 3X to account for i4nufacturing toLcrances and further increasing the value by SZ to account Eor measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied.

SHEARON HARRIS - UNIT 1 3/4 2"7b Amendment No. 4'e~ qa h I I I <C t tI 4 Jka POWER DISTRIBUTION LIMITS SURVEILLANCE RE VIREMENTS (Continued) C ~ Satisfyin e Eot.lowing relationship'. ZA x K(Z) frP>APLND P x WZ g,45 where'F M Q (Z) is the measured F Q (Z) ' ~ The F Limit is ~. K(Z) is given in Figure 3.2-2. P is the Eraction of RATED THERMAL POWER. W(Z)BL is the cycle dependent function that accounts Eor limited pover distribution transients encountered during normaL ~~ operation. This function is given in the as per Specification 6.9.1.6. Core Op~+ng Lirni+s @pod~

d. Measuring F (Z) in conjunction vith target flux diEference determinatiIIn according to the Eolloving scheduLe',

1. Prior to entering Base Load operation after satisfying Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermaL power having been maintained above APL for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and

2. At least once pcr 31 effcctivc full power days.

With measurements indicating F (Z) maximum [ ] ) M has increased since the previous determination F (Z) either of the Eolloving actions shall be taken: M 1 ~ F (Z) shaLL be increased by 2 percent over that specified in 4 ' 'o4 icy or 2~ F (Z) shall bc measured at least once per 7 EFPD until 2 successive maps indicate that F (Z) maximum fK Z ] is not increasing.

f. With the relationship speci. fied in 4.2.2.4.c above not being satisfied, either of the Eolloving actions shall be taken:

Place the core in an equilibrium condition vhere the limit in 4.2.2.2.c is satisfied, and remeasure F (Z) , or SHEARON HARRIS - UNIT 1 3/4 2-7c Amendinent No. ei-4r ~% lml 'w L>> POMER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued) 2., Comply with the requirements of SpeciEication 3.2.2, Eor F~(Z) exceeding its limit by the percent calculated with the following expression: [(max. of A/5 F P (Z) x M(Z)BL x K(Z) ] ) -1] x 100 Eor P > APL ND

g. The limits specified in 4.2.2i4.c, 4 '.2.4.e, and 4.2.2.4.f above are not applicable in the Eollowing core plane regions:

1. Lower core region 0 to 15 percent, inclusive.

2. Upper core region 85 to 100 percent, inclusive.

4.2.2.5 When F~(Z) is measured Eor reasons other than meeting the requirements of Specification 4.2.2.2 an overalI measured F (Z) shall be obtained from a power distribution map and increased by 3X 3o account for manufacturing tolerances and further increased by 5X to account Eor measurement uncertainty. SHEARON HARRIS - UNIT 1 3/4 2-7d Amendment Ho. 0 E$ I 'i' MAL Fq' 2.32 CORE HHGHT O.QOQ .QOO 1.000 0.840 0.047 6 8 10 CORE HEIG (FT) FlGURE: 3.2-2 K(Z) LOCAL AXIAL PENAITY FUNCTION FOR Fq (Z) tA 0.0 0.1 o.e Elevation Ihmal )zed Peakln Factor 1$ 0.0 1.0 6.0 l.o O.4 )2.0 0.925 OA 4 8 b COfK BBfAKN(FED) g tl l ~'~ 4 Ir r . ' lI P POWER DISTRIBUTION LIMITS 3/4.2e3 RCS fllN RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITINC CONDITION FOR OPERATION 3.2.3 The indioarad Reattot Coolant shall. be maintained as follova: 0 attn (RCS) total floe tata and )g

a. Meaa ed RCS floe atc > $ 05~!'pm x (1.0 + Cl), and i.4 2 b.

FnH ~a [l 0 + 0.3(I.C-P) for LOPAR +na),ond F < ~ qS ~l,O + O.aSPl.q-Ping@VZ~mCez 6~. P ~ THERMAL POWER, and RATED THERMAL POWER $ 5Pl. AcE Measured es of N F>H ob 'd by using movable in e Wiv'H detect tg tain a er ribut map, d t aaured AvlAcHE2 aof F>Mah e used for c riaon above ce the ta for ~t xQ$ <kT ~ 4 alue abov acc measur ent N F>H. a nce inca Cl ~ Meaaurcmcnt uncertainty for core flow aa described in the Bases. APPLICABILITY! MODE 1 ACTION: With RCS total floe rate or outside the above limits! a~ Within 2 hours either: Restore RCS total floe rate and to within thc above limits, or

2. Reduce THERMAL POWER to lese than 50Z of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55Z of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SHEARON HARRIS - UNIT 1 3/4 2-9 Amendment No. INSERT TO PAGE 3/4 2-9: F<H - Nuclear enthalpy rise hot channel factor obtained by using the movable incore detectors to obtain a power distribution map, with the measured value of the nuclear enthalpy rise hot channel factor (F N ) increased .At by an allowance of 4X to account for measurement uncertainty. ~ 1>'49<<c<t ~, 4' C,) P 1 p~f tg POWER DISTRIBUTION LIMITS 3/4.2.5 ONB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following ONB-related parameters shall be maintained within the following limits:

a. Indicated Reactor Coolant System T < . oF after addition for instrument uncertainty, and

b. Indicated Pressurizer Pressure > psig" after subtraction for instrument uncertainty.

2l85 APPLICABILITY: MODE 1. ACTION: With any of the above parameters exceeding its indicated limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5 of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . SURVEILLANCE RE UIREMENTS 4.2.5 Each of the parameters shown fn Specification 3.2.5 shall be verified to be wfthin its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . limit is not applicable during either a Thermal Power Ramp in excess of ~ "This of Rated Thermal Power per minute or a Thermal Power step change in excess &OX Rated Thermal Power. SHEARON HARRIS - UNIT 1 3/4 2-14 0 J 4', t TABLE 3.3-1 Continued ACTION STATEMENTS Continued ACTION 3- With the number of channels OPERABLE one less thah the Minimum Channels OPERABLE requirement and with the. THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Inter lock)

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below'10" of RATEO THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above ICE of RATED THERMAL POWER. ACTION 4.- With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes. I ACTION 5- a. With the number of OPLRABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoper 3.L1.2~ the shutdown margin requirements of Specification IZ hours thereafter. i i 1 I ~~ able channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers, and verify compliance with

b. With no channels OPERABLE, open the Reactor Trip System breakers within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and suspend all operations involving positive reactivity changes. Verify compliance with the 3.l.L.R~

MITOOWN MARGIN requirements of Specification per I2 hours thereafter. H f 1 %RC~ ACTION 6- With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7- With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive annunciator window(s) that the inter lock is in its required state for the existing plant condition, or apply Specification 3.0.3. SHEARON HARRIS - UNIT I 3/4 3-7 TABLE 3.3-4 Continued ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRljHENTATION TRIP SETPOINTS SENSOR TOTAL ERROR UNCTIONAL UNIT ALLNANCE TA ~S. TRIP SETPOINZ ALLOMABLE VALUE

9. Loss-of-Of fsite Power

a. 6.9 kV Eaergency Bus N.A. N:A. N.A. > 4830 volts > 4692 volts with Undervol tage Priaary with a < 1.0 a tile delay second tiwe < 1.5 seconds delay.

b. 6.9 kV Eaergency Bus N.A. N.A. N.A. > 6420 volts > 6392 volts Undervoltage- with a < 16 with a. time Secondary second tiae delay < 18 seconds delay (with (with Safety Safety Injection).

Injection). > 6420 volts > 6392 volts with a < 54.0 with a < 60 second tise second time delay (with- delay (with-out Safety out Safety Injection). Injection).

10. Engineered Safety Features Actuation Systea Interlocks

a. Pressurizer I'ressure, P-11 N.A. N.A. N.A. > 2000 psig > 1986 pgg Not P-11 N.A. N.A. N.A. < 2000 psig < 2014 psig

b. Low-Low T, P"12 N.A. N.A. N.A. > 553 F

if' tl fi Y 3/4.2 POWER DISTRIBUTION LIHITS Wc d~iS< DgBR. value BASES Thc specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Hoderate Frequency) events by: (1) maintaining the minimum DNBR in thc core greater than or equal to L during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet tcmperaturep and cladding mechanical proper-ties to within assumed design criteria. In addition, limiting the peak Linear power density during Condition I events provides assurance that the initiaL conditions assumed for the LOCA analyses arc met and the ECCS acceptance. criteria limit of 2200'F 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, is defined is the maximum local heat flux. on thc surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, aLLowing for manufacturing toleranccs on fuel pellets and rods> N F4H 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; 3/4.2.1 AXIAL F DIFFERENCE g.45 The limits on FLUX DIFFERENCE (AFD) assure that the F~(Z) upper bound envelope of times the notmalized axial peaking dattoz ts not extended during either normal operation or in thc 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 Limits and should be inserted near their normal position for steady-state operation at high power levels ~ Thc value of the target flux difference obtained under these conditions divided by the friction of RATED THERHAL POWER is the TARGET AFD at RATED THERHAL POWER for the associated core burnup conditions'ARGET AFD for other THERHAL POWER l.evels are obtained by multiplying the RATED THERHAL POWER value by the appropriate fractional THERHAL POWER levele The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. SHEARON HARRIS - UNIT 1 B 3/4 2-1 Amendment No. 'ky ti Pp S ~ 8 4 4 KNER DISTRIBUTION LIHITS BASES AXIAL FLUX DIFFERENCE (Continued) speciI'iel in %he SK Opem+nfI L)Inly RePog'q At povex levels belov APL , the limits on AFD arc i.e., that defined by the RAOC operating procedure and limits. These limits were calculated in a manner such that expected operational transients, e.g., load follow operations, vould not result in the AFD deviating outside of those limits. Hovcver, in the event such a deviation occurs, the ahox't period of time alloved outside of the Limits at reduced power levels vill not result in signifitant xenon redistribution auth that the envelope of gashing peters uouid change sufficiently to prevent operation in the vicinity of the APL power Level. ggplac+ At over Levels cr than APL tvo modes f operation are rmisaiblex ~)4h D ts of are ined b ure ~ , and e a ~gale~ pe on, is de ma ten c o th 3X gggkf a et val The RAOC opergxng proce ure above APL xs t e same aa that efine for operation beLov APL . Hovever, it is possible vhcn following extended load folloving maneuvers that the AFD Limits may result in restrictions in the maximum aLloved pover or AFD in order to guarantee operation vith F~(Z) less than ita limiting value. To allow operation at the maximum permissible value, the Base Load operating procedure reatri s t indicated to relativ target ver svin oi AFD sma nd and or Base Load operatxon, xt ia expected that the plant vx operate a, within the target band. Operation outside of the tax'get band for the short time period alloved vill not result in significant xenon redistribution such that the envelope of peaking factors vould change sufficiently to prohibit continued operation in the pover zegion defined above. To assure there is no residual xenon redistribution impact from past operation on thy Base Load operation, a 24-hour vaiting period at ~ pover level above APL and alloved by RAOC is necessary. During this time pexiod, load changes and rod motion arc restricted to that allowed by the Base Load pz'ocedure. After the vaiting period, extended Base Load operation ia permissible: The computer determines the one-minute average of each of the OPERABLE excorc detector outputs and provides an alarm message iaxncdiately if the AFD for tvo or morc OPERABLE excore channeLs are! 1) outside the alloved 4I pover operating space (for RAOC operation), or 2) outside the acceptable AFD target band (for Base Load operation). active greater than') These alarms are 50X of RATED THERMAL PIKER (for RAOC operation), when or power

2) hPLg (for Base Load operation). Penalty deviation minutes fox Base Load operation are not accumulated based on the short period of time during vhich operation outside 'of the target band is alloved.

SHEARON HARRIS - UNIT 1 B 3/a 2-2 Amendment No. g 'I <4 INSERT TO PAGE B 3/4 2-2:

1) RAOC with fixed AFD limits as a function of reactor power level and 2) Base Load operation which is defined as the maintenance of the AFD within a band about a target value. Both the fixed AFD limits for RAOC operation and the band for Base Load operation are specified for each reload cycle in the CORE OPERATION LIMITS REPORT per Specification 6.9.1.6.

q4 POWER DISTRIBUTION LIHITS BASES 3/4.2.2 AND 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The Limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak locaL power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance, criteria 1 lml t ~ Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overLapping groups as described in Specification 3.1.3.6; SHEARON HARRIS - UNIT 1 B 3/4 2-2a Amendnent No.

tN ak a s i 4'tl4 a 0 U 1 I'(v i.54 for LOPAR fuel yn J PM'ISTRIBUTION LIHITS l.59 for VANTAGE ~ fuel ~ BhSES HEhT FLUX HOT CHhMMEL FhCTOR hMD RCS FIAT RhTE hND NUCLEhR EMTHALPY RISE HOT CHhMNEL FhCTOR Continued

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained) and

d. Thc axial power distribution, expressed in terms of hXIhL FLUX DIFFEREMCE, is maintained vithin the limits.

vill. be maintained vithin its limits provided Conditions a. through d. ve are main a'ncd. The combinations of the RCS ELov requirement and the measurement o ensures that the cal ulatcd DMBR vill not be belov the design DMBR value. The relaxation oE. as a function of THERNhL PNER allovs changes in the radial povcr shape for all peraissib1e rod insertion Limits ~ F ese vaguer a F>H is evaluated as being less than oy aqua o ~ used ia the various accident analyses vhere F+ influences parameters other than DMBR> c.g., peak cLad temperature, and thus xs the maximum "as measured" value alloved. Fu rod bovi this rcduct compl cly ffset rcd es n in the any r e value neri.c mar bov pe '.ltics. ratio. plicable o e gener c e edit va ava'ble margins, t taling of rod to 9. o set DMBR v and any oth penaltics s pr ante in PShR Se 'on 4.4.2 '.4 ~ is rgin include c folio xng! Des'imit DMB of .30 vs 1 8,

b. Cri acing (K of 0 46 vs .059,

~ C ~ rma Diffus'on Coeffi e oE 0.038 vs 059,

d. BR Mul 'pl' of 0 '6 vs .88, and

c. itch redu on.

When an F< measurement is taken, an allovancc for both experimental error and manufactu&iag tolerance must be made. hn allovancc of 5X is appropriate for a EuLL-core map taken with the Incorc Detector Flux Happing System, and a 3X allowance is appropriate for manufacturing tolerance. The hot channel factor F~(Z) is measured periodically and increased by a cycLe and height dependent poser factor appropriate to either RhOC or Base Load operation, 'M(Z) or 'M(Z)BLi to provide assurance that the limit on the hot channel factor, P~(Z), xs met. M(Z) accouats for the effects of normaL opcratio'n transicBts and. vas determined from expected pover control maneuvers over the fuLL range of burnup conditioas in the cora. W(Z)BL accounts for the more restrictive operating Limits allowed by Base Load operation vhich result in less severe transient values. The M(Z) function for normaL operation is provided in thc per Speci.fication 6.9.1.6. pope 4)pergking Limits R8)el+ SHEhROM HhRRIS - UMIT 1 B 3/4 2-4 Amendeent No. g INSERT TO PAGE B 3(4 2-4: Hargin is maintained between the safety analysis limit DNBR and the design limit DNBR. This margin is more than sufficient to offset any rod bow penalty and transition core penalty. POWER DISTRIBUTION LIHITS BASES HEAT FLUX HOT CHANNEL FACTOR ANO RCS FLOW RATE ANO NUCLEAR ENTHALPY RISE HOT HANN L A antinued iS When RCS flow rate em& measured, na additional allowance) neces-sary prier ta caaparisan nith the liaitp at Speciticetian 3.2.3. A measurement error of 4X for F has een allowed for in determination of. the desi n ONBR value g normal RCS flcwrate error af will be included in Cq, w ich wi e mo ified as discussed below. The measurement error for RCS total flow rate is based upon per forming a precision heat balance and using the result to calibrate the RCS flaw rate ind'icators. Potential fouling of the feedwater venturi which might not be detected cauld bias the result from the precision heat balance in a nan-conservative manner. Therefore, a penalty of O.i for undetected fouling of ~ the feedwater venturi, raises the nominal fIow measurement allowance, Cz, to far no venturi fouling. Any fouling which might bias the RCS flow rate measurement greater than O.l" can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fo~ling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned ta eliminate the fouling. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect anly flaw degradation that could lead ta operation outside the accept-able region of operation. 3/4.2.4 UAORANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and period-ically during power operation. The limit of 1.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of L025 can be tolerated before the margin for uncertainty in F~ is depleted. A limit of LD2 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. The 2-hour time allowance for operation with a ta tilt condition greater than identification and correction 1.02 but less than 1.09 is provided allow of a dropped or misaligned control rcd. In the event such action does nat carr~et the tilt, the margin for uncertainty on F~ is reinstated by reducing the maximum allowed power by 3" for each percent of tilt in excess of 1. SHEARON HARRIS - UNIT 1 a S/4 2-S ~h 4' !- h 4i'* h V fI hp IT POWER OISTRIBUTION LIMITS BASES OUAORANT POWER TILT RATIO Continued For purposes of monitor fng /VAGRANT POWER TILT RATIO when one excora detector is inoperable, the moveable fncore detectors are used to confirm that the normalized symmetric power distribution fs consistent with the gUAORANT POWER TILT RATIO. The fncore dete~r monitoring fs done with a full fncore flux map or Cwo sets of four symmetric thimbles. The preferred sets of four sym-metric thimbles is a unique set of eight detector locations. These locations are C-S, E-5, E-11, H-3, H-33, L-5, L-I1, N-B. If other locations must be used, a special report ta NRR should be submitted wfthfn 30 days. in accozdanca with~ 10CFR50.4. 3/4. 2. 5 ONB PARAMETERS The limits on the ONB-related parameters assure Chat each of the parameters are maintained within the normal steady-state envelope af operation assumed in the transient and accident analyses. The limits are consistent with the ini- 'ial FSAR assumptians and have been analytically demonstrated adequate to in a minimum ONB throughout each analyzed transient. The indi-cated Tav va ue an the indicated Pressurize~ Pressure value are comPared to analytical limits of ~ d sfg, respectfvely, ance for measurement uncertainty. >Per o Is indudcd The XZ-hour periodic surveillance of these parameters through fns rument rea"- out is sufficient to ensure that the parameters are restared within their limits following load changes and other expected Cransient operation. +a+ is equi ko or korea<~ +<<+e design 2)HBR va,lue, SHEARQN HARRIS - UNIT 1 B 3/4 2 6 % ~4 ill" fq ~r S l C~ 3/4. 4 REACTOR COOLANT SYSTEN BASES 3/4.4.1 REACTOR COOLANT LOOPS ANO COOLANT CIRCULATION The plant is designed to operate with all reactor coolant Ioaps in operation and maintain DNBR above during all normal operatians and anticipated tran-sients. In NODES 1 an with one reactor coolant loop not in operation this specification requires at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 4c. esign DNHR value In NODE.3, two reactor coo an oops prov de sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufffcient heat removal capacity if a bank withdrawal accident can be prevented, i.e.; by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times. In NODE 4, and in NODE 5 with reactor coolant looped filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for remov-ing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In NODE 5 with reactor coolant loops not filled, a single'RHR loop provides sufficient heat removal capability far removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat remov-ing component, require that at least two RHR loops be OPERABLE. The ape~ation of one reactar coolant pump (RCP) or one RHR pump provides ade-quate flow to ensure mixfng, prevent stratification and produce gradual re-activity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with baron reduction will, there-fore, be within the capabflfty of operator recognftion and control. The restrictions on starting an RCP with one or more RCS cold legs less than or equal ta 335'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G ta 10 CFR Part 50. The RCS will be protected against over pressure transients ynd will not exceed the limits af Appendix G by restricting starting of the RCPs to when the secondary ~ater temperature of each steam generator is less than 50oF abave each of the RCS caid leg temperatures-3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate ta prevent the RCS from being pres-surized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 380,000 Ibs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occu~ during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protectian against RCS overpressurization at low temperatures. SHEARON HARRIS -. UNIT 1 B 3/4 4-1 I < 4'v 'I 1 4 Wp 'I l'j lllII 1 l,i4 Jf i h P tM ADMINISTRATIVE CONTROLS INC FACTOR LIMIT ~ O'EPLAQE WITH A~AgHEb INSERT 6 ' .6 The M(Z) F tio for RAOC and B e Loa operation d the vaLue Eor APL (aa required) shall established or each Load c e and implemented prior o use. The meth olo used to gencrat e M(Z) functions RAOC aad Base Lo operation the value for APL shall be those pr iou y revieved a approved thc NRC.+ If cha ca these method are de d access y, they vilL be val tcd in accord ce vit 10 CFR 50. and subai ed to e NRC Eor rcvicv ind ap ovaL prior o their us if thc hange is date 'n to involve an cvicvcd s fcty qu tion or if su a ge vould requir amendment of pr iously submi tcd cumcntation. A rcport contgni the M(Z) function or and Base d operat a and the value for APL aa cquircd) shall provi d to the C in accord ce vith 10 CFR 50.4 v hin 30 ays after c h cycle in ial c ticality. hny info tion needed t supp t Q(Z), W(Z)BL> APL vill be by request from the NRC and need not be ncludcd ia this r tto SPECIAL REPORTS 6.9.2 Special reports shalL be submitted to the Regional Administrator of the Regional Office of the NRC vithin the time period specified for each report. 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulationa, the folloving records shall be retained Eor at least the minimum period indicated. 6.10.2 The folloving records shill be retained for at least 5 years! ae Records and logs of unit operation covering time interval it each pover leveled

b. Records aad logs oE principal maintenance activities, inspections, repair, and replacement of principaL items of equipment related to nuclear safety)

Co hll REPORTABLE EVENTSI do Records of surveillance activities, inspections, aad cilibrationa required by these Technical Specifications).

M - 6, "R ation o atant Ax fact ol-P~ Su ance T x 1 cifici '

SHEARON HARRIS - UNIT 1 6-24 Amendment No.~ INSERT TO PAGE 6-24: CORE OPERATING LIMITS REPORT 6.9.1.6 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. The analyticaL methods used to determine the core operating Limits shall be those previousLy reviewed and approved by NRC in WCAP-10217"A t II Relaxation of Constant Axial Offset Control F~ SurveiLlance Technical Specification," 1983 (for AFD limits, APL" , and W(r) Functions) and in WCAP-9273-A, "Westinghouse Reload Safety Evaluation Methodology," 1985 (for Control Bank Insertion Limits). The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS Limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided within 30 days of their implementation, for each reload cycle to the NRC in accordance with LOCFR50.4. ADHINISTRATIUE CONTROLS (Continued)

e. Records of changes made to the procedures required by Specifica-cion 6.8.1;

f. Records of radioactive shipments>

g. Records of seaLed source and fission detector leak tests and results; and Q'eoRb RETEHTIDP4 SHEARON HARRIS - UNIT 1 6-24a Amendment No.

ATTACHHENT 3 NON-LOCA ACCIDENT ANALYSIS FOR THE SHEARON HARRIS NUCLEAR POWER PLANT TRANSITION TO 17 x 17 VANTAGE 5 FUEL (277CRS/lah) TABLE OF CONTENTS Section Descri tion ~Pe e

15. 0 ACCIDENT ANALYSES 15.0-1 15.1 CONDITION I - NORMAL OPERATION AND OPERATIONAL TRANS IENTS 15.1-1 15.1.1 Optimization of Control Systems 15.1-3 15.1.2 Initial Power Conditions Assumed in Accident Analyses 15.1-3 15.1.2.1 Power Rating 15.1-3 15.1.2.2 Initial Conditions 15.1-4 15.1.2.3 Power Distribution 15.1-5 15.1.3 Trip Points and Time Delays to Trip Assumed in Accident Analyses 15.1-5 15.1.4 Rod Cluster Control Assembly Insertion Characteristic 15.1-6 15.1.5 Reactivity Coefficients 15.1-7 15.1.6 Plant Systems and Components Available for 15.1-8 Mitigation of Accident Effects 15.1.7 Fission Product Inventories 15..1-8 15.1.8 Residual Decay Heat 15.1-8 15.1.9 Computer Codes Utilized 15.1-8 15.1.9.1 FACTRAN 15.1-9 15.1.9.2 LOFTRAN 15.1-9 15.1.9.3 LEOPARD 15.1-10 15.1.9.4 TURTLE 15.1-10 15.1.9.5 TWINKLE 15.1-11 15.1.9.6 THING 15.1-11 15.1.10 References 15.1-11 15.2 CONDITION II - FAULTS OF MODERATE FREQUENCY 15.2-1 15.2.1 Uncontrolled Rod Cluster Control Assembly Bank Hithdrawal from a Subcritical Condition 15.2-3 15.2.1.1 Identification of Causes and Accident Description 15.2-3 15.2.1.2 Analysis of Effects and Consequences 15.2-5 1575v:1D/11 1 188 15. 0-'i

'1 4 TABLE OF CONTENTS (Cont) Section Descri tion Pa<ac 15.2.1.3 Results 15.2-7 15.2.1.4 Conclusions 15.2-8 15.2.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power 15.2-9 15.2.2.1 Identification of Causes and Accident Description 15.2-9 15.2.2.2 Analysis of Effects and Consequences 15.2-10 15.2.2.3 Results 15.2-12 15.2.2.4 Conclusions 15.2-15 15.2.3 Rod Cluster Control Assembly Misoperation 15.2-16 15.2.3.1 Identification of Causes and Accident Description 15.2-15 15.2.3.2 Analysis of Effects and Consequences 15.2-18 15.2.3.3 Results 15.2-19 15.2.3.4 Conclusions 15.2-21 1S.2.4 Uncontrolled Boron Dilution 15.2-23 15.2.4.1 Identification of Causes and Accident Description 15.2-23 15.2.4.2 Analysis of Effects and Consequences 15.2-24 15.2.4.3 Results and Conclusions 15.2;24 15.2.5 Partial Loss of Forced Reactor Coolant Flow 15.2-26 1S.2.5.1 Identification of Causes and Accident Description 15.2-26 15.2.5.2 Analysis of Effects and Consequences 15.2-27 15.2.5.3 Results 15.2-28 15.2.5.4 Conclusions 1S.2-28 15.2.6 Startup of an Inactive Reactor Coolant Loop 15.2-29 15.2.6.1 Identification of Causes and Accident Description 15.2-29 15.2.6.2 Analysis of Effects and Consequences 15.2-30 15.2.6.3 Results 15.2-31 15.2.6.4 Conclusions 15.2-31 15.2.7 Loss of External Electrical Load and/or Turbine Trip 15.2-32 15,2.7.1 Identification of Causes and Accident Description 15.2"32 15.2.7.2 Analysis of Effects and Consequences 15.2"33 15.2.7.3 Results 15.2-36 15.2.7.4 'Conclusions 15.2-37 15.2.8 Loss of Normal Feedwater 15.2-38 15.2.8.1 Identification of Causes and Accident Description 15.2-38 15.2.8.2 Analysis of Effects and Consequences 15.2-39 1575v:1D/122288 15.0-ii g 11 c'4 l 'ft )h, fi> TABLE OF CONTENTS (Cont) Section Descri tion ~Pe e 15.2.8.3 Results 15.2-40 15.2.8.4 Conclusions 15.2-41 15.2.9 Loss of Offsite Power to the Station Auxiliaries (Station Blackout) 15.2-42 15.2.9.1 Identification of Causes and Accident Description 15.2-42 15.2.9.2 Analysis of Effects and Consequences 15.2-43 15.2.9.3 Results 15.2.43 15.2.9.4 Conclusions 15.2-44 15.2.10 Excessive Heat Removal Due to Feedwater System Mal functions 15.2-45 15.2.10.1 Identification of Causes and Accident Description 15.2-45 15.2.10.2 Analysis of Effects and Consequences 15.2-45 15.2.10.3 Results 15.2-47 15.2.10.4 Conclusions 15.2-48 15.2.11 Excessive Load Increase Incident 15.2-49 15.2.11.1 Identification of Causes and Accident Description 15.2-49 15.2.11.2 Analysis of Effects and Consequences 15.2-49 15.2.11.3 Results 15.2-51 15.2.11.4 Conclusions 15.2-52 15.2.12 Accidental Depressurization of the Reactor Coolant System 15.2-53 15.2.12.1 Identification of Causes and Accident Description 15.2-53 '5,2.12.2 Analysis of Effects and Consequences 15.2-53 15.2.12.3 Results 15.2-54 15.2.12.4 Conclusions 15.2-54 15.2.13 Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.2-55 15.2.13.1 Identification of Causes and Accident Description 15.2-55 15.2.13.2 Analysis of Effects and Consequences 15.2-56 15.2.13.3 Results 15.2-58 15.2.13.4 Conclusions 15.2-59 ,15.2.14 Inadvertent Operation of the Emergency Core Cooling System at Power 15.2-60 15.2.14.1 Identification of Causes and Accident Description 15.2-60 15.2.14.2 Analysis of Effects and Consequences 15.2-62 15.2.14.3 Results 15.2-63 15.2.14.4 Conclusions 15.2-63 15.2.15 References 15.2-65 1575v:1 0/020389 15.0-iii Wt I ~ O' J TABLE OF CONTENTS (Cont} Section Descri tion ~Pe e

15. 3 CONDITION III - INFREQUENT FAULTS 15.3-1 15.3.2 Complete Loss of Forced Reactor Coolant Flow 15.3"2 15.3.2.1 Identification of Causes and Accident Description 15.3-2 15.3.2.2 Analysis of Effects and Consequences 15.3-3 15.3.2.3 Resul ts 15.3-4 15.3.2.4 Conclusions 15.3-4 15.3.3 Single Rod Cluster Control Assembly Mithdrawal at Full Power 15.3-5 15.3.3.1 Identification of Causes and Accident Description 15.3-5 15.3.3.2 Analysis of Effects and Consequences 15.3-6 15.3.3.3 Results 15.3-6 15.3.3.4 Conolusions 15.3-7 15.3.4 References 15.3-8 15.4 CONDITION IV - LIMITING FAULTS 15.4-1 15.4.2 Major Secondary System Pipe Rupture 15.4-2 15.4.2.1 Rupture of a Main Steam Line 15.4-2 15.4.2.2 Major Rupture of a Main Feedwater Pipe 15.4-10 15.4.4 Single Reactor Coolant Pump Locked Rotor 15. 4-16 15.4.4.1 Identification of Causes and Accident Description 15.4-16 15.4.4.2 Analysis of Effects and Consequences 15.4-16 15.4.4.3 Results 15.4-19 15.4.4.4 Conclusions 15.4-19 15.4.6 Rupture of a Control Rod Drive Mechanism Housing, (Rod Cluster Control Assembly Ejection) 15.4-21 15.4.6.1 Identification of Causes and Accident Description 15.4"21 15.4.6.2 Analysis of Effects and Consequences 15.4-24 15.4.6.3 Results 15.4-29 15.4.6.4 Conclusions 15.4-31 15.4.7 References 15.4-32 1575v:1D/122988 15. 0" i v

TABLE OF CONTENTS (Cont) TABLES Table Title 15.1-1 Nuclear Steam Supply System Power Rating 15.1-2 Trip Points and Time Delays to Trip Assumed in Accident. Analyses 15.1-3 Summary of Initial Conditions and Computer Codes Used 4 15.1-4 Equipment Available for Transient and Accident Conditions 15.2-1 Time Sequence of Events for Condition 11 Events 15.3"1 .Time Sequence of Events for Condition III Events 15.4-8 Time Sequence of Events for Major Secondary System Pipe Ruptures 15.4"9 Summary of Results for Locked Rotor Transients 15.4"10 Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident 15.4-11 Time Sequence of Events for Rod Cluster Control Assembly Ejection Accident 1575v:1D/020389 15.0-v k4 a, gl I TABLE OF CONTENTS (Cont) FIGURES ~Fi ure Title 15,1-1 Overtemperature and Overpower Delta-T Protection 15.1-2 Rod Position versus Time After Rod Drop Begins 15.1-'3 Normalized RCCA Reactivity Worth versus Rod Insertion 15.1-4 Normalized RCCA Reactivity Worth versus Time After Rod Drop Begins 15.1-5 Doppler Power Coefficient Used in Accident Analyses 15.1-6 Figure .Deleted 15.1-7 1979 ANS Decay Heat 15.1-8 Fuel Rod Cross Section 15.2.1-1 Uncontrolled Rod Withdrawal from A Subcritical Condition-Nuclear Power and Core Heat Flux Versus Time 15.2.1-2 Uncontrolled Rod Withdrawal from A Subcritical Condition - Hot Spot Fuel Average and Clad Temperature Versus Time 15.2.2"1 Uncontrolled Rod Withdrawal From 100% Power Terminated by High Neutron Flux Trip - Pressurizer Pressure and Nuclear Power Versus Time 15.2.2-2 Uncontrolled Rod Withdrawal From 100% Power Terminated by High Neutron Flux Trip - DNBR and T Versus Time-- avg 15.2.2-3 Uncontrolled Rod Withdrawal From 100% Power Terminated by Overtemperature Delta-T Trip - Pressurizer Pressure and Nuclear Power Versus Time 15.2.2-4 Uncontrolled Rod Withdrawal From 100% Power Terminated by Overtemperature Delta-T Trip - DNBR and T avg Versus Time 15.2.2-5 Effect of Reactivity Insertion Rate on Minimum DNBR For a Rod Withdrawal Accident at 100% Power 1575v:1o/020389 15.0-vi h II TABLE OF CONTENTS (Cont) FIGURES ~Fi ure, Title 15.2.2-6 Effect of Reactivity Insertion Rate on Minimum DNBR For a Rod Withdrawal Accident at 60% Power 15.2.2-7 Effect of Reactivity Insertion Rate on Minimum DNBR For a Rod Withdrawal Accident at 10% Power 15.2.3-1 Transient Response to A Dropped RCCA - Nuclear Power and Heat Flux'ersus Time 15.2.3-2 Transient Response to A Dropped RCCA - T and Pressurizer Pressure Versus Time avg 15.2.5-1 All Loops Operating, One Loop Coasting Down - Vessel Flow and Faulted Loop Flow Versus Time 15.2.5-2 All Loops Operating, One Loop Coasting Down - Nuclear Power and Heat Flux Versus Time 15.2.5-3 All Loops Operating, One Loop Coasting Down - Pressurizer Pressure and DNBR Versus Time 15.2.6-1 Startup of an Inactive Loop - Nuclear Power Versus Time 15.2.6-2 Startup of an Inactive Loop - Average and Hot Channel Heat Flux Versus Time 15.2.6-3 Startup of an Inactive Loop - Pressurizer Pressure and Core ersus Time 15.2.6-4 Startup of an Inactive Loop - Core Flow and DNBR Versus Time 15.2.7-1 Loss of Load With Pressurizer Spray and Power-operated Relief Valves at Beginning of Life - Nuclear Power and DNBR Versus Time 15.2.7-2 Loss of Load With Pressurizer Spray and Power-operated Relief Valves at Beginning of Life - Pressurizer Pressure and Water Volume Versus Time 15.2.7-3 Loss of Load With Pressurizer Spray and Power-operated Relief Valves at Beginning of Life - Core T and Steam Temperature Versus Time v avg 15.2.7-4 Loss of Load With Pressurizer Spray and Power-operated Relief Valves at End of Life - Nuclear Power and DNBR Versus Time 15.2.7-5 Loss of Load With Pressurizer Spray and Power-operated Relief Valves at End of Life - Pressurizer Pressure and Water Volume Versus Time 1 575 v:1D/020389 15.0-vii TABLE OF CONTENTS (Cont) FIGURES ~Fi ure Ti tie 15.2.7-6 Loss of Load With Pressurizer Spray and Power-operated Relief Valves at End of Life - Core T and Steam Temperature Versus Time avg 15.2.7-7 Loss of Load Without Pressurizer Spray and Power-op'crated Relief Valves at Beginning of Life - Nuclear Power and DNBR Versus Time 15.2.7-8 Loss of Load Without Pressurizer Spray and Power-operated Relief Valves at Beginning of Life - Pressurizer Pressure and Water Volume Versus Time 15.2.7-9 Loss of Load Without Pressurizer Spray and Power-operated Relief Valves at Beginning of Life Core T and Steam Temperature Versus Time avg 15.2.7-10 Loss of Load Without Pressurizer Spray and Power-operated Relief Valves at End of Life - Nuclear Power and DNBR Versus Time'5.2.7-11 Loss of Load Without Pressurizer Spray and Power-operated Relief Valves at End of Life - Pressurizer Pressure and Mater Volume Versus Time 15.2.7-12 Loss of Load Without Pressurizer Spray and Power-operated Relief Valves at End of Life - Core T v avg and Steam Temperature Versus Time 15.2.8-1 Loss of Normal Feedwater - Nuclear Power and Core Heat Flux Versus Time 15.2.8-2 Loss of Normal Feedwater - Primary Temperature and Steam Generator Pressure Versus Time 15.2.8-3 Loss of Normal Feedwater - Pressurizer Pressure and Mater Volume Versus'Time 15.2.9-1 Station Blackout - Nuclear Power and Core Heat Flux Versus Time 15.2.9-2 Station Blackout - Primary Temperature and Steam Generator Pressure Versus Time 15.2.9-3 Station Blackout - Pressurizer Pressure and Mater Volume Versus Time 15.2.10-1 Feedwater System Halfunction with Rod Control - Nuclear Power and Core Heat Flux Versus Time 1575v:1D/120788 15.0-viii i \ ~ A I 'I F A 4. E TABLE OF CONTENTS (Cont) FIGURES F i<iure Ti tl e 15.2.10-2 Feedwater System Malfunction with Rod .Control - Pressurizer Pressure and DNBR Versus Time 15.2.10-3 Feedwater System Malfunction with Rod Control - Loop Delta-T and Core Tavg Versus Time 15.2.10-4 Feedwater System Malfunction without Rod Control - Nuclear Power and Core Heat Flux Versus Time 15.2.10-5 Feedwater System Malfunction without Rod Control - Pressurizer Pressure and DNBR Versus Time 15.2.10-6 Feedwater System Malfunction withw t: Rod Control - Loop Delta-T and Core Tavg Versus Time 15.2.11-1 Excessive Load Increase Without Control, Miniimum Feedback-Nuclear Power and Pressurizer Pressure Versus Time 15.2.11-2 Excessive Load Increase Without Control, Minimum Feedback-T av r i a n d D N B R Ve s u s T m e 15.2.11-3 Excessive Load Increase Without Control, Maximum Feedback-Nuclear Power and Pressurizer Pressure Versus Time'5.2.11-4 Excessive Load Increase Without Control, Maximum Feedback-T av r a n d D N B R Ve s u s 15.2.11-5 Excessive Load Increase With Control, Minimum Feedback - Nuclear Power and Pressurizer Pressure Versus Time 15.2.11-6 Excessive Load Increase With Control, Minimum Feedback - Tav avg and DNBR Versus Time 15.2.11-7 Excessive Load 'Increase With Control, Maximum Feedback - Nuclear Power and Pressurizer Pressure Versus Time 15.2.11-8 Excessive Load Increase With Control, Haximum Feedback - T and DNBR Versus Time avg 15.2.12-1 Accidental Depressurization of the Reactor Coolant System-Nuclear Power and Core T Versus Time r avg 15.2.12-2 Accidental Depressurization of the Reactor Coolant System-Pressurizer Pressure and Water Volume Versus Time 15.2. 12-3 Accidental Depressurization of the Reactor Coolant System - DNBR Versus Time 1575v:1O/120788 15.0-ix TABLE OF CONTENTS (Cont) FIGURES ~Fi ure Ti tie 15.2.13-1 Main Steam Depressurization - Variation of Temperature Keff with Core 15.2.13-2 Main Steam Depressurization - Safety Injection Flowrate 15.2.13-3 Transient Response For A Steam Line Break Equivalent to 268 lb/sec at 1200 psia With Offsite Power Available 15.2.13-4 Transient Response For A Steam Line Break Equivalent to 268 lb/sec at 1200 psia With Offsite Power Available 15.2.14-1 Spurious Actuation of the Safety Injection System - Nuclear Power, Steam Flow and Core T Versus Time avg 15.2.14-2 Spurious Actuation of the Safety Injection System - Pressurizer Water Volume, Pressurizer Pressure and DNBR Versus Time 15.3.2-1 All Loops Operating, All Loops Coasting Down - Vessel Flow and Heat Flux Versus Time 15.3.2-2 All Loops Operating, All Loops Coasting Down - Nuclear Power and DNBR Versus Time 15.3.2-3 All Loops Operating, All Loops Coasting Down - Pressurizer Pressure Versus Time 15.4.2-1 Variation of Reactivity with Power at Constant Core Average Temperature 15.4.2-2 Transient Response to a Steam Line Break Double Ended Rupture with Offsite 'Power Available (Case A) 15.4.2-3 Transient Response to a Steam Line Break Double Ended Rupture with Offsite Power Available (Case A) 15.4.2-3A Transient Response to a Steam Line Break Double Ended Rupture with Offsite Power Available (Case A) 15.4.2-4 Transient Response to a Steam Line Break Double Ended Rupture with No Offsite Power Available (Case B) 15.4.2-5 Transient Response to a Steam Line Break Double Ended Rupture with No Offsite Power Available (Case B) 15.4.2-5A Transient Response to a Steam Line Break Double Ended Rupture with No Offsite Power Available (Case B) 1575v:1 0/020389 15. 0-'x 4 E TABLE OF CONTENTS (Cont) FIGURES F iciure Ti tie 15.4.2-6 Main Feedline Rupture with Offsite Power -,Nuclear Power and Core Heat Flux Versus Time 15.4.2-7 Main Feedline Rupture with Offsite Power - Pressurizer Pressure and Water Volume Versus Time 15.4.2-8 Main Feedline Rupture with Offsite Power - Faulted and Intact Loop Coolant Temperatures Versus Time 15.4.2-9 Main Feedline Rupture with Offsite Power - Steam Generator Pressure and Water Mass Versus Time 15.4.2-10 Main Feedline Rupture without Offsite Power - Nuclear Power and Core Heat Flux Versus Time 15.4.2-11 Main Feedline Rupture without Offsite Power - Pressurizer Pressure and Water Volume Versus Time 15.4.2-12 Main Feedline Rupture without Offsite Power - Faulted, Isolated, and Intact Loop Temperatures Versus Time 15.4.2-13 Main Feedline Rupture without Offsite Power - Steam Generator Pressure and Water Mass Versus Time '15.4.4-1 All Loops Operating, One Locked Rotor - RCS Pressure Versus Time 15.4.4-2 All Loops Operating, One Locked Rotor - Total RCS Flow Versus Time 15.4.4-3 All Loops Operating One Locked Rotor - Faulted Loop Flow Versus Time 15.4.4-4 All Loops Operating One Locked Rotor - Nuclear Power Versus Time 15.4.4-5 All Loops Operating One Locked Rotor - Maximum Clad Temperature Versus Time 15.4.4-6 All Loops Operating One Locked Rotor - Hot Channel Heat Flux Versus Time 15.4.6-1 Rod Ejection Accident, BOL HFP - Nuclear Power, Hot Spot Fuel and Clad Temperature Versus Time 15.4.6-2 Rod Ejection Accident, BOL HZP - Nuclear Power, Hot Spot Fuel and Clad Temperature Versus Time 1575v:10/020389 15.0-xi 4, 4, r ,I, c Chapter 15 ACCIDENT ANALYSES Since 1970, the ANS classification of plant conditions has been used to divide plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to the public. The four categories are as follows: (1) Condition I: Normal Operation and Operational Transients (2) Condition II: Faults of Moderate Frequency (3) Condition III: Infrequent Faults (4) Condition IV: Limiting Faults. The basic principle applied in relating design requirements to each of the conditions is that the most frequent occurrences must yield little or no radiological risk to the public, and those extreme situations having the potential for the greatest risk to the public shall be those least likely to occur. Where applicable, reactor trip, system and engineered safety features functioning is assumed, to the extent allowed by considerations such as the single failure criterion, in fulfilling this principle. 1575v:1D/110888 'J ~r . 15.1 CONDITION I - NORMAL OPERATION AND OPER ..'IONAL TRANSIENTS Condition I occurrences are those that are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant. As such, Condition I occurrences are accommodated with margin between any plant parameter and the value of that parameter which would require either automatic or manual protective action. Inasmuch as Condition I occurrences occur frequently or regularly, they must be considered from the point of view of affecting, the consequences of fault conditions (Conditions II, III and IV). In this regard, analysis of each fault condition is generally based on a conservative set of initial conditions corresponding to the most 'adverse set of conditions that can occur during Condition I operation. A typical list of Condition I events is shown below: (1) Steady state and shutdown operations Mode 1 - Power operation (> 5% of rated thermal power) Mode 2 - Startup > 0.99, < 5% of rated thermal power) (Keff Mode' - Hot standby (K ' 0.99, > 350'F) ff eff < T avg Mode 4 - Hot shutdown (subcritical, residual heat removal system isolated, Keffff < 0.99, ' 200'F < T < 350'F) avg Mode 5 - Cold shutdown (subcritical, residual heat removal system in operation, Keff 0.99, T < 200'F) - Refueling (K 0.95, 140'F) Mode 6 ff < T < 1575v:lo/110888 ~ 1 V F (2) Operation wi <h permi ssibl e devi ati ons Various deviations that may'ccur during continued operation as permitted by the plant Technical Specifications (Reference 1) must be considered in conjunction with other operational modes. These include: (a) Operation with components or systems out of service (b) Leakage from fuel with cladding defects (c) Activity in the reactor coolant

1. Fission products

2. Corrosion products

3. Tritium (d) Operation with steam generator leaks up to the maximum allowed by the Technical Specifications (e) Testing as allowed by the Technical Specifications (3) Operational transients (a) Plant heatup and cooldown (up to 100'F/hour for the reactor coolant system (RCS); 200'F/hour for the pressurizer during cooldown, and 100'F/hour for the pressurizer during heatup).

(b) Step load changes (up to +10%) (c) Ramp load changes (up to 5% per minute) (d) Load rejection up to and including design load rejection transient 1575v:1D/111188 g j 1 y' ~ 15.1.1 0 timization of Control S stems A setpoint study (Reference 2) is performed in order to simulate performance of the Reactor Control and Protection Systems. Emphasis was placed on the development of a control system that will automatically maintain prescribed conditions in the plant even under the most conservative set of reactivity parameters with respect to both system stability and transient performance. For each mode of plant operation, a group of optimum controller setpoints is determined. In areas where the resultant setpoints are different, compromises based on the optimum overall performance are made and verified. A consistent set of control system parameters is derived satisfying plant operational requirements throughout the core life and for various power levels. The study comprises an analysis of the following control systems: rod cluster control assembly, steam dump, steam generator level, pressurizer pressure, and pressurizer level.

15. 1.2 Initial Power Conditions Assumed in Accident Anal ses Reactor power-related initial conditions assumed in the accident analyses presented in this chapter are described in this section.

15.1.2.1 Power Ratin Table 15. 1-1 lists the principal power rating values that are assumed in analyses performed in this section. Two ratings are given: ( 1) The guaranteed nuclear steam supply system (NSSS) thermal power output. This power output includes the tnermal power generated by the reactor coolant pumps. (2) The engineered safety features (ESF) design rating. The Westinghouse-supplied ESFs are designed for a thermal power higher than the guaranteed value in order not to preclude realization of future potential power capabilty. This higher thermal power value is designated as the ESF design rating. This power output includes the thermal power generated by the reactor coolant pumps. 1575v:10/111188 15.1-3 iver Mhere initial power operating conditions are assumed in accident analyses, the guaranteed NSSS thermal power output (plus allowance for errors in steady state power determination for some accidents) is assumed. Where demonstration of the adequacy of the containment and ESF is concerned, the ESF design rating plus allowance for error is assumed. The thermal power values for each transient analyzed are given in Table 15.1-3.

15. 1.2.2 Initial Conditions For most accidents which are DNH limited, nominal values of i'nitial conditions are assumed. The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the limit DNBR, as described in Reference 3. This procedure is known as the "Improved Thermal Design Procedure" ( ITDP) and these accidents utilize the MRB-1 and MRB-2 DNB correlations (References 4 and 5). The initial conditions for other key parameters are selected in such a manner to maximize the impact on DNBR.

Minimum measured flow is used in all ITDP transients. For accident evaluations that are not DNH-limited, or for which the Improved Thermal Design Procedure is not employed, the initial conditions are obtained by adding maximum steady state er'rors to rated values. The following steady state errors are considered: (1) Core power +2.0% allowance for calorimetric error (2) Average RCS temperature +5.3/-6.8'F allowance for. controller deadband and,measurement error (3) Pressurizer pressure +38 psi/-50 psi allowance for steady state fluctuations and measurement error. 1575v:1D/020389 yyl <I

15. 1.2.3 Power Distribution The transient response of the reactor system is dependent on the initial power distribution. The nuclear design of the reactor core minimizes adverse power distribution through the placement of fuel assemblies, control rods, and by operation instructions. The power distribution may be characterized by the radial peaking factor F H and the total peaking factor F<. The peaking factor limits are given in the Technical Specifications.

For transients that may be DNB-limited, the radial peaking factor is of importance. The radial peaking factor increases with decreasing power level due to rod insertion. This increase in F<H is included in the core limits illustrated on Figure 15. 1-1. All transients that may be DNB-limited are assumed to begin with an F H consistent with the initial power level defined in the Technical Specifications. The axial power shape used in the DNB calculation is discussed in FSAR Section 4.4. For transients that may be overpower-limited, the total peaking factor F is of importance. All transients that may be overpower-limited are assumed to begin with a value of F< consistent with the initial power level. as defined in the Technical Specifications. For overpower transients which are slow with respect to the fuel rod thermal time constant (for example, CVCS Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant Inventory), the fuel rod thermal evaluations are performed as discussed in Section 4.4 of the FSAR. For overpower transients which are fast with respect to the fuel rod thermal time constant (for example, RCCA Ejection events), a detailed fuel heat transfer calculation is performed. 15.1.3 Tri Points and Time Dela s to Tri Assumed in Accident Anal ses A reactor trip signal acts to open two trip breakers connected in series I feeding power to the control rod drive mechanisms. The loss of power to the 1575v:10/110888 15.1-5 ttC> &j

4 1e

mechanism coils causes the mechanism to release the rod cluster control assemblies (RCCAs) which then fall by gravity into the core. There are various instrumentation delays associated with each trip function, including delays in signal actuation, in opening the trip breakers, and in the release of the rods by the mechanisms. The total delay to trip is defined as the time delay from the time that trip conditions are reached to the time the rods are free and begin to fall. The difference between the limiting trip. point assumed for the analysis and the nominal trip point represents an allowance 'for instrumentation charm'el error and setpoint error. Nominal trip setpoints are specified in the plant V Technical Specifications. During startup tests, it is demonstrated that actual instrument time delays are equal to or less than the assumed values. In addition, protection system channels are calibrated and instrument response times are determined in accordance with the plant Technical Specifications.

15. 1.4 Rod Cluster Control Assembl Insertion Characteristic The negative reactivity insertion following a reactor trip is a function of the acceleration of the RCCA and the variation in rod worth as a function of rod position.

With respect to accident analyses, the critical parameter is the time of up to the dashpot entry or approximately 85% of the rod cluster 'nsertion travel. For accident analyses, the insertion time to dashpot entry is conservatively taken as 2.7 seconds. The RCCA position versus time assumed in accident analyses is shown on Figure 15. 1-2. Figure 15.1-3 shows the fraction of total negative reactivity insertion for a core where the axial distribution is skewed to the lower region of the core. This curve is used as input to all point kinetics core models used in transient analyses. There is inherent conservatism in the use of this curve in that it is based on a skewed axial power distribution that would exist relatively infrequently. For cases other than those associated with xenon oscillations, significant negative reactivity would have been inserted due to the more favorable axial power distribution existing prior to trip. 1575v:1D/110888 15.1-6 l g' ~% The normalized RCCA negative reactivity insertion versus time is shown on Figure 15.1-4. The curve shown in this figure was obtained from Fic:ures

15. 1-2 and 15. 1-3. A total negative reactivity insertion following a trip of 4.8% Lk is assumed in the transient analyses except where specifically noted otherwise. This assumption is conservative with respect to the calculated trip reactivity worth available as shown in FSAR Tab/e 4.3.3-4.

The normalized RCCA negative reactivity insertion versus time curve for an axial power distribution skewed to the bottom (Figure 15. 1-4) is used in transient analyses. Where special analyses require the use of three-dimensional or axial one-dimensional core models, the negative reactivity insertion resulting from reactor trip is calculated directly by the reactor kinetic code and is not separable from other reactivity feedback effects. In this case, the RCCA position versus time of Figure 15.1-2 is used as a code input.

15. 1.5 Reactivit Coefficients The transient response of the reactor coolant system is dependent on reactivity feedback effects, in particular the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients and their values are discussed in detail in FSAR Chapter 4. In the analysis of certain events, conservatism requires the use of large reactivity coefficient values, whereas in the analysis of other events, conservatism requires the use of small reactivity coefficient values. Some analyses, such as loss of reactor coolant from cracks or ruptures in the RCS, do not depend on reactivity feedback effects. The values used are given in Table. 15. 1-3; reference is made in that table to Figure 15. 1-5 that shows the upper and lower Doppler power coefficients, as a function of power, used in the transient analysis. The justification for use of conservatively large versus small reactivity coefficient values is treated on an event-by-event basis.

1575v:10/110888 1 1 '~ II +I e. "$ ~ ~ 4 15.1.6 Plant S stems and Com onents Available for Miti ation of Accident Effects The NSSS is protected by design from the possible effects of natural phenomena, postulated environmental conditions and dynamic effects of the postulated accidents. In addition, the design incorporates features which minimize the probability and effects of fires and explosions. The incorporation of these features in the NSSS, coupled with the reliability of the design, ensures that the normally operating systems and components listed in Table 15.1-4 will be available for mitigation of the events discussed in Chapter 15. In determining which systems are necessary to mitigate the effects of these postulated events, the classification system of ANSI-N18.2-1973 is utilized. The design of "systems important to safety" (including protection systems) is consistent with IEEE Standard 379-1972 and Regulatory Guide 1.53 in the application of the single failure criterion. In the analysis of the Chapter 15 events, control system action is considered only if that action results in more severe accident results. No credit is taken for control system operation if that operation mitigates the results of an accident. For some accidents, the analysis is performed both with and without control system operation to determine the worst case.

15. 1.7 Fission Product Inventories The fission product inventories existing in the core and fuel rod gaps are described in Section 15.0.9 of the FSAR.

15. 1.8 Residual Oeca Heat For the non-LOCA analyses the 1979 ANS decay heat curve is used (Reference 6). Figure 15.1-7 presents this curve as a function of time after shutdown.

15. 1.9 Com uter Codes Utilized Summaries of some of the principal computer codes used in transient analyses are given below. Other codes, in particular, very specialized codes in which the modeling has been developed to simulate one given accident, such as the 1575v:10/111188

f g a %E'l 4 ~4 '+I ,4 ~ SATAN-V1 code used in the analysis of the RCS pipe rupture (Section 15.4), and which consequently have a direct bearing on the analysis of the accident itself, are summarized in their respective accident analyses sections. The codes used in the analyses of each transient are listed in Table 15. 1-4.

15. 1. 9. 1 FACTRAN FACTRAN calculates the transient temperature distribution in a cross section of a metal clad U02 fuel rod (see Figure 15. 1-8) and the transient heat flux at the surface of the clad using as input the nuclear power and the time-dependent coolant parameters (pressure, flow, temperature and density).

The code uses a fuel model that exhibits the following features simultaneously: (1) A sufficiently large number of finite difference radial space increments to handle fast transients such as rod ejection acc'idents (2) Material properties that are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation (3) The necessary calculations to handle post-DNB transients: film boi ling heat transfer correlations, zircaloy-water reaction and partial melting of the materials. FACTRAN is further discussed in Reference 7. 15.1.9.2 LOFTRAN The LOFTRAN program is used for .studies of transient response of a PHR system to specified perturbations in process parameters. LOFTRAN simulates a multi loop system by modeling the reactor core and vessel, hot and cold leg piping, steam generator (tube and shell-sides), reactor coolant pumps and the pressurizer with up to four reactor coolant loops. The pressurizer heaters, spt.ay, relief and safety valves are also considered in the program. Point model neutron kinetics and reactivity effe'cts of the moderator, fuel, boron, and rods are included. The secondary side of the steam generator utilizes a homogeneous, saturated mixture for the thermal transients and a water level 1575v:1D/120788 'II ~: S I I rg V '~J correlation for indication and control. The reactor protection system is simulated to include reactor trips on neutron flux, overpower and overtemperature reactor coolant hT, high and low pressure, low flow, and high pressurizer level. Control systems are also simulated including rod control, steam dump, feedwater control, and pressurizer pressure control. The safety injection system (SIS), including the accumulators, is also modeled. LOFTRAN is a versatile program that is suited to both accident evaluation and control studies as well as parameter sizing. LOFTRAN also has the capability of calculating the transient value of DN8R based on the input from the core limits illustrated on Figure 15. 1-1. The core limits represent the minimum value of DN8R as calculated for a typical or thimble cell. LOFTRAN is further discussed in Reference 8.

15. 1.9.3 LEOPARD The LEOPARD computer program determines fast and thermal spectra using only basic geometry and temperature data. The code optionally computes fuel depletion effects for a dimensionless reactor and recomputes the spectra before each discrete burnup step.

LEOPARD is further discussed in Reference 9.

15. 1.9. 4 TURTLE TURTI.E is a two-group, two-dimensional neutron diffusion code featuring direct treatment of the nonlinear effects of xenon, enthalpy, and Doppler feedback:

Fuel depletion is allowed. TURTI.E was written for the study of azimuthal xenon osci llations, but the code is useful for general analysis. The input is simple, fuel management is handled directly, and a boron criticality search is allowed. TURTLE is further described in Reference 10. 1575v:1D/120788 15.1-10 'I E t1 II <1 ~i 4 1>>, 15.1.9. 5 TWINKLE The TWINKLE program is a multidimensional spatial neutron kinetics code. which was patterned after steady state codes presently used for reactor core design. The code uses an implicit finite-difference method to solve the two-group transient neutron diffusion equations in one, two, and three dimensions. The code uses six delayed neutron groups and contains a detailed multiregion fuel-clad-coolant heat transfer model for calculating pointwise Doppler and moderator feedback effects. The code handles up to 2000 spatial points and performs its own steady state initialization. Aside from basic cross sect'ion data and thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits provide channelwise power, axial offset, enthalpy, volumetric surge, pointwise power, fuel temperatures, and so on. The TWINKLE code is used to predict the kinetic behavior of a reactor for transients that cause a major perturbation in the spatial neutron flux distribution. TWINKLE is further described in Reference 11. 15.1.9. 6 TH INC The THING code is described in Section 4.4 of the FSAR. 15 F 1.10 REFERENCES

1. Technical S ecifications, Shearon Harris Nuclear Power Plant Unit One Technical Specifications through Amendment 7, May 22, 1988.

2. S. Abedin, Set oint Stud Carolina Power and Li ht Com an Shearon Harris Nuclear Plant, WCAP-10183, November 1982.

1575v:1D/120788 15.1-11 6

3. Chelmer, H., et al;, "Itiiproved thermal Design Procedure," WCAP-8567 (Proprietary) and WCAP-8568 (Non-Proprietary), July 1975.

4. Motleg, F. E., et al., "New Westinghouse Correlations WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids,"

WCAP-8762-P=A and WCAP-8763-A, July 1984.

5. Davidson, S. L. and Kramer, W. R.; (ed.) "Reference Core Report Vantage 5 Fuel Assembly," Appendix A.2.0, September 1985.

6. ANSI/ANS-5. 1-1979, "Decay Heat Power In Light Water Reactors",

August 29, 1979.

7. H. G. Hargrove, FACTRAN - A Fortran IV Code for Thermal Transients in a U02 Fuel Rod, WCAP-7908, June 1972.

8. T. W. T. Burnett et al, LOFTRAN Code Descri tion, WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.

9. R. F. Barry, LEOPARD - A S ectrum De endent Non-S atial De letion Code for the IBM-7904, WCAP-3269-26, September 1963.

10. R. F. Barry and S. Altomare, The TURTLE 24.0 Diffusion De letion Code, WCAP-7213-P-A (Proprietary), WCAP-7758-A (Non-Proprietary), January 1975.

11. D. H. Risher, Jr. and R. F. Barry, TWINKLE - A Multi-Dimensional Neutron Kinetics Com uter Code, WCAP-7979-P-A (Proprietary), WCAP-8028-A (Non-Proprietary), January 1975.

1575v:10/111) 88 15.1-12 TABLE 15. 1-1 NUCLEAR STEAM SUPPLY SYSTEM POWER RATING Core thermal power (license level) 2775 Thermal power generated by the reactor coolant pumps 10 Nuclear steam supply system thermal power output 2785 Engineered safety features design rating Ig~ximum calculated turbine rating) 2910 (a) The unit will not license rating. be operated at this rating because it exceeds the 1575v:1D/111588 TABLE 15.1-2 Sheet 1 of 2 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Limiting Trip Trip Point Assumed Time Delay, Function In Anal ses sec Power'range high neutron flux, high setting 118% 0..5 Power range high neutron flux, low setting 35% 0.5 High Neutron Flux, P-8 79% 0.5 Overtemperature aT Variable, see 60 Figure 15.1-1 Overpower aT Variable, see Figure 15.1-1 High pressurizer pressure 2445 psig Low pressurizer pressure 1920 psig Low reactor coolant flow (from loop flow detectors) 87% loop flow Undervoltage trip (b) 1.5 15 7 5 v:1 o/11158 & lt, yl TABLE 15.1-2 Sheet 2 of 2 Limiting Trip Tl lp Point Assumed Time Delay, Function A sec Turbine trip Not applicable Low-low steam generator level 32% of narrow 3.5 range level span 19% of narrow 3.5 range level span High-high steam generator level trip 90.2% of narrow 2.5 (for of the feedwater pumps and turbine; range level span turbine trip) closure of feedwater system 10* (for valves* feedwater isolation) (a) Total time delay (including RTD and thermowell time response, trip circuit and channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to fall. ~ (b) A specific undervoltage setpoint was not assumed in the safety analysis. (c) Loss of Normal Feedwater/Station Blackout Analyses (d) Feedline Brea'k Analysis 1575v:1D/020889 Sheet 1 of 4 TABLE 15.1-3

SUMMARY

OF INITIAI CONDITIONS AND COMPUTER CODES USED Assumed Reactivity Coefficients Initial NSSS

'Moderator Moderator Thermal Power Output Computer Density Assumed Faults Codes Utilized 13k/gm/cc Doppler MWt CONDITION II Uncontrolled RCCA bank TWINKLE, See Section withdrawal from a subcritical FACTRAN, THINC 15.2.1.2 condition Uncontrolled RCCA bank LOFTRAN 0.50 Lower and 279/1674/2790 withdrawal at power Upper RCCA misoperation THINC, (e) (e) (e) 2785 LOFTRAN Uncontrolled boron dilution 0 and 2785 Partial loss of forced reactor LOFTRAN Upper 2785 coolant flow FACTRAN, THINC Startup of an inactive reactor LOFTRAN, 0.50 Lower 1671 coolant loop FACTRAN, THINC Loss of external electrical load LOFTRAN 0.50 Lower and 2785 and/or turbine trip Upper Loss of normal feedwater LOFTRAN Upper 2910 Loss of offsite power to the LOFTRAN Upper 2910 plant auxiliaries (station blackout) 1575v:10/120188

Sheet 2 of 4 TADL'E 15.1-3 Assumed Reactivity Coefficients Initial NSSS Moderator Moderator Thermal Power a ~ a Output Computer Temp. Densi ty Assumed Faul ts Codes Utilized pcm/ F hk/gm/cc Doppler MHt CONDITION I I (Con t ')

Excessive heat removal due to LOFTRAN 0.50 Lower 0 and 2785 feedwater system malfunctions Excessive load increase LOFTRAN 0 and 0.50 Lower and 2785 Upper Accidental depressurization of LOFTRAN Lower 2785 the reactor coolant system Accidental depressurization of LOFTRAN Function of See Figure 0 the main steam system the modera- 15.4.2-1 (Subcritical) tor density.

See Sec. 15.2.13 (Figure 15.2.13-1)

Inadvertent operation of the ECCS LOFTRAN Lower 2785 at power CONDITION III Loss of reactor coolant from small NOTRUMP 2775(')

ruplured pipes or from cracks in SBLOCTA large pipe which actuate emergency core cooling I 57,"iv" I n/ I? n I 88

Sheet 3 of 4 TABLE 15. 1-3 Assumed Reaclivily Coefficients Initial NSSS Mndera Lor Modera t.or Thermal Power a Output Computer Temp. l1eesi~ty Assumed(

Faults Codes Utilized pcm/'F hk/gm/cc Doppler( Mtlt CONDITION III (Cont'd)

Complete loss of force reactor LOFTRAN, Upper 2785 coolant flow 'FACTRAN, THINC Single RCCA withdrawal at TURTLE, THINC, 2785 full power LEOPARD CONDITION I V Major rupture of pipes containing SATAN-VI Function of Function 2775(')

reactor coolant up to and including COCO moderator of fuel double-ended rupture of the largest BASH density. temp. See pipe in the reactor coolant system HREFLOOD See Sec. Sec. 15.4.1 (loss-of-coolant accident) LOCBART 15.4.1 Ha.jnr secondary system pipe rupture LOFTRAN Function of -

See Figure 0 up to and including double-ended . the Modera- 15'.4.2-1 (Subcri tical) rupture (rupture of a steam pipe) tor Density see Section 15.2.13 (Figure 15.2.13-1) 157sv'10/120188

Sheet 4 of 4 TABLE 15.1-3

. Assumed React i vi ty Coef fi ci ents Initial NSSS Moderator Moderator Thermal Power Output Computer Assumed Faults Codes Utilized pcm/'F hk/gm/cc Doppler MMt CONDITION IV (Cont'd)

Major secondary system pipe rupture LOFTRAN 0.50 Upper 2910 up to and including double-ended rupture (rupture of a feedline)

Single reactor coolant pump locked LOFTRAN Upper 2785 rotor FACTRAN, THING Rupture of a control rod mechanism TMINKLE, +5.2 BOL Consistent 0 and 2775(

housing (RCCA ejection) FACTRAN, -23. EOL with lower LEOPARD limit on Fig. 15.1-5 (a) Only one is used in analysis, i.e., either moderator temperature or moderator density coefficient.

(b) Reference Figure 15.1-5.

(c) Appropriate calorimetric error considered where applicable.

(d) Pcm means percent mille. See footnote Table 4.3-1.

(e) Provided in Reference 9 (Section 15.2.15)

(f) Core power.

1575v:10/020389

Sheet I oF 3 TABLE 15.1-4 i

EQUIPNCNT AVAILABLE FOR IRANSIENI AND ACCIDENI CONDITIONS Incident Reactor rrl Cunrtions ESf'ctuation f'unct Irma Other E ui ment ~ESP I <in high Bank Withdraval from Flux Flow s.p. I. source Subcritical range hi qh f lux. hi gh flux rate. manual Uncontrolled RCCA Power range high flux Pressurizer safety valves, Bank Withdraval at order. oidr. stram generator saFety Power hi prcssurizlr v<<11 ves pressure. hi pressurizer level, itmnuar

3. RCCA Power range negative Ni sa1 I gnment flux rate. Order.
manual
4. Uncontrolled Source range high flux. Lnv insertion limit Boron Dilution power range high flux, annunciators for Older. manual boration. source range count rate (while shutdown)
5. Startup of an Pownr range high flux - PB.
Inactive Reactor manual Coolant Loop
6. Loss of External High pressurizer pressure Pressurizer safety valves, Electrical Load and/ Ore. steam steam generator safety or Turbine Trip generator lo-lo level, v<<ilvnS Itlanual
7. Loss of Normal St> am gene> atnr ln-lo Stnam gnncrator lo-lo Stnam grnerator Auxiliary feedwater Fcodvatcr I evnl . manua I I I'.ve I safety valves SyS'tern B. Loss of Offsite Same as 7 S<<llllc aS 7 Simr as 7 Salllc as 7 Powel to the Statinn Auxiliaries
9. Excess Heat Re- Pover range high flux, High-hiqh Stnam gener- Fecdvater ISOlatiOn moval due to Feed- high-high steam generator .itor level producrd valves vater System Nal- level. manual f>iodwater isolition functions and turbine trip
10. Exressive Load Power range high flux. Pressurizer safety valves lncrc<<ise Incident orhr. Ophr steam qrnorator safety Inw prl Ssurir> r pr.l.nnurn. valvnn IILIAI>;II II. Accidental Orpres- Prrssuriror lnv suriratinn of thr RCSI proos urI.. 0151. minuil
~
1575v ID/121588
Sheet 2 of 3 TABI,F. $ 5. I-0 (CnEEt'd)
EOUIPMENT AVAILABLE FOR 1RANSIENI ANO ACCIDENT CONDITIONS l <Id t Reactor Tri functions ESF hctuaiion functions ~ESPE l ct
$ 2. Inadvertent Operation Pressurizer low of the ECCS at Power pressurizer. safety injection trip. manual
13. Major Rupture of Main SIS, low pressurizer Low pressurizer I end 1 i ne I SO 1 a t I nn va I Ved. Emergency feed-Steam Line pressure. OTBT. pressure. low comp- steam line isolation water system. SI power range high flux. ensated steam line valves equipment minus manual pressure. hi-I con- either one SI tainment pressure. charging pump. or manual one diesel generator.
$ 4. Complete Loss of Low flow. undervoltaqe.
Forced Reactor underfrequency. manual Coolant Flow
$ 5. Locked Rotor Low Flow. manual
$ 6. Rupture of a Control Power range high flux.
Rod Drive Mechanism high positive flux Housing rate. manual
$ 7. Single RCCA Ifith- OTBT. manual drawal at Full Power IB. Major Rupture of a Lo steam generator level High Containment Steam I ine Isolat IOI> Emergency feod-Main Feedwater Line plus steam/feed mismatch. pressure. steam valvula. feed line isolation water pumps SIS. high pressurizer generator low-low pr essuE izer self-actuated pressure. OTAT water level. low safety valves. steam gen-low-low steam generator compensated steam line erator safety va)ver level. manual pressure
19. Large Break LOCA ReactoE trip system Engineered safety Service water system. Emergency core f<.atures actuation co<<<pone nt cool ing rooling system.
sys'Le<El water system containment heat removal system.
emergency power system 20 Small Break LOCA Reactor trip system Engineered safety Service water system. Emergency core f<.atures actualion component cooiing cooling system.
system water system, generator auxiliary feedwater safety and/or relief valves system containment heat removal system emergency power sy tom 1575v 10/120788
Sheet 9 of 3
'IABI.E 15. I-4 (Cont'd)
EQUIPMENT AVAILABLE FOR TRANSIENT ANO ACCIOENT CONDITIONS Incident Reactor Tri Functions ESF Actuation Functions Other E ui ment ESF E ui ment
21. Steam Generator Reactor trip system Engineered safety Service vater system, Emergency core Tube Rupture features actuation component cooling vater cooling system.
system system. steam generator emergency I'eed-shell s1de fluid operating vater system.
system, steam generator emergency power safety and/or relief valve, systems steam line isolation valves 1575v;10/ I 20788
75 OVERPOWER dT TRIP LINE 65 2000 PSIA
- 2250 PSIA 1935 PSIA
-- -2400 PSIA 55 CORE LIMITS OVERTEHPERATURE dT TRIP LINES 45 LOCUS OF POINTS WHERE SG SAFETY VALVES OPEN zs 8
588 5EI5 5q8 SIS 688 685 618 6ls 628 625 6""8 655 T-avg ('F)
Shearon Harris Figure 15,1-1 Overtemperature and Overpower Delta-T Protection
0.9 0.8 I
o
?:
Cl 0.6 o.s C) 04 0.3 aC) 0.2 0.1 DASHPOT TINE AFTER ROD DROP BEGINS (SECONDS)
Shearon Harris Figure 15.1-2 Rod Position vs.
Time After Rod Drop Begins
09 I
0.8 I
0.7 0.6 0.5 0.4 0.3 5 0.2 0.1 0.2 0.4 0.6 0.8 ROD INSERTION (FRACTION)
Shearon Harris Figure 15.1-3 Normalized RCCA Reactivity Worth vs.
Rod Insertion
0.9 0.8 0.7 0.6 0.5 OA 0.3 0.2 DASHPOT 0.t TINE AFTER ROD DROP BEGINS (SECONDS)
Shearon Harris Figure 15.1-4 Normalized RCCA Reactivity Worth vs.
Time After Rod Drop Begins
20
-19
-18
-17 NOTE 1
-16
-15
-14
-13
-12 10
- '9 NOTE 2 8
7 6
NOTE 1: "UPPER CURVE" HOST NEGATIVE DOPPLER ONLY POWER DEFECT
5. ~ -1.6% hK 4
3 2
NOTE 2: "LOWER CURVE" w LEAST NEGATIVE DOPPLER ONLY POWER DEFECT 1 ~ -0.78% hK 20 40 60 80 100 PERCENT POWER Shearon Harris Figure 15.1-5 Doppler Power Coefficient Used In Accident Analyses
H Cy fi"
10 0 10-1 I
2 10 101 10 0 10 10 10 TIME AFTER SHUTDOWN (SECONDS)
Shearon Harris Figure 15. 1-7 1979 ANS Decay Heat
N+ 4 FILM N- I CLAD GAP I - I FUEL FUEL ROD CROSS SECTION Shearon Harris Figure 15.1-8 Fuel Rod Cross Section
15.2 CONDITION II - FAULTS OF MODERATE FREQUENCY These faults result at worst in reactor shutdown with the plant being capable of returning to operation. By definition, these faults (or events) do not propagate to cause a more serious fault, i.e., a Condition III or IV faul't.
In addition, Condition II events are not expected to result in fuel rod failures or reactor coolant system (RCS) overpressurization. For the purposes of this report the following faults have been grouped into this category:
(1) Uncontrolled rod cluster control assembly (RCCA) bank withdrawal from a subcritical condition (2} Uncontrolled RCCA bank withdrawal at power (3) RCCA misoperation (4) Uncontrolled boron dilution (5) Partial loss of forced reactor coolant flow (6) Startup of an inactive reactor coolant loop (7) Loss of external electrical load and/or turbine trip (8) Loss of normal feedwater (9) Loss of offsite power to the station auxiliaries (station blackout)
(10) Excessive heat removal due to feedwater system malfunctions (11) Excessive load increase (12} Accidental RCS depressurization 1578v:1D/120688 15.2-1
(13) Inadvertent opening of a steam generator relief or safety valve.
( 14) Inadvertent Operation of the Emergency Core Cooling System. at power.
Each of these faults of moderate frequen'cy are analyzed in this section. In general, each analysis includes an identification of causes and description of the accident, an analysis of effects and consequences, a presentation of results, and relevant conclusions.
The time sequence of events for the Condition II faults are shown in Table 15.2-1.
1 578 v:10/020389 15.2-2
15.2. 1 Uncontrolled Rod Cluster Control Assembl Bank Withdrawal from a Subcritical Condition 15.2. 1. 1 Identification of Causes and Accident Descri tion An RCCA withdrawal accident is defined as an uncontrolled increase in reactivity in the reactor core caused by withdrawal of RCCAs resulting in a power excursion. Such a transient could be caused by a malfunction of the reactor control or control rod drive systems. This could occur with the reactor at either subcritical, hot zero power, or at power. The at-power case is discussed in Section 15.2.2.
I.
The RCCA drive mechanisms are wired into preselected bank configurations that are not altered during core reactor life. These circuits prevent the assemblies from being withdrawn in other than their respective banks. Power supplied to the banks is controlled so that no more than two banks can be withdrawn at the same time. The RCCA drive mechanisms are of the magnetic latch type and coil actuation is sequenced to provide variable speed travel.
The maximum reactivity insertion rate analyzed in the detailed plant analysis is that occurring with the simultaneous withdrawal of the two control banks having the maximum combined worth at maximum speed.
The neutron flux response to a continuous reactivity insertion is characterized by a very fast rise terminated by the reactivity feedback effect of the negative Doppler coefficient. This self-limitation of the power burst is of primary importance since it limits the power to a tolerable level during the delay time for protection action. Should a continuous RCCA withdrawal accident occur, the transient will be terminated by the following automatic features of the reactor protection system:
1578v:1D/120688 15.2-3
4'4l 15.2. 1. 1. 1 Source Ran e Hi h Neutron Flux Reactor Tri The source range high neutron flux'eactor trip is actuated when either of two independent source range channels indicates a neutron flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed when either intermediate range flux channel indicates a flux level above a specified level. It is automatically reinstated when both intermediate range channels indicate a flux level below a specified level.
15.2. 1. 1.2 Intermediate Ran e Hi h Neutron Flux Reactor Tri The intermediate range high neutron flux reactor trip is actuated when either of two independent intermediate range channels indicates a flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed when two of the four power range. channels give readings above approximately 10% of full power and is automatically reinstated when three of the four channels indicate a power below this value.
15.2. 1. 1.3 Power Ran e Hi h Neutron Flux Reactor Tri (Low Settin )
The power range high neutron flux trip (low setting) is actuated when two-out-of-four power range channels indicate a power level above approximately 25% of full power. This trip. function may be manually bypassed when two of the four power range channels indicate a power level above approximately 10% of full power and is automatically reinstated when three. of the four channels indicate a power level below this value.
15.2. 1.1.4 Power Ran e Hi h Neutron Flux Reactor Tri (Hi h Settin )
The power range high neutron flux reactor trip (high setting) is actuated when two-out-of-four power range channels indicate a power level above a preset setpoint. This trip function is always active. In addition, control rod stops on high intermediate range flux level (one-of-two) and high power range flux level (one-out-of-four) serve to discontinue rod withdrawal and prevent the need to actuate the intermediate range flux level trip and the power range flux level trip, respectively.
1578v:1D/120588 15.2-4
ES 15.2.1.1.5 Hi h Neutron Flux Rate Tri The high neutron flux rate trip is actuated when the rate of change in power exceeds the positive or negative setpoint in two-out-of-four power range channels. This function is always active.
15.2. 1.2 Anal sis of Effects and Conse uences The analysis of the uncontrolled rod withdrawal from subcritical accident is performed in three stages: first a core average nuclear power transient calculation is performed, followed by an average core heat transfer calculation, and finally a DNBR calculation. The core average nuclear power transient calculation is performed using a spatial neutron kinetics code, T'k'INKLE(1) , to determine the average power generation with time including the various total core feedback effects, i.e., Doppler and moderator reactivity. The average heat flux and temperature transients are determined by performing a fuel rod transient heat transfer calculation in FACTRAN The average heat flux is next used in THINC((6,8) ' for the transient DNBR calculation.
The core axial power distribution is severely peaked to the bottom of the core for the limiting transient. The M-3 DNB correlation is used to evaluate DNBR in the span between the lower nonmixing vane grid and the first mixing vane grid. The MRB-1 correlation (LOPAR fuel) and the MRB-2 correlation (VANTAGE 5 fuel) remain applicable for the rest of the fuel assembly.
In order to give conservative results for a startup accident, the following assumptions are made concerning the initial reactor conditions:
(1) Since the magnitude of the power peak reached during the initial part of the transient for any given rate of reactivity insertion is strongly dependent on the Doppler coefficient, conservative values (low absolute magnitude) as a function of power are used. See Section 15.1.5 and Table 15. 1-3.
1578v:1D/120688 15.2-5
II (2) Contribution of the moderator reactivity coefficient is negligible during the initial part of the transient because the heat transfer time between the fuel and the moderator is much longer than the neutron flux response time. However, after the initial neutron flux peak, the succeeding rate of power increase is affected by the moderator reactivity coefficient. A conservative value, given in Table 15. 1-3, is used in the analysis to yield the maximum peak heat flux.
(3) The reactor is assumed to be at hot zero power. This assumption is more conservative than that of a lower initial system temperature.
The higher initial system temperature yields a larger fuel-water heat transfer coefficient, larger specific heats, and a less negative (smaller absolute magnitude) Doppler coefficient, all of which tend to reduce the Doppler feedback effect thereby increasing the neutron flux peak. The initial effective multiplication factor is assumed to be 1 since this results in maximum neutron flux peaking.
(4) Reactor trip is assumed to be initiated by power range high neutron flux (low setting). The most adverse combination of instrument and setpoint errors, as well as delays for trip signal actuation and RCCA release, is taken into account. A 10% increase is assumed for the power range flux trip setpoint, raising it from the nominal value of 25 to 35%. Previous results, however, show that the rise in neutron flux is so rapid that the effect of errors in the trip setpoint on the actual time at which the rods are released is negligible. In addition, the reactor trip insertion characteristic is based on the assumption that the highest worth RCCA is stuck in its fully withdrawn position. See Section 15. 1.4 for RCCA insertion characteristics.
1578v:1D/120688 15.2"6
(5) The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two control banks having the greatest combined worth at maximum speed (45 inches/minute). Control rod drive mechanism design is discussed in Section 4.2.3 of the FSAR.
(6) The initial power level was assumed to be below the power level expected for any shutdown condition. The combination of highest reactivity insertion rate and lowest initial power produces the highest peak heat flux.
(7) Two reactor coolant pumps are assumed to be operating.
15.2. 1. 3 Results The calculated sequence of events for this accident is shown on Table 15.2-1.
Figures 15.2. 1-1 and 15.2. 1=2 show the transient behavior for the indicated reactivity insertion rate with the accident terminated by reactor trip at 35%
nominal power. This insertion rate is greater than that for the two highest worth control banks, both assumed to be in their highest incremental worth region.
Figure 15.2. 1-1 shows the nuclear power transient. The nuclear power overshoots the full power nominal value but this occurs for only a very short time period. Hence, the energy release and the fuel temperature increase are relatively smally The thermal flux response, of interest for departure from nucleate boiling (ONB) considerations, is shown on Figure 15.2. 1-1. The beneficial effect on the inherent thermal lag in the fuel is evidenced by a peak heat flux less than the full power nominal value.
Figure 15.2. 1-2 shows the response of the hot spot fuel average and clad temperatures.
157Bv:1D/120588 15.2-7
15.2.1.4 Conclusions In the event of an RCCA withdrawal accident from the subcritiral condition, the core and the RCS are not adversely affected since the combination of thermal power and the coolant temperature r'esult in a departure from nucleate boiling ratio (DNBR) greater than the design limit value. Thus, no fuel or clad damage is predicted as a result of ONB.
157Sv:1D/120588 15.2-8
0 15.2.2 Uncontrolled Rod Cluster Control Assembl Sank Withdrawal at Power 15.2.2.1 Identification of Causes and Accident Oescri tion Uncontrolled RCCA bank withdrawal at power results in an increase in the core heat flux: Since the heat extraction from the-steam generator lags behind the core power generation until the steam generator pressure reaches the relief or safety valve setpoint, there is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result in ONS. Therefore, in order to avert damage to the cladding, the reactor protection system is designed to terminate any such transient before the DNBR falls below the safety analysis limit values.
The automatic features of the reactor protection system that prevent core damage following the postulated accident include the following:
(1) The power range neutron flux instrumentation actuates a reactor trip if two-out-of-four channels exceed a high Flux setpoint; (2) The reactor trip is actuated if any two-out-of-three aT channels exceed an overtemperature aT setpoint. This setpoint is automatically varied with axial power imbalance, coolant temperature, and pressure to protect against ONB; (3) The reactor trip is actuated if any two-out-of-three .hT channels exceed an overpower aT setpoint to ensure that the allowable heat gen'eration rate (kw/ft) is not exceeded; (4) A high pressurizer pressure reactor trip actuated from any
.two-out-of-three pressure channels that are set at a fixed point.
This set pressure is less than the set pressure for the pressurizer safety valves; (5) A high pressurizer water level reactor trip actuated from any two-out-of-three level channels that are set at a fixed point.
1578v:lo/120588 15.2-9
)
1n addition to the above listed reactor trips, there are the following RCCA withdrawal blocks:
(1) High neutron flux (one-out-of-four);
(2) Overpower aT (two-out-of-three);
(3) Overtemperature aT (two-out-of-three).
Figure 15. 1-1 presents allowable reactor coolant loop average temperature and hT for the design power distribution and flow as a function of primary coolant pressure. The boundaries of operation defined by the overpower hT trip and the overtemperature aT are represented as protection lines on this diagram. The protection lines are drawn to include all adverse instrumentation and setpoint errors so that under nominal conditions a trip would occur well within the area bounded by these lines. The utility of this diagram is in the fact that the limit imposed by a given DNBR can be represented as a line. The DNB lines represent the locus of conditions for which the DNBR equals the safety analysis limit value. All points below and to the left of a DNB line for a given pressure have a DNBR greater than the limit. The diagram shows that DNS is prevented for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable DNBR line at any point.
The area of permissible operation (power, pressure, and temperature) is bounded 'by the combination of reactor trips: high neutron flux (fixed setpoint); high-press0re (fixed setpoint); low-pressure (fixed setpoint);
overpower and overtemperature AT (variable setpoints).
15.2.2.2 Anal sis of Effects and Conse uences The uncontrolled RCCA bank withdrawal at power transient's analyzed by the LOFTRAN code (3) . This code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent 1578v:1D/lz0588 15.2-10
plant variables i'ncluding temperatures, pressures, and power level. The core limits as illustrated on Figure 15. 1-1 are used as input to LOFTRAN to determine the minimum DNBR during the transient.
This accident is analyzed with the Improved Thermal Design Procedure as described in Reference 4. In order to obtain conservative results, the following assumptions are made:
(1) Initial conditions of nominal core power and reactor coolant average temperatures and nominal reactor coolant pressure are assumed.
Uncertainties in initial conditions are included in the limit DNBR as described in Reference 4; (2) Reactivity Coefficients - two cases are analyzed:
(a) Minimum reactivity feedback. A positive moderator coefficient of reactivity of +5 pcm/'F is assumed. A variable Doppler power coefficient with core power is used in the analysis. A conservatively small (in absolute magnitude) value is assumed; (b) Haximum reactivity feedback. A conservatively large positive moderator density coefficient and a large (in absolute magnitude) negative Doppler power coefficient are assumed; (3) The reactor trip on high neutron flux is assumed to be actuated at a conservative value of 118% of nominal full power. The hT trips include all adverse instrumentation and setpoint 'errors, while the delays for the trip signal actuation are assumed at their maximum values; (4) The RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position; l578v:10/1205B8
Fg f lf I
(5) The maximum positive reactivity insertion rate is greater than that which would be obtained from the simultaneous withdrawal of the two control rod banks having the maximum combined worth at maximum speed.
The effect of RCCA movement on the axial core power distribution is accounted for by causing. a decrease in overtemperature and overpower aT trip setpoints proportional to a decrease in margin to ONB.
15.2.2.3 Results Figures 15.2.2-1 and 15.2.2-2 show the response of nuclear power, pressure, average coolant temperature, and DNBR to a rapid RCCA withdrawal starting from full power. Reactor trip on high neutron flux occurs shortly after the start of the accid'ent. Since this is rapid with respect to the thermal time constants of the plant, small changes in T and pressure result and a avg large margin to ONB is maintained.
The response of nuclear power, pressure, average coolant temperature, and DNBR for a slow control rod assembly withdrawal from full power is shown on Figures 15.2.2-3 and 15.2.2-4. Reactor trip on overtemperature aT occurs after a longer period and the rise in temperature and pressure is consequently larger than for rapid RCCA'ithdrawal. Again, the minimum DNBR is never less than the safety analysis, limit values.
Figure 15.2.2-5 shows the minimum DNBR as a function of reactivity insertion rate from initial full power operation for the minimum and for the maximum reactivity feedbacks. It can be seen that two reactor trip channels provide protection over the whole range of reactivity'nsertion rates. These are the high neutron-flux and overtemperature aT trip channels. The minimum ONBR is never less than the safety analysis limit values.
Figures 15.2.2-6 and 15.2.2-7 show the minimum DNBR as a function of reactivity insertion rate for RCCA withdrawal incidents starting at 60% and 10% power, respectively. The results are similar to the 100% power case, except that as the initial power is decreased, the range over which the overtemperature aT trip is effective is increased. In neither case does the ONBR fall below the safety analysis limit values.
1578v:1D/120588 . 15.2-12
The 'shape of the curves of minimum DNH ratio versus reactivity insertion rate in the reference figures is due both to reactor core and coolant system transient response and to protection system action in initiating a reactor trip.
Referring to Figure 15.2.2-7, for example, it is noted that:
1. For reactivity insertion rates above - 15 pcm/sec, reactor trip is initiated by the high neutron flux trip for the minimum reactivity feedback cases. The neutron flux level in the core rises rapidly for these insertion rates while core heat flux and coolant system temperature lag behind due to the thermal capacity of the fuel and coolant system fluid. Thus, the reactor is tripped prior to a significant increase in heat flux or water temperature with resultant high minimum DNH ratios during the transient. As the reactivity insertion rate decreases, core heat f)ux and coolant temperatures can remain more nearly in equi librium with the neutron flux. Minimum DNBR during the transient thus decreases with decreasing insertion rate.
2. The overtemperature hT reactor trip circuit initiates a reactor'rip
, when measured coolant loop hT exceeds a setpoint based on measured Reactor Coolant System average temperature and pressure. It is important to note .that the average temperature contribution to the circuit is lead-lag compensated in order to decrease the effect of the thermal capacity of the Reactor Coolant System in response to power increases.
3. For reactivity insertion rates below - 15 pcm/sec, the overtemperature
~T trip terminates the transient.
For reactivity insertion rates between - 15 pcm/sec and - 7 pcm/sec, the effectiveness of the overtemperature hT trip increases (in terms of increased minimum DNBR) due to the fact that with lower insertion rates the power increase rate is slower, the rate of rise of average coolant temperature is slower and the system lags and delays become less significant.
1 578 v:1D/120588 15.2-13
4. For reactivity insertion rates less than - 7 pcm/sec, the rise in the reactor coolant temperature is sufficiently high 'so that the steam generator safety valve setpoint is reached prior to trip. Opening of these valves, which act as an additional heat load on the Reactor Coolant System, sharply decreases the rate of increase of Reactor Coolant System average temperature. This decrease in rate of increase of the average coolant system temperature during the transient is accentuated by the lead-lag compensation causing the overtemperature hT trip setpoint to be reached later with a resulting lower minimum DNBR.
For transients initiated from higher power levels (for example, see Figure 15.2.2-5), the effect described in item 4 above, which results in the sharp peak in minimum DNHR at approximately 7 pcm/sec, does not occur since the steam generator safety valves are not actuated prior to trip.
Figures 15.2.2-5, 15.2.2-6, and 15.2.2-7 illustrate minimum DNHRs calculated for minimum and maximum reactivity feedback.
Since the RCCA withdrawal at power incident is an overpower transient, the fuel temperatures rise during the transient until after reactor trip occurs.
For high reactivity insertion rates, the overpower transient is fast with respect to the fuel rod thermal time constant, and the core heat flux lags behind the neutron flux response. Due to this lag, the peak core heat flux does riot exceed 118 percent of its nominal value (i.e., the high neutron flux trip setpoint assumed in the analysis). Taking into account the effect of the RCCA withdrawal on the axial core power distribution, the peak fuel centerline temperature will still remain below the fuel melting temperature.
For slow reactivity insertion rates, the core heat flux remains more nearly in equilibrium with the neutron flux. The overpower transient is terminated by the overtemperature hT reactor trip before a DNH condition is reacPed. The peak heat flux again is maintained below 118 percent of its nominal value.
Taking into account the effect of the RCCA withdrawal on the axial core power distribution, the peak fuel centerline temperature will remain below the fuel melting temperature.
1578v:1D/120588 15.2-14
Since DNB does not occur at any time during the RCCA withdrawal at power transient, the ability.of the primary coolant to remove heat from the fuel rod is not reduced. Thus, the fuel cladding temperature does not rise significantly above its initial value during the transient.
The calculated sequence of events for this accident is shown on Table 15.2-1.
'ith the reactor tripped, the plant eventually returns to a stable condition.
The plant may subsequently be cooled down further by following normal plant shutdown procedures.
15.2.2.4 Conclusions The high neutron flux and overtemperature hT trip channels provide adequate protection over the entire range of possible reactivity insertion rates; i.e.,
the minimum value of DNBR is always larger than the safety analysis limit values.
1578v:10/120588 15.2-15
i.2.3 t d C1 C 1 A b This section discusses RCCA misoperation that can result either from system malfunction or operator error.
15.2.3. 1 Identification of Causes and Accident Descri tion RCCA misalignment accidents include:
( 1) One or more dropped RCCAs within the same group; (2) A dropped RCCA bank; (3) Statically misaligned RCCA.
Each RCCA has a position indicator channel that displays the position of the assembly. The displays of assembly positions are grouped for the operator's convenience. Fully inserted assemblies are further indicated by a rod at bottom signal, which actuates a local alarm and a control room annunciator.
Group demand position is also indicated.
RCCAs are always moved in preselected banks, and the banks are always moved in the same preselected sequence. Each bank of RCCAs is divided into two groups. The rods comprising a group operate in parallel through multiplexing thyristors. The two groups in a bank move sequentially such that the first group is always within one step of the second group in the bank. A definite schedule of actuation (or deactuation of the stationary gripper, movable gripper, and lift coils of a mechanism) is requi.red to withdraw'the RCCA attached to the mechanism. Since the stationary gripper, movable gripper, and lift coils associated with the four RCCAs of a rod group are driven in parallel, any single failure that would cause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction of insertion, or immobility.
A dropped RCCA, or RCCA bank, is detected by:
(1) A sudden drop in the core power level as seen by the nuclear instrumentation system;
. 1578v:10/120588 15.2-16
(2) Asymmetric power distribution as seen on out-of-core neutron detectors or core-exit thermocouples; (3) Rod at bottom signal; (4) Rod deviatio'n alarm; (5) Rod position indication; (6) Negative neutron flux rate trip circuitry.
Hisaligned RCCAs are detected by:
(1) Asymmetric power distribution as seen on out-of-core neutron detectors or core-exit thermocouples; (2) Rod deviation alarm; (3) Rod position indicators.
The deviation alarm alerts the operator whenever an individual rod position signal deviates from the other rods in the bank by a preset limit. If the rod deviation alarm is not .operable, the operator is required to take action as required by the Technical Specifications If one or more rod position indicator channels should be out of service, detailed operating instructions are followed to .ensure the alignment of the nonindicated RCCAs. The operator is also required to take action as required by the Technical Specifications.
1578v:1D/120588 15.2-17
AP-.'y
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15.2.3.2 Anal sis of Effects and Conse uences Method of Analysis (1) One or More Dropped RCCAs from the Same Group For evaluation of the dropped RCCA event, the transient system response is calculated using the LOFTRAN code (3) . The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief safety valves,. pressurizer spray, steam .generator, and steam
'nd generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
Statepoints are calculated and nuclear models are used to obtain a hot channel factor consistent with the primary system conditions and reactor power. By incorporating the primary conditions from the transient and the hot channel factor from the nuclear analysis, the DNB design basis is shown to be met. The DNB analysis is performed using the THINC code (6,8)
' The transient response, nuclear peaking factor analysis, and DNB design basis confirmation are performed in accordance with the methodology described in
'I Reference 9.
(2) Dropped RCCA Bank Analysis is not required since the dropped RCCA bank results in a trip.
(3) Statically Misaligned RCCA Steady state power distributions are analyzed using the computer codes as described in Table 4.1-2 of the FSAR. The peaking factors are then used as input to the THING code to'alculate the DNBR.
1578v:1D/120688 15.2-18
~ '
15.2.3.3 Results (1) One or More Dropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion that may be detected by the power range negative neutron flux rate trip circuitry. If detected, the reactor is tripped within approximately 2.7 seconds following the drop of the RCCAs. The core is not adversely affected during this period since power is decreasing rapidly. Following reactor trip, normal shutdown procedures are followed. The operator may manually retrieve the RCCA by following approved operating procedures.
For those dropped RCCAs that do not result in a reactor trip, power may be reestablished either by reactivity feedback or control bank withdrawal. Following a dropped rod event in manual rod control, the plant will establish a new equi librium condition. The equilibrium process without control system interaction is monotonic, thus removing power .overshoot as a concern and establishing the automatic rod control mode of operation as the limiting case.
For a dropped RCCA event in the automatic rod control mode, the rod control system detects the drop in power and initiates control bank withdrawal. Power overshoot may occur due to this action by the automatic rod controller after which the control system will insert the control bank to restore nominal power. Figures 15.2.3-1 and 15.2.3-2 show a typical transient response to a dropped RCCA (or RCCAs) in automatic control.. In all cases, the minimum DNBR remains above the safety analysis limit value.
(2) Dropped RCCA Bank A dropped RCCA bank typically results in a reactivity insertion of greater than 400 pcm which will be detected by the power range negative neutron flux rate trip circuitry. The reactor is tripped within approximately 2.7 seconds following the drop of a RCCA bank.
1578v:10/120588 15.2-19
The core. is not adver'sely affected during this period since power is decreasing rapidly. Following the reactor trip, normal shutdown procedures are followed to further cool down the plant. Any action required of the operator to ma'intain the plant in a stabilized condition will be in a time frame in excess of 10 minutes following the incident.
(3) Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR 'at significant power levels arise from cases in which one RCCA is fully inserted, or where Bank D is fully inserted with one RCCA fully withdrawn. Multiple independent alarms, including a bank insertion limit alarm, alert the operator well before the postulated conditions are approached. The bank can be inserted to its insertion limit with any one assembly fully withdrawn without the DNHR falling below the safety analysis limit value.
The insertion limits in the Technical Specifications may vary from time to time depending on a number of limiting'riteria. It is preferable, therefore, to analyze the misaligned,RCCA case at full power for a position of the control bank as deeply inserted as the criteria on minimum DNBR and power peaking factor will allow. The full power insertion limits on control Hank D must then be chosen to be above that position and will usually be dictated by other criteria. Detailed results will vary from cycle to cycle depending on fuel arrangements.
For this RCCA misalignment, with Bank D inserted to its full power insertion limit and one RCCA fully withdrawn, DNBR does not fall below the safety analysis limit value. This case is analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values but with the increased radial peaking factor associated with the misaligned RCCA.
1578v:1D/120588 15.2-20
1 h f 1 4'
DNB calculations have not been performed specifically for RCCAs missing from other banks; however, power shape calculations have been done as required for the ful ly wi thdrawn anal ysi s. Inspection of the power shapes shows that the DNB and peak kW/ft situation is less severe than the Bank D case discussed above assuming insertion limits on the other banks equivalent to a Bank D full-in insertion 1 imi t.
For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value. This case is analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values, but with the increased radial peaking factor associated with the misaligned RCCA.
DNB does not occur for the RCCA misalignment incident and thus the ability of the primary coolant to remove heat from the fuel rod is not reduced. The peak fuel temperature corresponds to a linear heat generation rate based on the radial peaking factor penalty associated with the misaligned RCCA and the design axial power distribution. The resulting linear heat generation is well below that which would cause fuel melting.
Following the identification of an RCCA group misalignment condition by the operator, the operator is required to take action as required by the plant Technical Specifications and operating instructions.
15.2.3.4 Conclusions For all cases of dropped RCCAs or dropped banks, for which the reactor is tripped by the power range negative neutron flux rate trip, there is no reduction'in the margin to core thermal limits and, consequently, the DNB design basis is met. It is shown for all cases which do not result in reactor trip that the DNBR remains greater than the safety analysis limit value and, therefore, the DNB design basis is met.
1578v:1oi120588 15.2-21
For all cases of any RCCA inserted, or Bank D inserted to its rod insertion limits with any single,RCCA in that bank fully withdrawn (static misalignment), the DNBR remains greater than the safety analysis limit value.
1578v:lo/120588 15.2-22
0 15.2.4 Uncontrolled Boron Dilution 15.2.4. 1 Ideni,indication of Causes and Accident Descri tion Reactivity can be added to the core by feeding primary grade water into the reactor coolant system (RCS) via the reactor makeup portion of. the chemical and volume control system (CVCS). Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on the rate and duration of dilution. A boric acid blend system is provided to permit the operator to match the boron concentration of reactor coolant makeup water during normal charging to that in the RCS. The CVCS is designed to limit, even under various postulated failure modes, the potential rate of dilution to a value which, after indication through alarms and instrumentation, provides the operator sufficient .time to correct the situation in a safe and orderly manner.
The opening of the primary water makeup control valve provides makeup to the RCS which can dilute the reactor coolant. Inadvertent dilution from this source can be readily terminated by closing the control valve. In order for makeup water to be added to the RCS at pressure, at least one charging pump must be running in addition to a primary makeup water pump.
The rate of addition of unborated makeup water to the RCS is limited by a flow limiting orifice between the reactor makeup water pumps and the boric acid blender. As demonstrated by tests at the plant, flow is within the bounds of unborated water used in analyses in this section.
The boric acid from the boric acid tank is blended with primary grade water in the blender and the composition is determined by the preset flow rates of boric acid and primary grade water on the control board. In order to dilute two separate operations are required:
1. The operator must switch from the automatic makeup mode to the dilute mode;
2. The start/stop switch is in the start position.
1 578 v:1D/120588 15.2-23
Omitting either step would prevent dilution.
The status of the RCS makeup is continuously available to the operator by:
1. CVCS and RMWS pump status lights,
2. Deviation alarms if the boric acid or blended flow rates deviate from the preset values.
15.2.4,2 Anal sis of Effects and Conse uences Due to the change in the 'steady state uncertainties and the Overtemperature bT trip, only the Dilution at Power Mode was analyzed. Table 15.2-1 contains the time sequence of events for this mode of the Boron Dilution event.
During power operation, the dilution rate is limited by the capacity of the charging pumps. (The analysis is performed assuming all charging pumps are in operation, although only one is normally in operation,) The effective reactivity addition rate is a function of the reactor coolant temperature and boron concentration. The reactivity insertion rate calculated is based on a conservatively high value for the expected boron concentration at power (1875 ppm) as well as a conservatively high charging flow rate capacity (285 gpm). The RCS volume assumed (7450 ft3
) corresponds to the active volume of the RCS excluding the pressurizer, for one reactor coolant pump in operation.
15.2.4,3 Results and Conclusions Pith the reactor in automatic control, the power and temperature increase from boron dilution results in insertion of the RCCAs and a decrease in the shutdown margin. The rod insertion limit alarms (low and low-low settings) provide the operator with adequate time (over 15 minutes) to determine the cause'f dilution, isolate the primary grade water source, and initiate reboration before the loss of all shutdown margin.
1578v:1D/120588 15.2-24
With the reactor in manual control and if no operator action is taken, the power and temperature rise will cause the reactor to reach the overtemperature aT trip setpoint. The boron dilution accident in this case is similar to a RCCA withdrawal accident. The maximum reactivity insertion rate for boron dilution is approximately 1.78 pcm/sec and is within the range of insertion rates analyzed. Prior to the overtemperature aT trip, an overtemperature hT alarm and turbine runback would be actuated. There is adequate time available (over 15 minutes after a reactor trip for the operator to determine the cause of dilution, isolate the primary grade water sources and initiate reboration before the loss of all shutdown margin.
The results presented above show that there is adequate time for the operator to manually terminate the source of dilution flow. Following termination of the dilution flow, the reactor will be in a stable condition. The operator can then initiate reboration to recover the shutdown margin.
1578v:10/120588 15.2-25
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15.2.5 Partial Loss of Forced Reactor Coolant Flow 15.2.5. 1 Identification of Causes and Accident Descri tion A partial loss of coolant flow accident can result from a mechanical or electrical failure in a reactor coolant pump, or from a faul.t in the power supply to the pump. If the reactor is at power at the time of the accident, the immediate effect of a loss of forced reactor co'olant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor is not tripped promptly.
The necessary protection against a partial loss of coolant flow accident is provided by the low primary coolant flow reactor trip that is actuated by two-out-of-three low flow signals in any reactor coolant loop. Above Permissive 8, low flow in any loop will actuate a reactor trip. Between Permissive 7 and the power level corresponding to Permissive 8, low flow in any two loops will actuate a reactor trip. Reactor trip on low flow is blocked below Permissive 7.
A reactor trip signal from the pump breaker position is also provided. Rhen operating above Permissive 7, a breaker open signal from any two pumps will actuate a reactor trip. This serves as a backup to the low flow trip.
Reactor trip on reactor coolant pump breakers open is blocked below Permissive 7.
Normal power for each pump is supplied through individual buses connected to the isolated phase bus duct between the generator circuit breaker and the main transformer. Faults in the substation may cause a trip of the main transformer high side circuit breaker leaving the generator to supply power to the reactor coolant pumps. -when a generator circuit breaker trip occurs because of electrical faults, the pumps are automatically transferred to an alternate power supply and the pumps will continue to supply coolant flow to the core. Following any turbine trip where there are no electrical faults, ~
the generator circuit breaker is tripped and the reactor coolant pumps remain connected to the network through the transformer high side breaker.
Continuity of power to the pump buses is achieved without motoring the 1578v:1D/120588 15.2-26
generator since means are provided to isolate the generator without isolating the pump-buses from the external power lines (e.g., a generator output breaker is provided as well as a station output breaker).
15.2.5.2 Anal sis of Effects and Conse uences 15.2.5.2. 1 Method of Anal sis The following case has been analyzed:
All loops operating, one loop coasting down This transient is analyzed by three digital computer codes. First the LOFTRAN code (3) is used to calculate the loop and core flow during the transient.
~
The LOFTRAN code is also used to calculate the time of reactor trip, based on the calculated flows and the nuclear power transient following reactor trip.
~
on the nuclear power and flow from LOFTRAN. Finally, the THING code 's The FACTRAN code (2) is then used to calculate the heat flux transient based used to calculate the minimum DNBR during the transient based on the heat flux from FACTRAN and the flow from LOFTRAN. The DNBR transient presented represents the minimum of the typical and thimble cells for Standard and VANTAGE 5 fuel.
I 15.2.5.2.2 Initial Conditions The assumed initial operating conditions are the most adverse with respect to the margin to DNB, i.e., nominal steady state power level, nominal steady state pressure, and nominal steady state coolant average temperature. See Section 15.1.2 for an explanation of initial conditions. The accident is analyzed using the Improved Thermal Design Procedure as described in Reference 4.
15.2.5.2.3 Reactivit Coefficients A conservatively large absolute value of the Doppler-only power coefficient is used (see Table 15. 1-3). The total integrated Doppler reactivity from 0 to 100% power is assumed to be -0.016 Ak.
1578v:10/120688 15.2-27
The least negative moderator temperature coefficient at full power (+5 pcm/'F) is assumed since this results in the maximum hot spot heat flux during the initial part of the transient when the minimum DNBR is reached.
15.2.5.2.4 Flow Coastdown The flow coastdown analysis is based on a momentum balance around each reactor coolant loop and across the reactor core. This momentum balance is combined with the continuity equation, a pump momentum balance, and the pump characteristics and is based on high estimates of system pressure losses to calculate the flow coastdown.
15.2.5.3 Results The calculated sequence of events is shown in Table 15.2-1. Figures 15.2.5-1 and 15.2.5-2 show the vessel flow coastdown, the faulted loop flow coastdown, the nuclear power and heat flux transient. The minimum DNBR is not less than the safety analysis limit value. A plot of DNBR vs. time is given in Figure 15.2.5-3 for the most limiting typical or thimble cell for Standard and VANTAGE 5 fuel.
15.2.5.4 Conclusions The analysis shows that the DNBR will not decrease below the safety analysis limit values at any time during the transient. Thus, no core safety limit is violated.
~ 1578 v:1D/120588 15.2-28
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15.2. 6 Star tu of an Inacti ve Reac tor Cool ant Loo In accordance with Technical Specification 3/4.4.1, Shearon Harris operation during startup and power operation with less than three loops operating is not permitted. This analysis is presented for completeness.
15,2.6. 1 Identification of Causes and Accident Oescri tion If a plant is operating with one pump out of service, there is reverse flow through the loop due to the pressure difference across the reactor vessel.
The cold leg temperature in an inactive loop is identical to the cold leg temperature of the active loops (the reactor core inlet temperature). If the reactor is operated at power, and assuming the secondary side of the steam generator in the, inactive loop is not isolated, there is a temperature drop across the steam generator in the inactive loop and, with the reverse flow, the hot'leg temperature of the inactive loop is lower than the reactor core inlet temperature.
Administrative procedures require that the unit be brought to a load of less than 25% of full power prior to starting a pump in an inactive loop in order to bring ihe inactive loop hot leg temperature closer to the core inlet temperature. Starting of an idle reactor coolant pump without bringing the inactive loop hot leg temperature close to the core inlet temperature would result in the injection of cold water into the core which causes a rapid reactivity .insertion and subsequent power increase.
This event is classified as an ANS Condition II incident (an incident of moderate frequency) as defined in Section 15.0.
Should the startup of an inactive reactor coolant pump at an incorrect temperature occur, the transient will be terminated automatically by a reactor trip on low coolant loop flow when the power range neutron flux (two-out-of-four channels) exceeds the Permissive 8 setpoint, which has been previously reset for two loop operation.
157Bv:1D/120588 15.2-29
15.2.6.2 Anal sis of Effects and Consequences This transient is analyzed by three digital computer codes. The LOFTRAN Code (3) is used to calculate the loop and core flow, nuclear power and core
~
pressure'nd temperature transients following the startup of an idle pump.
(2) is used to calculate the core heat flux transient based on core
~
FACTRAN flow and nuclear power from LOFTRAN. The THINC Code'(6 8) 's then used to calculate the ONBR during the transient based on system conditions (pressure, temperature, and flow) calculated by LOFTRAN and heat flux as calculated by FACTRAN.
In order to obtain conservative results for the startup of an inactive pump accident, the following assumptions are made:
(1) Initial conditions of maximum core power and reactor coolant'average temperatures and minimum reactor coolant pressure resulting in minimum initial margin to DNB. These values are consistent with the assumed maximum steady state power level corresponding to all but one loop in operation including allowances for calibration and instrument errors. The high initial power gives the greatest temperature difference between the core inlet temperature and the inactive loop hot leg temperature.
(2) Following the start of the idle pump, the inactive loop flow reverses and accelerates to its nominal full flow value.
(3) A conservatively large (absolute value) negative moderator temperature coefficient associated with end of life conditions.
(4) A conservatively low (absolute value) negative Doppler power coefficient is used.
(5) The initial reactor coolant loop flows are at the appropriate values for one pump out of service.
1578v:1D/120588 15.2-30
(6) The reactor trip is assumed to occur on low coolant flow when the power range neutron flux exceeds the Permissive 8 setpoint. The Permissive 8 setpoint is conservatively assumed to be 79 percent of rated power.
Results '5.2.6.3 The results following the startup of an idle pump with the above listed assumptions are shown in Figures 15.2.6-1 through 15.2.6-4. As shown in these curves, during the first part of the transient, the increase in core flow with cooler water results in an increase in nuclear power and a decrease in core average temperature. The minimum DNBR during the transient is considerably greater than the safety analysis limit values.
Reactivity addition for the inactive loop startup accident is due to the decrease in core water temperature. During the transient, this decrease is due both to (1) the increase in reactor coolant flow and, (2) as the inactive loop flow reverses, to the colder water entering the core from the hot leg side (colder temperature 'side prior to the start of the transient) of the steam generator in the inactive loop. Thus, the reactivity insertion rate for this transient changes with time. The resultant core nuclear power transient, computed with consider'ation of both moderator and Doppler reactivity feedback effects, is shown on Figure 15.2.6-1.
The calculated sequence of events for this accident is shown in Table 15.2-1.
The transient results illustrated in Figures 15.2.6-1 through 15.2.6-4 indicate that a stabiliied plant condition, with the reactor tripped, is approached rapidly. Plant cooldown may subsequently be achieved by following normal shutdown procedures.
15.2.6.4 Conclusions The transient results show that the core is not'dversely affected. There is considerable margin to the safety analysis DNBR limit values; thus, no fuel or clad damage is predicted.
1578v:1D/120588 15.2-31
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15.2.? Loss of External Electrical Load and/or Turbine Tri 15.2.7. 1 Identification of Causes and Accident Descri tion A major load loss on the plant can result from either a loss of external electrical load or from a turbine trip. For either case, offsite power is available for the continued operation of plant components such, as the reactor coolant pumps. The case of loss of all ac power (station blackout) is analyzed in Section 15.2.9.
For a turbine trip, the reactor would be tripped directly (unless it is below approximately 10% power) from a signal derived from the turbine stop emergency trip fluid pressure and turbine stop valves. The automatic Steam Dump System accommodates the excess steam generation. Reactor coolant temperatures and pressure do not significantly increase if the Steam Dump System and Pressurizer Pressure Control System are functioning properly. If the turbine condenser were not available, the excess steam generation would be dumped to the atmosphere. Additionally, main feedwater flow would be lost if the turbine condenser were not available. For this situation, steam generator level would be maintained by the Auxiliary Feedwater System.
For a loss of external electrical load without subsequent turbine trip, no direct reactor trip signal would be generated. The plant would be expected to trip from the Reactor Protection System if a safety limit were approached. A continued steam load of approximately 5% would exist after total loss of external electrical load because of the steam demand of plant auxiliaries.
In the event the steam dump .valves fail to open following a large loss of load, the steam generator safety valves may lift and the reactor may be tripped by the high pressurizer pressure signal, the high pressurizer water level signal, low-low steam generator water level signal, or the overtemperature hT signal. The steam generator shell-side pressure and reactor coolant temperatures will increase rapidly. The pressurizer safety valves and steam generator safety valves are, however, sized to protect the RCS and steam generator against overpressure for all load losses without 1578v:1D/120588 15.2-32
dt assuming the operation of the Steam Dump System, p:. "surizer spray, pressurizer power-operated relief valves, automatic rod cluster control assembly control, or direct reactor trip on turoine .':rip.
The steam generator safety valve capacity is sized to remove the steam flow at the engineered safeguards design rating (104.5% of steam flow at rated power) from the steam generator without exceeding 110% of the steam system design pressure. The pressurizer safety valve capacity is sized bas'ed on a complete loss of heat sink with the plant in'itially operating at the maximum calculated turbine load along with operation of the steam generator safety valves. The pressurizer safety valves are then able to maintain the RCS pressure within 110% of the RCS design pressure without direct or immediate reactor trip action.
A more complete discussion of overpressure protection can be found in Reference 7.
15.2.7.2 Anal sis of Effects and Conse uences In this analysis, the behavior of the unit is evaluated for a complete loss of steam load from full power without a direct reactor trip. This 'is done to show the adequacy of the pressure-relieving devices and to demonstrate core protection margins. The reactor is not tripped until conditions in. the RCS or steam generator result in a trip. The turbine is assumed to trip without actuating all the turbine stop valve limit switches. This assumption delays reactor trip until conditions result in a trip due to other signals. Thus, the analysis assumes a worst case transient.. In addition, no credit is taken for steam dump. Hain feedwater flow is terminated at the time of turbine trip, with no credit taken for auxiliary feedwater (except for long-term recovery) to mitigate the consequences of the transient.
The loss of load transients are analyzed with the LOFTRAN computer program (see Section 15. 1). The program simulates the neutron kinetics, RCS, pressurizer, 'pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The program computes pertinent plant variables including temperatures, pressures, and power level.
1578v:1D/120688 15.2-33
Hajor assumptions are summarized below:
1 (1) Hoderator and Doppler Coefficients of Reactivity The turbine trip is analyzed with both maximum and minimum reactivity feedback. The maximum feedback (EOL) cases assume a large negative moderator temperature coefficient and the most negative Doppler power coefficient. The minimum feedback (BOL) cases assume a minimum moderator temperature coefficient and the least negative Doppler coefficient.
(2) Reactor Control From the standpoint of the maximum pressures attained, it is conser'vative to assume that the reactor is in manual control. If the reactor were in automatic control, the control rod banks would move prior to trip and reduce the severity of the transient.
(3) Steam Release No credit is taken for the operation of the Steam Dump System or steam generator power-operated reli'ef valves. The steam generator pressure rises to the safety valve setpoint where steam release through safety valves limits secondary steam pressure at the setpoint value.
(4) Pressurizer Spray and Power-operated Relief Valves Two cases for both the BOL and EOL are analyzed:
(a) Full credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are also available.
1578v:1D/120588 15.2",34
t E
(b) No credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are operable.
(5) Feedwater Flow Main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. No credit is taken for auxiliary feedwater flow since a stabilized plant condition will be reached before auxiliary feedwater initiation is normally assumed to occur; however, the auxiliary feedwater pumps would be expected to start on a trip of the main feedwater pumps. The auxiliary feedwater flow would remove core decay heat following plant stabilization.
(6) Initial Operating Conditions
.The cases which assume operation of pressurizer spray and power-operated relief valves will present a greater challenge to the limit DNBR than the cases which do not assume their operation. For these 'cases, initial reactor power, pressure and RCS temperature are assumed to be at their nominal values. Uncertainties in initial conditions are included in the limit DNBR as described in Reference 4.
The cases for which no credit is taken for the effect of pressurizer spray and power-operated relief valve will result in a greater challenge to the design pressure limit than the cases that do assume their operation. For these cases, reactor power and RCS pressure are assumed at their maximum values consistent with the steady-state full power operation including allowances for calibration and instrument errors. The initial RCS temperature is assumed at a minimum value consistent with steady-state full power operation including allowances for calibration and instrument errors. This results in a conservative predictjon of peak pressures for this event.
1 578 v:1D/120688 15.2-35
4
,I
(7) Reactor trip is actuated by the first Reactor Protection System trip setpoint reached with no credit taken for the direct reactor trip on the turbine trip.
15.2.7.3 Results The tr'ansient responses for a total loss of load from full power operation are shown for four cases; two cases for the BOL and two cases for the EOL on Figures 15.2.7-1 through 15.2.7-12.
Figures 15.2.7-1, 15.2.7-2 and 15.2.7-3 show the transient responses for the total loss of steam load at BOL assuming full credit for the pressurizer spray and pressurizer power-operated relief valves. No credit is taken for the steam dump. The reactor is tripped by the overtemperature aT trip channel.
The minimum DNBR is well above the limit value. The pressurizer safety valves are actuated for this case and maintain system pressure below 110 percent of the design value. The steam generator safety valves open and limit the secondary steam pressure increase.
Figures 15.2.7-4, 15.2.7-5 and 15.2.7-6 show the responses for the total loss of load at EOL assuming a large (absolute value) negative moderator temperature coefficient. All other plant parameters are the same as in the above case. The reactor is tripped by the low-low steam generator water level channel. The DNBR increases throughout the transient and. never drops below its initial value.
Total loss of load was also studied assuming the plant to be initially operating at full power with no credit taken for the pressurizer spray, pressurizer power-operated relief valves, or steam dump. The reactor is tripped .on the high pressurizer pressure signal. Figures 15.2.7-7, 15.2.7-8 and 15.2.7-9 show the BOL transients. The nuclear power remains at or above full power until the reactor is tripped. The DNBR generally increases throughout the transient. In this case, the pressurizer safety valves are actuated and maintain the system pressure below 110 percent of the design value.
1578v:1D/120588 15.2-36
Figures 15.2.7-10, 15.2.7-11 and 15.2.7-12 show the transient at EOL with the other assumptions being the same as on Figures 15.2.7-7 through 15.2.7-9.
Again, the .DNBR increases throughout the transient and the pressurizer safety valves are actuated to limit the primary pressure.
Reference 7 presents additional results for a complete loss of heat sink including loss of main feedwater. This report shows the overpressure protection that is afforded by the pressurizer and steam generator safety valves.
15.2.7.4 Conclusions Results of the analyses, including those in Reference 7, show that the plant design is such that a total loss of external electrical load without a direct or immediate reactor trip presents no hazard to the integrity of the RCS or the Main Steam System. Pressure-relieving devices incorporated in the two systems are adequate to limit the maximum pressures to within the design limits.
The integrity of the core is maintained by operation of the reactor protection system; i.e., the DNBR will be maintained above the safety analysis limit values. Thus, no core safety limit will be violated.
1578v:10/120588 15.2" 37
+ e 15.2.8 Loss of Normal Feedwater 15.2.8.1 Identification of Causes and Accident Descri tion A loss of normal feedwater (from pump failures, valve malfunctions, or loss of offsite ac power) results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the reactor were not tripped during this accident, core damage would possibly occur from a sudden loss of heat sink. If an alternative supply of feedwater were not supplied to the plant, residual heat following reactor trip would heat the primary system water to the point where water relief from the pressurizer would occur.
Si'gnificant loss .of water from the RCS could conceivably lead to core damage.
Since the plant is tripped well before the steam generator heat transfer capability is reduced, the primary system variables never approach a DNH condition.
The following provide the necessary protection against a loss of normal feedwater:
(1) Reactor trip on low-low water level in any steam generator; (2) Reactor trip on steam flow-feedwater flow mismatch in coincidence with low steam generator water level; (3) Two motor-driven auxiliary feedwater (AFW) pumps that are started on:
(a) Low-low level in any steam generator, (b) Trip of all main feedwater pumps, (c) Any safety injection signal, (d) Loss of offsite power (automatic transfer to diesel generators),
(e) Manual actuation.
1578v:1D/1205S8 15.2-38
(4) One turbine-driven AFR pump that is started on:
(a) Low-low level in any two steam generators, (b) Loss of offsite power, (c) Manual'ctuation.
The motor-driven AFW pumps are connected to vital buses and are supplied by the diesels if a loss of offsite power occurs. The turbine-driven pump utilizes steam from the secondary system and exhausts it to the atmosphere.
The controls are designed to start both types of pumps within 1 minute even if a loss of all ac power occurs simultaneously with loss of normal feedwater.
'The AFh'umps take suction from the condensate storage tank for delivery to the steam generators.
The analysis shows that following a loss of normal feedwater, the AFh'ystem is capable of removing the stored and residual heat thus preventing either overpressurization of the RCS or loss of water from the reactor core.
15.2.8.2 Anal sis of Effects and Consequences A detailed analysis using the LOFTRAN code is performed in order to determine the plant transient following a loss of normal feedwater. The code describes the plant thermal kinetics, RCS including natural circulation, pressurizer, steam generators, and feedwater system, and computes pertinent variables, includin'g the pressurizer pressure, pressurizer water level, and reactor coolant average temperature.
Major assumptions are:
(1) Reactor trip occurs on steam generator low-low level at 32.3% of narrow range span.
(2) The plant is initially operating at 102% of the ESF power rating (2910.3 Mwt).
1578v:1D/120588 15. 2-39
C~'
(3) Conservative core residual heat generation based on long-term operat'ion at the initial power level preceding the trip is assumed.
The 1979 decay heat ANSI 5.1 + 2 SIGMA was used for calculation of residual decay heat levels.
(4) The AFM system is actuated by the low-low steam generator water level signal.
(5) The AFM system is assumed to supply a total of 430 gpm to three steam generators from one motor-driven pump.
(6) The pressurizer sprays and PORVs are assumed operable. This maximizes the peak transient pressurizer water volume.
(7) Secondary system steam relief is achieved through the self-actuated safety valves. Note that steam relief will, in fact, be through the power-operated relief valves or condenser dump valves for most cases of loss of normal feedwater. However, for the sake of analysis these have been assumed unavailable.
(8) The initial reactor coolant average temperature is 5.3'F higher than the nominal value to allow for uncertainty on nominal temperature.
The initial pressurizer pressure uncertainty is 38 psi.
15.2.8.3 Results Figures 15.2.8-1 through 15.2.8-3 show plant parameters following a loss of normal feedwater.
Following the reactor and turbine trip from full load, the water level in the steam generators will fall due to the reduction of steam generator void fraction and because steam flow through the safety valves continues to dissipate the stored and generated heat. About one minute following the initiation of the low-low level trip, the motor-driven AFW pump is automatically started, reducing the rate of water level decrease.
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The capacity of the motor-driven AFR pump is such that the water level in the steam generator being fed does not recede below the lowest level at which
'ufficient heat trans; r area is available to dissipate core residual heat without water relief from the RCS relief or safety valves.
From Figure 15.2.8-3 it can be seen that at no time is there water .relief from the pressurizer. If the auxH iary feedwater delivered is greater than that of one motor-driven pump, the initial reactor power is less than 102% of the ESF design rating, or the steam generator water level in one or more steam generators is above the low-low level trip point at the time of trip, then the results for this transient wi 1] be less limiting.
The calculated sequence of events for this accident is listed in Table. 15.2-1.
As shown in Figures 15.2.8-1 through 15.2.8-3, the plant approaches a
'tabilized condition following reactor trip and auxiliary feedwater initiation. Plant procedures may be followed to further cool down the plant.
15.2.8.4 Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the AFH capacity is such that the reactor coolant water is not relieved from the pressurizer relief or safety valves.
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15.2.9 Loss of Offsite Power to the Station Auxiliaries Station Blackout) 15.2.9. 1 Identification of Causes and Accident Descri tion During a complete loss of offsite power and a turbine trip there will be loss of power to the plant auxiliaries, i.e., the reactor coolant pumps, condensate pumps, etc.
The events following a loss of ac power with turbine and reactor trip are described in the sequence listed below:
(1) Plant vital instruments are supplied by emergency power sources.
(2) As the steam system pressure rises following the trip, the steam system power-operated relief valves are automatically opened to the atmosphere. Steam dump to the condenser is assumed not to be available. If the power-operated relief valves are not available, the steam generator self-actuated safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the residual heat produced in the reactor.
(3) As the no-load temperature is approached, the steam system power-operated relief valves (or thy self-actuated safety valves, if the power-operated relief valves are not available) are used to dissipate the residual heat and to maintain the plant at the hot standby condition.
(4) The emergency diesel generators started on loss of voltage on the plant emergency buses begin to supply plant vital loads.
The AFH system is started automatically as discussed in the loss of normal feedwater analysis. The steam-driven AFH pump utilizes steam from the secondary system arid exhausts to the atmosphere. The two motor-driven AFW pumps are supplied by power from the diesel generators. The pumps take suction directly from the condensate storage tank for delivery to the steam generators.
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f,4 e,
Upon the loss of power to the reactor coolant pumps, coolant flow;, cessary for core cooling and the removal of residual heat is maintained by natu". al circulation in the reactor coolant loops.
15.2.9.2 Anal ~sis of Effects and Conse uences A detailed analysis using the LOFTRAN code is performed in order to determine the plant transient following a station blackout. The code describes the plant thermal kinetics, RCS including natural circulation, pressurizer, steam generators, and feedwater system, and computes pertinent variables, including the pressurizer pressure, pressurizer water, level, and reactor coolant average temperature.
Hajor assumptions differing from those in a loss of normal feedwater are:
(1) No credit is taken for immediate response of control rod drive mechanisms caused by a loss of offsite power.
(2) A heat transfer coefficient in the steam generator associated with RCS natural circulation is assumed following the reactor coolant pump coastdown.
(3) The initial reactor coolant average temperature is 6.8 F lower than nominal.
I (4) The AFM system is assumed to supply a total of 430 gpm to two steam generators from one motor-driven pump.
15.2.9.3 Results The time sequence of events for the accident is given in Table 15.2-1. The first few seconds after the loss of power to the reactor coolant pumps will closely resemble a simulation of the complete loss of flow incident (see Section 15.3.2); i.e., core damage due to rapidly increasing core temperatures is prevented by promptly tripping the reactor. After the reactor trip, stored and residual heat must be removed to prevent damage to either the RCS or the core. Figures 15.2.9-1 through 15.2.9-3 show that the natural circulation 1578v:1D/120588 15.2-43
flow available is sufficient to provide adequate core decay heat removal following reactor trip and RCP coastdown.
15.2.9.4 Conclusions Results of the analysis show that, for the loss of offsite power to the station auxiliaries event, all safety criteria are met. Since the DNBR remains above the safety analysis limit, the core is not adversely affected.
AFh'apacity is sufficient to prevent water relief through the pressurizer relief and safety valves; this assures that the RCS is not overpressurized.
Anal'ysis of the natural circulation capability of the RCS demonstrates that sufficient long-term heat removal capability exists following reactor coolant pump coastdown to prevent fuel or clad damage.
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15.2.10 Excessive Heat Removal Due to Feedwater S stem Malfunctions 15.2. 10. 1 Identification of Causes and Accident Descri tion Excessive feedwater additions are a means of increasing core power above full power. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The overpower and overtemperature protection (high neutron flux, overtemperature hT, and overpower hT trips) prevent any power increase that could lead to a DNBR that is less than the DNBR limit.
An example of excessive feedwater flow would be a full opening of a feedwater control valve due to a feedwater control system malfunction or an 'operator error. At power, this excess flow causes a greater load demand on the RCS due to increased subcooling in the steam generator. With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator coefficient of reactivity. Continuous excessive feedwater addition is prevented by the steam generator high-high level trip, which closes the feedwater valves.
15.2. 10.2 Anal sis of Effects and Conse uences The excessive heat removal due to a feedwater system malfunction transient is analyzed with the LOFTRAN code (3) . This code simulates a multiloop system, neutron kinetics, the pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety -valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
The system is analyzed to evaluate plant behavior in the event of a feedwater system malfunction.
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b Excessive feedwater addition due to a control system malfunction or operator error that allows a feedwater control valve to open fully is considered. Four cases are analyzed as follows:
(1) Accidental opening of one feedwater control valve with the reactor just critical at zero load conditions assuming a conservatively large moderator density coefficient characteristic of end-of-life conditions. Cases with and without automatic rod control are analyzed.
(2) Accidental opening of one feedwater control valve with the reactor at full power. Cases with and without automatic rod control're analyzed.
The reactivity insertion rate following a feedwater system malfunction is calculated with the following assumptions:
(1) For the feedwater control valve accident at full power, one feedwater control valve is assumed to malfunction, resulting in a step increase to 184% of nominal feedwater flow to one steam generator.
(2) For the feedwater control valve accident at zero load conditions, a feedwater valve malfunction occurs that results in a step increase in flow to one steam generator from zero to 120% the nominal full load value for one steam generator.
(3) For the zero load condition, feedwater enthalpy is at a conservatively low value of 50 Btu/ibm.
(4) No credit is taken for the heat capacity of the RCS and steam
'enerator thick metal in attenuating the resulting plant cooldown.
(5) Ho credit is taken for the heat capacity of the steam and water in the unaffected steam generators.
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i II (A
(6) The feedwater flow resulting from a fully open control valve is terminated by the steam generator high-high level signal that closes all feedwater control valves, closes all feedwater bypass valves, trips the main feedwater pumps, and shuts the feedwater isolation valves. The steam generator high-high level signal also produces a signal to trip the turbine.
15.2.10.3 Results In the case of an accidental full opening of one feedwater control valve with the reactor at zero power and the above mentioned assumptions, the maximum reactivity insertion rate .is less than the maximum reactivity insertion rate analyzed in Section 15.2.1, Uncontrolled RCCA Bank 'Withdrawal from a Subcritical Condition, and its analysis is, therefore, covered by that of the latter. It should be noted that if the incident occurs with the unit just critical at no-load, the reactor may be tripped by the power range high neutron flux trip (low setting) set at approximately 25% of nominal full power.
The full power case (end-of-life, without control) gives the largest reactivity feedback and results in the greatest power increase. Assuming the reactor to be in the automatic control mode results in a slightly less severe transient. The rod control system is not required to function for an excessive feedwater flow event. A turbine trip is actuated when the steam generator level reaches the high-high level setpoint. For convenience, reactor trip is assumed to be initiated upon turbine trip. However, this function is not necessary. Should turbine trip not initiate a reactor trip signal, reactor trip will occur on power range high neutron flux.
For all cases of excessive feedwater, continuous addition of cold feedwater is prevented by closure of all feedwater control valves, closure of all feedwater bypass valves, a trip of the feedwater pumps, and closures of the feedwater isolation valves on steam generator high-high level.
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T Transient results (see Figures 15.2. 10-1 and 15.2. 10-6) show the core heat flux, pressurizer pressure, T , and DNBR, as well as the increase in power and loop aT associated with the increased thermal load on the avg'uclear reactor. Steam generator level rises until the feedwater is terminated as a result of the high-high steam generator level trip. The ONBR does not drop below the limit safety analysis DNBR value.
15.2. 10.4 Conclusions The reactivity insertion rate that occurs at no-load following excessive feedwater addition is less than the maximum value considered in the analysis of the rod withdrawal from a subcritical condition. Also, the ONBRs encountered for excessive feedwater addition at power are well above the safety analysis limit ONBR value.
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15.2. 11 Excessive Load Increase Incident 15.2. 11. 1 Identification of Cause and Accident Oescri tion An excessive load increase incident is defined as a rapid increase in the steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The reactor control system is designed to accommodate a 10% step-load increase or a 5% per minute ramp load increase in the range of 15 to 100% of full power. Any loading rate in excess of these values may cause a reactor trip actuated by the reactor protection system.
This accident could result from either an administrative violation such as excessive loading by the operator or an equipment malfunction in the steam dump control or turbine speed control.
During power operation, steam dump to the condenser is controlled by reactor coolant condition signals; i.e., high reactor coolant temperature indicates a need for steam dump. A single controller. malfunction does not cause steam dump; an interlock is provided that blocks the opening of the valves unless a large turbine load decrease or a turbine trip has occurred.
Protection against an excessive load increase accident is provided by the following reactor protection system signals:
(1) Low Pressurizer Pressure, (2) Overtemperature hT, (3) Power Range High Neutron Flux.
15.2. 11.2 Anal sis of Effects and Conse uences This accident is analyzed using the LOFTRAN code (3)'. The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, feedwater system, steam generator, and steam generator safety valves. The code computes pertinent. plant variables including temperatures, pressures, and power level.
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-1 8 Four cases are analyzed to demonstrate the plant behavior following a 10% step load increase from rated load. These cases are as follows:
(1) Reactor control in manual with BOL minimum moderator reactivity feedback, (2) Reactor control in manual with EOL maximum moderator reactivity feedback, (3) Reactor control in automatic with BOL minimum moderator reactivity feedback, (4) Reactor control in automatic with EOL maximum moderator reactivity feedback.
For the BOL minimum moderator feedback cases, the core has the least negative moderator temperature coefficient of reactivity and the least negative Doppler
, only power coefficient curve; therefore the least inherent transient response capability. For the EOL maximum moderator feedback cases, the moderator temperature coefficient of reactivity has its highest absolute value and the most negative Doppler only power coefficient curve. This results in the largest amount of reactivity feedback due to changes in coolant temperature.
A conservative limit on the turbine valve opening is assumed, and all cases are studied without credit being taken for pressurizer heaters.
This accident is analyzed with the Improved Thermal Design Procedure as described in Reference 4. Initial reactor power, RCS pressure and temperature are assumed to be at their nominal values. Uncertainties in initial conditions are included in the limit DNBR as described in Reference 4.
Plant characteristics and initial conditions are further discussed in Section 15.1.
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'Hg ~
Normal reactor control systems and engineered safety systems are not required to function. The reactor protection system is assumed to .be operable; however, reactor trip is not encountered for most cases due to the error allowances assumed in the setpoints. No single active failure will prevent the reactor protection system from performing its intended function.
The cases which assume automatic rod control are analyzed to ensure that the worst case is presented. The automatic function is not required.
15.2.11.3 Results The calculated sequence of events for the excessive load increase incident are shown on Table 15.2-1.
Figures 15.2, 11-1 through 15.2. 11-4 illustrate the transient with the reactor in the manual control mode. As expected, for the BOL minimum moderator feedback case, there is a slight power increase, and the average core temperature shows a large decrease. This results in a DNBR which increases above its initial value. For the EOL maximum moderator feedback manually controlled case, there is a much larger increase in reactor power due to the moderator feedback.' reduction in DNBR is experienced but DNBR remains above the limit value.
Figures 15.2. 11-5 through 15.2. 11-8 illustrate the transient assuming the reactor is in the automatic control mode. Both the BOL minimum and EOL maximum moderator feedback cases show that core power increases, thereby reducing the rate of decrease in coolant average temperature and pressurizer pressure. For both of these cases, the minimum DNBR remains above the limit value.
For all cases, the plant rapidly reaches a stabilized condition at the higher power level. Normal plant operating procedures would then be followed to reduce power.
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The excessive load increase incident is an overpower transient for which the fuel temperatures will rise. Reactor trip does not occur for any of the cases analyzed, and the plant reaches a new equilibrium condition at a higher power level corresponding to the increase in steam flow.
Since DN8 does not occur at any time during the excessive load increase transients, the ability of the primary coolant to remove heat from the fuel rod is not reduced. Thus, the fuel cladding temperature does not rise significantly above its initial value during the transient.
15.2. 11. 4 Conclusions The analysis presented above shows that for a 10% step load increase, the DNHR remains above the safety analysis limit value, thereby precluding fuel or. clad damage. The plant reaches a stabilized condition rapidly, following the load increase.
1578v:1D/120588 15.2-'52
J 15.2. 12 Accidental Oe ressurization of the Reactor Coolant S stem 15.2. 12.1 Identification of Causes and Accident Oescri tion An accidental depressurization of the Reactor Coolant System could occur as a result of an inadvertent opening of a pressurizer relief or safety valve.
Since a safety valve is sized to relieve approximately twice the steam flowrate of a relief valve, and will therefore allow a much more rapid depressurization upon opening, the most severe core conditions resulting from an accidental depressurization of the RCS are associated with an inadvertent opening of a pressurizer safety valve. Initially, the event results in a rapidly decreasing RCS pressure until this pressure reaches a value corresponding to the hot leg saturation pressure. At that time, the pressure decrease is slowed considerably. The pressure continues to decrease, however, throughout the transient. The. effect of the pressure decrease would be to increase the neutron flux via the moderator density feedback. The reactor control system (if in the automatic mode) also functions to maintain the power and average coolant temperature essentially constant throughout the initial stage of the transient. Pressurizer level increases initially due to expansion caused by depressurization and then decreases following reactor trip.
The reactor will be tripped by one of the following reactor protection system signals:
(1) Pressurizer low pressure, (2) Overtemperature aT.
15.2.12.2 Anal sis of Effects and Conse uences The accidental depres'surization transient is analyzed with the LOFTRAN code (3) . The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator and steam generator safety valves. The code computes pertinent plant variables including temperatur'es, pressures and power level.
This accident is analyzed with the Improved Thermal Oesign Procedure as described in Reference 4.
1578 v:1D/120588 15.2-53
f f1 j
~
In calculating the DNBR, the following conservative assumptions are made.
(1) Plant characteristics and initial conditions are discussed in Section 15.1. Uncertainties and initial condi'tions are included in the limit DNBR as described in Reference 4.
(2) A positive moderator temperature coefficient of reactivity for BOL operation is used in order to provide a conservatively high amount of positive reactivity feedback due to changes in moderator temperature. The spatial effect of voids due to local or subcooled boi ling is not considered in the analysis with respect to reactivity feedback or core power shape. These voids would tend to flatten the core power distribution.
(3) A low (absolute value) Doppler coefficient of reactivity such that the resultant amount of negative feedback is conservatively low is used in order to maximize any power increase due to moderator reactivity feedback.
15.2. 12. 3 Results Figure 15.2.12-1 illustrates" the nuclear power transient following the RCS depressurization accident. The flux increases until the time reactor trip occurs on overtemperature aT, thus resulting in a rapid decrease in the nuclear flux. The time of reactor trip is shown in Table 15.2-1. The pressure decay transient following the accident is given on Figure 15.2. 12-2.
The resulting DN8R never goes below the safety analysis limit value as shown on Figure 15.2.12-3.
15.2. 12. 4 Conclusions The pressurizer low pressure and the overtemperature aT reactor protection system signals provide adequate protection against this accident and the minimum ONBR remains in excess of the safety analysis limit value.
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15.2.13 Inadvertent 0 enin of a Steam Generator Relief or Safet Valve 15.2.13. 1 Identification of Causes and Accident Oescri tion The most severe core conditions resulting from an accidental depressurization of the main steam system are associated with an inadvertent opening of a single 'steam dump, relief, or safety valve. The analyses, assuming a rupture of a main steam pipe, are discussed in Section 15.4.
The steam released as a consequence of this accident results in an initial increase in steam flow that decreases, during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin.
The analysis is performed to demonstrate that the following criterion is satisfied: assuming a stuck RCCA and a single failure in the engineered safety features (ESF) the limit ONBR value will be met after reactor trip for a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief, or safety valve.
The following systems provide the necessary mitigation of an accidental depressurization of the main steam system.
(1) Safety injection system (SIS) actuation from any of the following:
(a) Two-out-of-three low pressurizer pressure signals, (b) Two-out-of-three low steam line pressure signals in any one main steam line, (2) The overpower reactor trips (neutron flux and hT), low pressurizer pressure reactor trip, and the reactor trip occurring in conjunction with receipt of the safety injection signal.
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(3) Redundant isolation of the main feedwater lines: sustained high feedwater flow would cause additional cooldown. Therefore, in addition to the normal control action wh'ich will close the main feedwater valves following reactor trip, a safety injection signal will rapidly close all feedwater control valves and back up feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
.(4) Trip of the fast-acting main steam line isolation valves (designed to close in less than 5 seconds) on:
(a) Two-out-of-three high-2 containment pressure signals, (b) Safety injection system actuation derived from two-out-of-three low steam line pressure signals in any one main steam line (above Permissive P-ll),
(c) High negative steam pressure rate indication from two-out-of-three signals in any one main steam line (below Permissive P- 1) .
1 15.2.13.2 Anal sis of Effects and Conse uences The following analyses of a secondary system steam release are performed.
(j) A full plant digital simulation using 'o determine LOFTRAN((3) RCS temperature and pressure during cooldown, and the effect of safety injection.
(2) An analysis to ascertain that the reactor does not exceed the limit DNBR value.
The following conditions are assumed to exist at the time of a secondary system steam release accident.
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(1) EOL shutdown margin at no-load, equilibrium xenon conditions, and with the most reactive rod cluster control assembly (RCCA) stuck in its fully withdrawn position. Operation of RCCA banks during core burnup in accordance with Technical Specifications will ensure that the addition of positive reactivity in a secondary system steam release accident will not lead to a more adverse condi tion than the case analyzed.
(2) A negative moderator coefficient corresponding to the EOL. rodded
.core with the most reactive RCCA in the fully withdrawn position.
The variation of the coefficient with temperature and pressure is
'ncluded. The keff versus temperature curve at 1150 psia corresponding to the negative moderator temperature coefficient plus the Ooppler temperature effect used is shown on Figure 15.2. 13-1.
1 (3) Minimum capability for injection of high concentration boric corresponding to the most restrictive single failure in the acid'olution safety injection system. The injection curve is'shown on Figure 15.2.13-2. This corresponds to the flow delivered by one charging pump delivering its full contents to the cold leg header. No credit has been taken for the low concentration boric acid that must be swept from the safety injection lines downstream of the refueling water storage tank (RMST) isolation valves prior,to the delivery of high concentration boric acid (2000 ppm) to the reactor coolant loops.
(4) The case studied is an initial total steam flow of 268 lb/sec at 1200 psia with offsite power available. This is the maximum capacity of any single steam dump, power-operated reli'ef, or safety valve. Initial hot shutdown conditions at time zero are assumed since this represents the most conservative initial condition.
Should the reactor be just critical or operating at power at the time of a steam release, the reactor will be tripped by the normal overpower protection when the power level reaches a trip point.
1578v:1D/120588 15.2-57
Following a trip at power the RCS contains more stored energy than at no-load, the average coolant temperature is higher than at no-load, and there is appreciable energy stored in the fuel. Thus, the additional energy stored is removed via the cooldown caused by the steam release before the no-,load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition. at time zero. Since the initial steam generator water inventory is greatest at no-load, the magnitude and duration of the RCS cooldown are less for steam release occurring at power.
(5) In computing the steam flow, the Moody Curve for fL/0 = 0 is used.
(6) Perfect moisture separation in the steam generator is assumed.
15.2.13.3 Results The results presented are a conservative indication of the events that would occur assuming a secondary system steam release since it is postulated that all of the conditions described above occur simultaneously.
Figures 15. 2. 13-.3 and 15.2. 13-4 show the transient arising as the result of a steam release having an initial steam flow of 268 lb/sec at 1200 psia with steam release from one safety valve. The assumed steam release is the maximum capacity of any single'team dump, power-operated relief, or safety valve. In this case, safety injection is initiated automatically by low pressurizer pressure. Operation of one centrifugal charging pump is considered. Boron solution at 2000 ppm enters the RCS providing sufficient negative reactivity to prevent core damage. The reactivity transient for the case'hown on Figure 15.2. 13-4 is quite conservative with respect to cooldown since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Since the transient occurs over a period of about 5 minutes, the neglected stored energy is likely to have a significant effect in slowing the cooldown.
1578 v:1D/120588 15.2-58
15.2. 13.4 Conclusions The analysis has shown that the criteria stated earlier in this section are satisfied. For an accidental depressurization of the main steam system, the ONB design basis is met. This case is less limiting than the rupture of a main steam pipe case presented in Section 15.4.
1 578 v:1D/120588 15.2-59
t lqp
15.2.14 Inadventent 0 eration of the Emer enc Core Coolin S stem at Power 15.2.14.1 Identification of Causes and Accident Descri tion Inadverent ECCS operation at power, could be caused by operator error or a false electrical actuating signal. A spurious signal in any of the following channels could cause this accident.
(1) High containment pressure, (2) Low pressurizer pressure, (3) Low steamline pressure, (4) Manual actuation.
Following the actuation signal, the suction of the centrifugal charging pumps is diverted from the volume control tank to the refueling water storage tank (RMST). The charging pumps then force highly concentrated (2200 ppm) boric acid solution from the RMST through the header and injection line and into the cold legs of each loop. The passive injection system and the low-head system provide no flow at normal RCS pressure.
A safety injection system (SIS) signal normally results in a reactor trip followed by a turbine trip. However, it cannot be assumed that any single fault that actuates the SIS will also produce a reactor trij. Therefore, two different courses of events are considered.
Case A: Trip occurs at the same time spurious injection starts.
Case B: The reactor protection system produces a trip later in the transient.
1578v:1D/020389 15.2-60
For Case A, the operator should determine if the spurious signal was transient or steady state in nature, i.e., an occasional occurrence or a definite fault. The operator will determine this by following approved procedures. In the transient case, the operator would stop the safety injecti'on and bring the plant to the hot shutdown condition. If the SIS must be disabled for repair, boration should continue and the plant brought to cold shutdown.
For Case B, the reactor protection system does not produce an immediate trip and the reactor experiences a slight positive reactivity excursion, followed by a negative reactivity excursion due to the injected boron causing a decrease in the reactor power. At beginning of life, the power mismatch causes a drop in T and consequent coolant shrinkage, and, ultimately, pressurizer pressure and level drop. Load will decrease due to the effect of reduced steam pressure on load when the turbine throttle valve is fully open.
If automatic rod control is used, these effects will be lessened until the rods have moved out of the core. The transient is eventually terminated by the low pressurizer pressure reactor trip or by manual reactor trip.
Results at end of life are similar except that moderator feedback effect's result in a slower transient. The pressurizer pressure and level increase slowly and the coolant T decreases slowly. The transient is eventually avg terminated by the high pressurizer pressure, high pressurizer water level reactor trip, or by manual reactor trip.
The time to trip is affected by initial operating conditions including core burnup history that affects initial boron concentration, rate of cnange of boron concentration, and Doppler and moderator coefficients.
Recovery from this incident for Case B is in the same manner as for Case A.
The only difference is the lower T and pressure associated with the power avg imbalance during this transient. The time at which reactor trip occurs's of no concern for this occurrence. At lower loads; coolant contraction wi 11 be slower resulting in a longer time to trip.
1578v:1D/120588 15.2-61
15.2.14.2 Anal sis of E.'f=cts and Conse uences The inadvertent operation nf the ECCS system is analyzed with the LOFTRAN code (3) . The code simulates the neutron kinetics, RCS, pressurizer pressurizer relief and safety valves, pressurizer spray, steam generator steam generator safety valves, and the effect of the SIS ~ The program computes pertinent plant variables including temperatures, pressures, and power level.
Because of the power and temper..ature reduction during the transient, operating conditions do not exceed the core limits'nalyses of several cases show that the res'ults are relatively independent of time to trip A typi.cal transient is consicered representing conditions at HOL This accident is analyzed with the Improved Thermal Design Procedure as described in Reference 4 ~ The assumptions made in the analysis are (1) Initial Operating Conditions The initial reactor power, pressure and RCS temperatures are assumed to be at their aominal values. Uncertainties in initial conditions are included in the limit DNBR as described in Reference 4 (2) Moderator and Doppler Coefficients of Reactivity A positive BOL moderator temperature coefficient was us'ed. A low absolute value Doppler power coefficient was assumed (3) Reactor Control The reactor was assumed to be in manual control (4) Pressurizer Heaters Pressurizer heaters were assumed to be inoperative in order to increase the rate of pressure drop when it occurs 1578v:1D/1205ee 15.2-62
Jl+
Ml a l.
(5) Boron Injection At time zero, two charging pumps inject 2200 ppm borated water into the cold legs of each'oop after purging 0 ppm water from the SIS lines.
(6) Turbine Load Turbine load was assumed constant until the governor drives the throttle valve wide open. Then turbine load drops as steam pressure dl opso (7) Reactor Trip Reactor trip was initiated by low pressurizer pressure. The trip was conservatively assumed to be delayed until the pressure reached 1935 psia.
15.2. 14. 3 Results The transient response for the minimum feedback case is shown on Figures 15.2.14-1 through 15.2.14-2. Nuclear power increases and then starts decreasing rapidly due to boron injection, but steam flow does not decrease until 70 seconds into the transient when the turbine throttle valve goes wide open. The mismatch between load and nuclear power causes T , pressurizer level, and pressurizer pressure to drop. The low-pressure trip setpoint avg'ater is reached at 85 seconds and rods'tart moving into the core at 87 seconds.
15.2. 14.4 Conclusions Results of the analysis show that inadvertent ECCS actuation with or without immediate reactor trip presents no hazard to the integrity of the RCS.
The calculated DNBR is never less than the design limit. Thus, there will be no cladding damage an'd no release of fission products to the reactor coolant system.
1578v:10/120588 15.2-63
If the reactor does not trip immediately, the low-pressure reactor trip will be actuated for beginning of life, or a manual trip (or high pressurizer water level in the long term) for end of life. This trips the turbine and prevents excess cooldown thereby expediting recovery from the incident.
1578v:1D/120588 15.2-64
15.2. 15 References
1. Risher, D. H. Jr. and Barry, R. F,, "TWINKLE-A Multi-Dimensional Neutron Kinetics Computer Code," MCAP-7979-P-A (Proprietary) and MCAP-8028-A (Non-proprietary), January 1975.
2. H. G. Hargrove, FACTRAN - A Fortran IV Code for Thermal Transients in A UO2 Fuel Rod, MCAP-7908, June 1972.
3. T. W. T. Burnett, et al., LOFTRAN Code Descri tion, MCAP-7907-P-A (Proprietary), MCAP-7907-A (Non-proprietary), April 1984.
4. Chelemer, H., et al., "Improved Thermal Design Procedure," MCAP-8567 (Proprietary) and WCAP-8568 (Non-proprietary), July 1975.
5. Technical S ecifications, Shearon Harris Nuclear Power Plant Unit One Technical Specifications through Amendment 7, May 22, 1988.
6. Hochreiter, L. E., Chelemer, H. and Chu, P. T,, "THING-IV An Improved Program for Thermal-Hydraulic Analysis of Rod Bundle Cores," MCAP-7956, June, 1973.
7. M. A. Mangan, Over ressure Protection for Mestin house Pressurized Mater Reactor, WCAP-7769, October 1971.'.
Hochreiter, L. E., and Chelemer, H., "Application of the THINC-IV Program to PWR Design," WCAP-8054 (Proprietary), October, 1973, and MCAP-8195 (Non-Proprietary), September, 1973.
9. Morita, T., et. al., "Dropped Rod Hethodology for Negative Flux Rate Trip Plant," WCAP-10297-P-A (Proprietary) and WCAP-10298-A (Non-proprietary),
June 1983.
1578v:1D/120588 15.2-65
TABLE 15.2-1 Sheet 1 of .11 TIME SEQUENCE OF EVENTS FOR CONDITION II EVENTS Accident Event Time, sec Uncontrolled RCCA Initiation of uncontrolled Mithdrawal from a rod withdrawal 9.0 x 10 Subcritical hk/sec reactivity insertion Condition rate from 10 of nominal power 0.0 Power range high neutron flux low setpoint reached 8.8 Peak nuclear power occurs 8.9 Rods begin to fall into core 9.3 Peak heat flux occurs Peak hot spot average clad temperature occurs 11.8 Peak hot spot average fuel temperature occurs 12.1 1578v:1D/120588
TABLE 15.2-1 Sheet 2 of ll Accident Event Time, sec Uncontrolled RCCA h'ithdrawal at Power
1. Case A Initiation ef uncontrolled RCCA withdrawal at a high reactivity insertion rate (8.0 x 10 hk/sec) 0.0 Power range high neutron flux high trip setpoint reached 1.5 Rods begin to fall into 2.,0 core Minimum DNBR occurs 2.8
2. Case B Initiation of uncontrolled RCCA withdrawl at a small reactivity insertion rate (1.0 x 10 hk/sec) 0.0 Overtemperature aT 'reactor trip signal initiated 58.4 Rods begin to fall into core 60.4 Minimum DNBR occurs 60.6 1578v:1D/120588
TABLE 15.2-1 Sheet 3 of ll Accident Event Time, sec Uncontrolled Boron Dilution During Full Power Operation
a. Automatic reactor . Dilution begins control Shutdown margin lost 993
b. Manual reactor Dilution begins control Overtemperature hT reactor trip Rods begin to drop 47
/
Shutdown margin lost (if dilution continues after trip) 948 Partial Loss of Forced Reactor Coolant Flow: All loops operating, one pump coasting down Coastdown begins 0.0 Low-flow reactor trip 1.43 Rods begin to drop 2.43 Minimum DNBR occurs 3.2 1578v:1O/120588
TABLE 15.2-1 Sheet 4 of ll Accident Event Time, sec Startup of an Inactive Reactor Coolant Loop Initiation of pump startup 0.0 Power reaches high nuclear
'flux trip 3.6 Rods begin to drop .4. 1 Minimum DNBR occurs 4.6 Loss of External Electrical Load
1. With pressurizer control (BOL) Loss of electrical load 0.0 Initiation of steam release from steam generator safety valves 7.5 Overtemperature tT setpoint 9.7 reached Rods begin to drop 11.7 Minimum DNBR occurs 13.0 Peak pressurizer pressure 13.0 occurs 1578v:1D/120588
TABLE 15.2-1 Sheet 5 of 11 Accident Event Time, sec
2. Nith pressurizer control (EOL) Loss of electrical load ,0.0 Peak pressurizer pressure occurs 5.5 Initiation of steam release from steam generator safety valves 7.5 Low-low steam generator water 45.1 level setpoint reached Rods begin to drop 48. 6 Minimum ONBR occurs (a)
3. Without pressurizer control (BOL) Loss of electrical load 0.0 High pressurizer pressure reactor trip setpoint reached 3.9 Rods begin to drop 5.9 Peak pressurizer pressure occurs 7.5 1578v:1D/1205SS
TABLE 15.2-1 Sheet 6 of ll Accident Event Time, sec initiation of steam release from steam generator safety valves 9.0 Minimum DNBR occurs (a)
4. Without pressurizer control (EOL) Loss of electrical load 0.0 High pressurizer pressure reactor trip setpoint reached 3.9 Rods begin to drop 5.9 Peak pressurizer pressure occurs 6.'5 Initiation of steam release from steam generator safety valves 9.5 Minimum DNBR occurs (a) 1578v:1D/120588
TABLE 15.2-1 Sheet 7 of 11 Accident Event Time, sec Loss of Normal w/Power w/o Power Feedwater and Loss of Offsite Power to the Station Auxiliaries (Station Blackout) Main feedwater flow stops 10 10 Low-low steam generator water level reactor trip 54.3 54.6 Rods begin to drop 57.8 58.1 Reactor coolant pumps begin to coast down 60.1
\
Pe'ak water level in pres-surizer occurs 2288 62 Steam generators begin to receive auxiliary feed-water from one motor-driven AFW pump -117 -117 Cold auxiliary feedwater is delivered to the steam generators -204 -175 Core decay heat plus pump heat decreases to auxiliary feedwater heat removal capacity -2400 -350 1578v:10/120588
TABLE 15.2-1 Sheet 8 of ll Accident Event Time, sec Excessive Feedwater w/Rod Control w/o Rod Control Flow at Full Load One main feedwater control valve fails fully open High-high steam generator level signal generated 54.1 54.2 Turbine trip occurs due to high-high steam generator level 56.6 56.7 Minipum DNBR occurs 57.0 57.0 Reactor trip due to turbine trip rod motion begins 59.1 59.2 Feedwater isolation valves fully closed 64.1 64.2 1578v:1D/120588
TABLE 15.2-1 Sheet 9 of 11 Accident Event Time, sec Excessive Load Increase
1. Manual reactor control (BOL minimum moderator feedback) 10% step .load increase 0.0 Equi 1 ibrium condi tions reached (approximate times only) 200
2. Manual reactor control (EOL maximum moderator feedback) 10% step load increase 0.0-Equilibrium conditions reached (approximate times only) 150
3. Automatic reactor control (BOL minimum moderator feedback) 10% step load increase 0.0 Equilibrium conditions reached (approximate times only) 150 1578v:10/120588
TABLE 15.2-1 Sheet 10 of 11 Accident Event Time, sec
4. Automatic reactor control (EOL maximum moderator feedback) 10% step load increase 0.0 Equi librium conditions reached (approximate times only) 100 Accidental Oepressuri-zation of the Reactor Coolant System .Inadvertent opening of one RCS safety valve 0.0 Overtemperature hT Trip Setpoint Reached 25.8 Rods begin to drop 27.8 Minimum DNBR occurs 28.2 Accidental Depressuri-zation of the Main Steam System Inadvertent opening of one main steam safety or relief valve 0.0 Pressurizer empties 172 Boron from the RHST reaches RCS loops 257 15 78v:10/120588
TABLE 15.2-1 Sheet ll of 11 r
Accident Event Time, sec Inadvertent Operation of ECCS During Power Operation Charging pumps begin injecting borated water 0.0 Minimum DNBR occurs 7.5 Low-pressure trip setpoint reached 85 Rods begin to drop 87 (a) DNBR does not decrease below the initial value.
(b) Not a required safety function.
1578v:1D/120588
IG 2
4. 6. BE IQ. 12. I<. 16. IB. 28. 22. 2< ~ 26 'B ~ 56 ~
T ICE I BEC I ISA 15. 20. 25.
T ICE (SEC I Shearon Harris Figure 15.2.1-1 Uncontrolled Rod Withdrawal From A Subcritical Condition Nuclear Power and Core Heat Flux vs. Time
P
\ E (coo ~
C
.=,:,.-.:
I I
~ L 'L ~
~
) )ecp..
~ )-"on.
2 12OO.
)OCO.
500.
5. 15. 29. 25. 30.
T I ME ( SEC )
C
~ ~
I eeo.
5. 15 ~ 28. 25.
TIME (SEC )
Shearon Harris Figure 15.2.1-2 Uncontrolled Rod withdrawal From A Subcritical Condition Hot Spot Fuel Average and Clad Temperature vs. Time
2688.
2588.
U) 2488.
2388.
2288.
K N 2188.
K g 2888.
1988.
1B88 8, l. 2. 3. 4. S. 6. 7. B. 9. 18.
1.6 z
0 1.4 z
0 1.2 z0 I-V CC
~ B CC
.6 0
K .4 O ~ 2 D
z
8. l. 2. 3.= 4. S. 6. 7. B. 9. 18.
TIME (SEC)
Shearon Harris Figure 15.2.2-1 Uncontrolled Rod Withdrawal From 100/ Power Terminated by High Neutron Flux Trip Pressurizer Pressure and Nuclear Power vs. Time
S.'.5 3.5 K
z lD 3.
2.5 2.
1.5
8. 2. 3. 4. 5. 6. 7. B. 9. 18.
628.
o 618.
g 688.
LL a- 598.
I-588.
U K
578.
o~ 568 U
558
8. 1. 2. 3. 4. 5. 6. 7. 8. I.. 18.
TiME (SEC I Shearon Harris Figure 15.2.2-2 Uncontrolled Rod Withdrawal From 100% Power Terminated by High Neutron Flux Trip DNBR and Tavg vs. Time
2688.
2588.
2488.
K 2388.
a. 2288.
K.,
21ee.
K D
2888.
1988.
1888 B. S. 18. 15. 28. 25. 38. 3S. 48. <S. 58. 55. 68. 65. 78.
1.6 R
0 z 1.2 0
0 O
.8 K
K .6 0 .4
~ 2 O
D R
8. S. 18. 15. 28. 25. 38. 35. 48. 45. 58. 55. 68. 65. 78.
TIME tSECI Shearon Harris Figure 15.2.2-3 Uncontrolled Rod Withdrawal From 100% Power Terminated by Overtemperature Delta-T Pressurizer Pressure and Nuclear Power vs. Time
S.
4.5 3.5 K
3.
2.5 2.
1.5 l.
8. 5. 18. 15. 28. 25. 38. 35. 48. 45. 58. SS. 68. 6S. 78.
628.
618.
688.
K 598.
I-588.
K 578; K
O 568.
U SSB.
8. S. 18. IS. 28. 25. 38. 35. 48. 45. 58. 55. 68. 65. 78.
TIIHE (SEC)
Shearon Harris Figure 15.2.2-4 Uncontrolled Rod Withdrawal From 100/. Power Terminated by Overtemperature Delta-T.
DNBR and Tavg vs. Time
2A 2.3 MAXIMUM FEEDBACK MINIMUM FEEDBACK a 2.1 HIGH NEUTRON Z
1 OTbT Ft VX TRIP 2.0 TRIP X
i.9 I
I I
%I I V
1.6 0.1 X.O iO. 0 joo. 0 4000. 0 mcmtn wsami vz t ethic)
Shearon Harris Figure 15.2.2-5 Effect of Reactivity Insertion Rate on Hinimum DNBR For a Rod Withdrawal Accident at 10% Power
MAXIMUM FEEDBACK MINIMUM FEEDBACK 2.3 OTBT TRIP a 2.1 X
A 2.O z
I.S HIGH NEUTRON FLUX TRIP 1.6 f.5 O.i 1.0 80. 0 100. 0 1000. 0 xAcmmY mme vm (eavxc)
Shearon Harris Figure 15.2.2-6 Effect of Reactivity Insertion Rate on Hinimue DNBR For a Rod Mithdrawal Accident at 6K Power
- - MAXIMUM FEEDBACK HINIMUM FEEDBACK HIGH NEUTRON/
FLUX TRIP
]
I
/
OTclT TRIP 8.0 10. 0 100. 0 .000.0 mCnnn SSDIIi ATE 0 @VS)
Shearon Harris Figure 15.2.2-7 Effect of Reactivity Insertion Rate on Minimum DNBR For a Rod Mithdrawal Accident at 1& Power
1 ~ 2000
1. 1000
l. 0000
.90000
.e0000
.70000
.600DO
.50000 o
o oCI C)
CI CI CI CI CI o
Vl oo CI Vl IQ TlHE (SEC) 1.2000
1. 10DO.
1. 0000
. 900QO
. 80000
.70000
.60000
.50000 o oo oo CI o
o CI oo o lA CI o
ci IA TtHE (sEC)
Shearon Harris Figure 15.2.3-1 Transient Response To A Dropped RCCA Nuclear Power and Heat Flux vs. Time
COO.OO LLI 580.00 5CO. 00 I us LLI Sio.oo CC O IV O Gl 52D. 00 CC 500. 00 8
C7 ei 8
fv TlXE 1SEC) 2i00. d 2300. 0 22DO. 0 2100. 0 2000 0 1900. 0 1800. 0 aCI C7 O Vl CI Tlef 1SEC)
Shearon Harris Figure 15.2.3-2 Transient Response To A Dropped RCCA Tavg and Pressurizer Press. vs. Time
l.2 o 1 ~
4
.6 C)
La EA IA
~ 2 8 ~
8. l. 2. 3. 4. 5. 6. 7. 8. cl. 18.
.6 3
C)
~ 2 8.
8. 1. 2. 3. 4. 5. 6 ~ 7. 8. Cl. 18.
TINE (SEC)
Shearon Harris Figure 15.2.5-1 All Loops Operating One Loop Coasting Down Vessel Flow and Faulted Loop Flow vs. Time
.6
.6 CD O
~ 4 4J
~ 2
8. I. 2. 3, 4. 5. 6. 7. 8. 9. 18.
I ~ 2 x
CD I ~
CD LL I
cC
.2 8.
8. I ~ 2. 3. 4. 5. 6. 7. 8. 9. 18.
TINE (SEC)
Shearon Harris Figure 15.2.5-2 All Loops Operating One Loop Coasting Down Nuclear Power and Heat Flux vs. Time
L4 1.
Time (sec) 2688.
2588.
O.
2488.
2388.
s 2288.
N P
2188.
2888.
I cIBB.
B. I . 2. 3. 4. S. 6. 7. 8. 'I. I e.
Time (sec)
Shearon 'Harris Figure 15.2.5-3 All Loops Operating One Loop Coasting Oown Pressurizer Pressure and ONBR vs. Time
1.4 z
O 1.2 z
0 zO I-V .8 5
,4 D
z
5. 18. 15. 28. 25, 38.,
T1ME (SECONDS)
Shearon Harris Figure 15.2.6-1 Startup of an Inactive Loop Nuclear Power vs.Time
1.4 2
X 1 ~
,i L'.. ~
R Lu
+0 W Z Lu zC Z0 X0 Lu U U cC Q K u
LLJ 0 .2 C
5. 18. 15. 28 ~ 25.
TIME (SECONDS) 1 ~ 2 X~
u- Z C0 .e x
~o 2g ZO
.6
~ IV V
I g
~,4 0 u.
.2
5. 18. 15, 28. 25. 38.
TIME (SECONDS)
Shearon Harris Figure 15.2.6-2 Startup of an Inactive Loop Average and Hot Channel Heat Flux vs.Time
2488.
2358.
2388.
2258.
g 2288.
2I58.
2I88.
2658 S. 18. IS. =
28, 25.
TIME (SECONDS) 648.
I 628.
688.
588.
U 568.
0 548.
528
'88.
8. S. I8. IS. 28. 25. 38.
TIME tSECONDS)
Shearon Harris Figure 15.2.6-3 Startup of an Inactive Loop Pressurizer Press.
and Core Tavg vs.Time
2.2 2.
'.8 I-1.6 0 1.4 z0 1 ~ 2 I
V I.
CC
.8 0 .6 ill . 4 K
O S. 18. 15 ~ 28. 25. 38.
TIME (SECONDS) 2' 2.6 2 '
2.2 D 2.
1.8 1.6 1.4 1.2 I ~ 2 3. 4 ~ 5. 6. 8. 9. "
18.
TIME (SECONDS)
Shearon Harris Figure 15.2.6-4 Startup of an Inactive Loop Core F1ow and DNBR vs.Time
I.4 E I.2 0 76. IS. IGG.
~
C. IS. 20. 36. 46. 56. 66. BG.
T IHE ISEC )
C I 5 I.
8. IS. 28. .38. 48. 58. BS ~ 78. BG. lb. ISS.
TIHE ISECJ Shearon Harris Figure 15.2.7-1 Loss of Load w/ Pressurizer Spray and PORVs at BOL Nuclear Power and ONBR vs. Time
2603.
c 2=03.
C~.'.
2403.
Lj (7) 2=03.
~ 22QD.
LJ tV
~ 21 QQ.
~~ 2833.
IROD.
Q. IC. 28 '8. 48. 58 T1M:
'0.15EC 73.
I
60. RG. IED.
14DC.
IEDC.
1200.
D C7 IIQC.
I 5 IQDC.
0'OD.
Q &QD.
780.
6GD.
G. 10. 28. 38. 48. 58 TIME
'8.ISEC)
78. 80. 98, IGD.
Shearon Harris Figure 15.2.7-2 Loss of Load w/ Pressurizer Spray and PORVs at BOL Pressurizer Pressure and Water Volume vs. Time
6'o.
630.
I
~ 620.
a 6IO.
m 688.
I
~ 598.
56".
)C
~~
~ 578 ~
CI 56Q ~
ceo 548.
o. IQ. 20. 30. 48. 50. 68. 78. 88. 98. 188.
TINE I SEC I 608 ~
a90 568 70, cr 568.
g W
55p I
~ 548.
C? 8 Q
520.
5IQ.
80
8. IQ. 28 '8, 48. 58.
TINE 68.
(SECI
78. 88. 90 'OO.
Shearon Harris Figure 15.2.7-3 Loss of Load w/ Pressurizer Spray and PORVs at BOL Core Tavg and Steam Temperature vs. Time
1.4 E
Z l.2 CD l.
CD I
cc .8
.6 CD 0:
.4
<Z LJ
~ 2
8. )G. 28. 38. 48. 50.
TIHE 68.
(5EC)
70. 80. 98 '08 CD 3 c (Zl Z
~ ~
CD 2.5 I 5 I ~
8. lG. 28. 38. 48. 58. 68. 78. 88. 98. IGG.
I TINE ) SEC )
Shearon Harris Figure 15.2.7-4 Loss of Load w/ Pressurizer Spray and PORVs at EOL Nuclear Power and DNBR vs. Time
2600.
2500 ~
th G
2400.
LJ g 2300.
0' 2280.
LJ M
~ 2100.
g 20M.
l9CQ.
1980 8 ~ 18. 28. 38. 48. 58. 68. 78. 60. 98. IGD.
TINE (SEC I
'1400.
IA 1300
'200.
C)
I 100.
$ 1000.
900.
w~ BOO 700.
600.
8. 10. 28. 38. 48. 58. 68. 78. 80. 98. 188.
T IHE I SEC I Shearon Harris Figure 15.2.7-5 Loss of Load w/ Pressurizer Spray and PORVs at EOL Pressurizer Pressure and Water Volume vs. Time
640.
650.
~ 628.
K 618.
z 600.
I
~ 598.
~w SBG.
~ ~ 0.
~ 568.
58 c40
'. 10. 20. 58. 40.'G. 68. 78. eG. 98 'cos TINE (SEC )
603.
-9n L 568.
578o Q.'z 568.
G:
4J
< 558 4J I
~ S48.
I cx'8 520, 18 588.
8. 18. 28. 30. 48. 58. 68. 78. 88. 90. 180.
TINE 1 SEC)
Shearon Harris Figure 15.2.7-6 Loss of Load w/ Pressurizer Spray and PORVs at EOL Core Tavg and Steam Temperature vs. Time
~ 6 LJ Cl 0
4 OC LJ O
Z
8. IG ~ 20. 38. 48.
TINE
58. 68 '8.
(SEC!
80. RG. 100.
4.5 O
~ 3.5 CC GC C) 2 C) 2' 1.5
8. 'l8, 28. 38. 48. 58. 68. 78. 80. '18. 108.
T I HE I SEC 1 Shearon Harris Figure 15.2.7-7 Loss of Load w/o Pressurizer Spray and PORVs at BOL Nuclear Power and DNBR vs. Time
1$
J MI
2600.
a: 2580.
Q 2480.
Pn2 80 G.
2288.
LJ tV
~ 2100.
D
~~ 2880 ~
I tQQ.
I6CQ.
8. IQ. 28. 38. 48. 58. 68. 78. 68. 98 'QQ.
TIME (SEC)
I 100.
QD.
Y I280.
5 ID QP I l80.
I g IQQD.
LJ 980.
Qt 880.
QI 788.
680 '. IQ. 28. 38. 48. 58. 68. 78. BQ. 98. I88.
TlME (SECI Shearon Harris Figure 15.2.7-8 Loss of Load w/o Pressurizer Spray and PORYs at BOL Pressurizer Pressure and Water Volume vs. Time
t
- 650.
w 620.
K 610.
~ 600.
w 590.
~ 560.
w S70.
w 560.,
550.
54Q.
G. IO. 28. =C. ~0. 58. 68. 7C. BG. 98. ICO.
T I HE I SEC I 600.
598.
L 5BG.
w 570.
<<z" 568.
w
~ 558.
4J I
~ 548.
~ 530.
528.
518.
508.
8. 18. 28. 38. 48. 58.
TIIIE 68.
(SECI
78. 88 '8. IGG ~
Shearon Harris Figure 15.2.7-9 Loss of Load w/o Pressurizer Spray and PORVs at BOL Core Tavg and Steam Temperature vs. Time
).4 CD 1.2 CD z
CD I
~ B Q
4
.6 CD
.4 D'Z
.2 R
8.
8. IG. 28. 38. 48. 58. 68. 78. BG. 98. IGG.
TIIIE (SEC) 4,c 4,
CD I ~
c C
DC 2 ~ S 2 ~
I c I.
8. IG ~ 28. 38. 48. SG. 68. 78. BG ~ 98. 188.
T IIIE I SEC I Shearon Harris.
Figure 15.2.7-10 Loss of Load w/o Pressurizer Spray and PORVs at EOL Nuclear Power and DNBR vs. Time
'l 4l 0
2680.
a 2588.
Us 2400.
Ld Q.
0'~08 ~
2280.
0 LI fV
~ 2100.
th C/l
~~ 2800.
CL 1980.
8. 18. 20. SC. 48. 58. 68. 78. B8. 90. 180.
T I IIE I SEC I 1408.
rn 1300.
1208.
ID Qt I I 80.
g 180'80.
608.
G. 18. 28. 38. 48. 58..68. 78. BG. 90. 188.
TIIIE ISEC)
Shearon Harris Figure 15.2.7-ll Loss of Load w/0 Pressurizer Spray and PORVs at EOL Pressurizer Pressure and Water Volume vs. Time
tq ~
0
640.
638.
w 628.
cz 618.
m 608.
w I
w 598.
~~ 568.
578.
o w 568.
550.
c48
8. IG. 28. 30. 48. 58. 68. 78. BG. 90. 180.
T I HE < SEC I Pt 60 598.
~ 5BG.
w 578 '
568.
w 5 0.
w I
~ 548.
~ cue Q) 520
'18
~
8. 18. 28. 38. 48. 58. 68. 78. BG. 90. 180.
T111E I SECI Shearon Harris Figure 15.2.7-12 .
Loss of Load w/o Pressurizer Spray and PORVs at EOL Core Tavg and Steam Temperature vs. Time
U - l dl ' I P 0
R I.2 C)
I.
G.
Q:
.6 Q1 CC LJ
.4 z
~ 2 I88 Iol l82 I8S TIIIE (SEC) 1.4 o I.2 C)
I.
.6 K
w,4
~ 2 Io'82 TIIIE (SEC)
I84 Shearon Harris Figure 15.2.8-1 Loss of Normal Feedwater Nuclear Power and Core Heat Flux vs. Time
ll rs ~ y+ " "~ '
the s
c- 660.
n. 640.
a: 620.
~
I 608.
SBQ.
CL
~S68.
~
S28.
SQQ ~
181 182 184 TIME ( SEC 1
! 6CE.
c~
C,
'CG c 1208.
1888.
888.
c 688.
<z 4QQ.
'81 I
c')
288 182 184 TItlE (SEC)
Shearon Harris Figure 15.2.8;2 Loss of Normal Feedwater Primary Temperature and Steam Generator Pressure vs. Time
~ s ~ ~ o. r y, ~
2680.
~ 2600 '
> 2488.
~ 2280.
fV Pn 2888.
s G:
1688.
1608.
I 00 181 182 18" 184 T IIIE I SEC I 20"D.
In IPGG.
16CG.
3C) 1408.
a: 12QQ.
g 1888.
G.'
608.
688.
488.
208.
)QQ IQI 182 184 T IIIE I SEC I Shearon Harris Figure 15.2.8-3 Loss of Normal Feedwater Pressurizer Pressure and Mater Volume vs. Time
f I ' A ~. J>> '%l "V ~ ~ I ~
M
( 0
.2 182 TINE (SEC)
.2 182 104 TINE I SEC )
Shearon Harris Figure 15.2.9-1 Station Blackout Nuclear Power and Core Heat Flux vs. Time
>ft 700.
~ 660, Y 640.
I
= 620.
~ 600.
~60.
8~ 560.
548.
520.
580 101 182 184 AT]HE ISEEI 1600.
CZ v>
C 1400.
a. 12DD.
1888.
2 880.
w~ 688.
488.
280.
181 ]82 ]85 184 T]HE (SECI Shearon Harris Figure 15.2.9-2 Station Blackout Primary Temperature and Steam Generator Pressure vs. Time
4 4 ~%
g
2":OG.
~ 2408.
tL g., 2208
'V R 2008.
1888.
leGG 180 181 162 LBS 184 TINE 'SEC) 2000.
~~ 1&GO.
1 EGO.
aCI l 400.
a. 1200.
LJ I
1008.
~P 888.
668.
488.
288.
8.
18 181 162 16""
TINE 1 SEC )
Shearon Harris Figure 15.2.9-3 Station Blackout Pressurizer Pressure and Water Volume vs. Time
l4 f a~
~ ~
I 0 1.2 0.
G. 28. 40. 68. 88.. 100. 128. 148. 160. 190. 208.
TI))E (SEC) 1.2 I
.6
.4
.G. 28. 48. 68. 98. 180. 128. 148 '68. IBG. 280.
TI))E (SEC)
Shearon Harris Figure 15.2.10-1 Feedwater System Mal function with Rod Control>Nuclear Power and Core Heat Flux versus Time
'C lC I
g A +W
C4e1 (lI c>> i.'3 G C '
Vl C1 4f C 4>> ~
Q.
G
~
LJ 22SQ.
~ 2203.
4')
>> C ~ C>>'
~
f 2)83.
23S3.
20 ~ 48. 60. 88. 108. 128. 140. ) 60. I &8. 283.
TINE !SEC) m SOS x
D 2.S 2 ~
I.S
8. 28. 48. 68 ~ 88. ! 88. 128. 148. 160 ~ 180. 208.
TINE ! SEC)
Shearon Harris Figure 15.2.10-2 Feedwater System Malfunction with Rod Control-Pressurizer Pressure and ONBR vs. Time
~ 7i ~
t e 6" LJ C-Q LJ Cga Q h.'
4wC ~
I 0
C
0. 20. 48. 68.. BG. 188. 128. 148. 168. 1BG. 208.
71HE ( SEC )
Vl CB G.
o 580.
< C7Cw ~
<Z I
w c78 G:
CD C6C 568.
cc5 558.
28. 48. 68. BG. 188. 128. 148. 168, 1BG. 208.
T lHE 1 SEC 1 Shearon Harris Figure 15.2. 10-3 Feedwater System Halfunction with Rod Control-Loop Delta-T and Core Tavg versus Time
N s t.-
PA %
,I I~
1.2 0.
Q. 20. 48. 60. BQ ~ 180. 128 ~ 1 48. I60. 168. 280.
TINE I SEC I I 2 P
'Z t
'J 0
C)
CJ I 4
~ 2 8.
8. 28. 48. 68. BQ. 188. 128 ~ I.40. 168. I BQ. 200.
T I IIE ( SEC )
Shearon Harris Figure 15.2.10-4 Feedwater System Malfunction without Rod Control-Nuclear Power and Core Heat Flux versus Time
2v50e
~
0
~~ 2550.
I:4 ~
C
~ 2250.
lJ f'4
~ 2200.
h C:)
Q 2150:
2100.
2058.
2080.
0. 28. 48. 68. 80. 188. 120. I48. 168. I80. 280.
71IIE ISECI C
~zc Cl z
Ci 2.5 2 ~
1.5
8. 28. 48. 68. 88. 188. 128. ) 40. 169. 168. 208.
7 I IIE I SEC )
Shearon Harris Figure 15.2.10-5 Feedwater System Mal function without Rod Control-Pressurizer Pressure and ONBR versus Time
P 0 L' G.
~
c en~ ~
I 40.
I LJ c =0 0
18.
0. 28. 48. 60. 88. 188. 128. 148. 160 ~ 183. 280.
T I ME I SEC )
1J) 4J Ce LJ 4 a~/~I c c70 565 560.
555.
50 ~
28. 48. 68 ~ 88. 188. 128. 148. 168. 168 ~ 288.
T I ME ( SEC 1 Shearon Harris figure 15.2.10-6 Feedwater System Malfunction without Rod Control-Loop Delta-T and Core Tavg versus Time
i~
53. 10". )50. 203. 250 'OO. *58. 400 (S:C) '!H.:
a r .g, c4.D.
a. ~
g 22DD.
~ 2)DD. j
"" 2COD.
~q )ROD.
~ )608.
8. 50. )88. )58. 288 ' 258. 308 '58. 488.
I )1E )SEC)
Shearon Harris Figure 15.2.11-1 Excessive Load Increase w/0 Control, Minimum Feedback Nuclear Power and Pressurizer Pressure vs.Time
II 4
UA
ccG G. 50 'GO. ISG. 288. 258. 388. 358. 480 TlME (SECI 2 2
~
2.
l .75
58. 188. 158. 288. 258. 388. 358. 488.
T1HE ( SEC I Shearon Harris Figure 15.2. 11-2 Excessive Load Increase w/o Control, Minimum Feedback Tavg and DNBR vs. Time
1.2 CL 4J
0. co ~ 180. 150. 2CD. 250. 300. 350. 480.
TINE (5EC)
CZ P4 Vl C @~a. ~ ~
2= CD ~
'C'7
(,J CCI.I.'J C.
~ 2)CD.
C 4J N (L~I. ~
O.
(~ il 'U ~
LJ 0'EDD.
8. 50. 100. )58. 288. 250. 300. 350. 48D ~
T I))E )SEC )
Shearon Harris Figure 15.2.11-3 Excessive Load Increase w/0 Control, Maximum Feedback Nuclear Power and Pressurizer Pressure vs. Time
IS q >ki ~ ~ %
s ~ ql I tll
'%0'A I'
P
I c2C ~
l
~t 5)0.,
LJ Q.
G.
4J
~ 5~0.
~
C 560.
a.
~~ 570.
550
8. 50. 180. ISB, 208.
TIHE 258.
(sEc) 3BB. 350 '80.
c ~ i 2
2 2 C)
D 1 ~ 75
1. 25 1 ~
8 . 58. 188. 158. 288. 258. 388. 358. 488.
TINE (SEC)
Shearon Harris Figure 15.2.11-4 Excessive Load Increase w/0 Control, Maximum Feedback Tavg and DNBR vs. Time
Vga X
O z ).2 I
l. 15 o I. I C.'.
05 O.'
I.
9 C.
LJ ~ 9 oD Q. 50. 100. )50. 200. 258. 308. 350. 480.
Tll;E )SEC)
CC pl Vl c~L, a o 2500.
22Cc e L
~ 2100.
"" 2008.
C'~
~I %08.
O.'600.
8. 58. IGG. )58. 288. 258. 380 '58. 488 T IIIE I SEC')
Shearon Harris Figure 15.2.11-5 Excessive Load Increase w/
Control, Minimum Feedback Nuclear Power and Pressurizer Pressure vs. Time
Sco fl V ~
58. 188. 158. 288. 2S8 ~ 300. 350. 488.
TINE (SEC) 2 I
~
C 2 2
~
0
1. 75 C
1.25 I.
8. 58. 180. 158. 288. 258. 388. 35'88.
TjME (SEC)
Shearon Harris Figure 15.2.11-6 Excessive Load Increase w/
Control, Hinimum Feedback Tavg and DNBR vs. Time
1.2 C.
C.'i CJ 6. SQ. 190. 156. 286. 258. 366. 35'66.
TIVE (SEC) 2 Ds ~
G g @cod 4'a.
~ 210.".
G.
fg ft&tt Ciiv ~
G:
(1 198.
0 1680.
58. 188. 156. 288. 258. 368. 359. 488.
TINE l SEC 1 Shearon Harris Figure 15.2. 11-7 Excessive Load Increase w/
Control, Haximum Feedback Nuclear Power and Pressurizer Pressure vs. Time
WSf
~,k ~ ~ ~
st t
0
Q 20 ~
610 ~
LJ G:
600.-
C'J 5g0
~) 570.
550
8. 58. 188 ~ 158. 288. 258. 388. 358 ~ 488.
T I NE ( SEC1 C
C 2o2 D
1.75 1 ~ 25
8. 58. 188.. 158. 288. 258. 388. 358. 408.
TINE (SEC)
Shearon Harris Figure 15.2.11-8 Excessive Load Increase w/
Control, Maximum Feedback Tavg and ONBR vs. Time
4 CZ z'D 1.2 4
CD I.
CD I
CC .8 Of 4.
.6 CD O.
,4 a
LI z ~ 2 628.
~
4J 618.
CD 688.
LJ 5'18.
Cf
o. 588.
578.
4J g4J 568.
) 558.
Ld K
O o 548.
538.
528.
8. S. 18. 15. 28. 25. 38. 35. 48. 45. 58. 55. 68; T INE I SEC )
Shearon Harris Figure 15.2.12-1 Accidental Depressurization of the Reactor Coolant System Nuclear Power and Core Tavg vs. Time
~ r a a ~ >a'i I' 2498.
2288.
Ch CL w 2888, ac 1888.
1688.
w 1488.
1288.
1888.
1288,
~ 1188.
1888.
3 e88.
)
OC 888.
4J I
CL Oa 788.
~w 688.
g 588.
488
'88.
8. 5. 18. 15. 28. 25. 38. 35. 48. 45. 58. 55. 68.
T1HE l SEC1 Shearon Harris Figure 15.2.12-2 Accidental Depressurization of the Reactor Coolant System Pressurizer Pressure and Mater Volume vs. Time
( ll 5.
4.5 4,
3.5 Q
2.5 1.5 I.
8. S. 18. 15. 28. 25. 38 TIME
'5. 48. 45. 58. SS. 68 (SEC)
Shearon Harris Figure 15.2.12-3 Accidental Depressurization of the Reactor Coolant System DNBR vs. Time
'k ZERO POWER. j150 PSIA END OF LIFE RODDED CORE h'I TH ONE RC CA STUCK FULL DUT Shearon Harris Figure 15.2.13-1 Hain Steam Depressurization Variation of Keff with Core Temperature
~
y
'el
~ '
i L
lN
%J lm g )4$
a L
12M 0
a 100Ij mm szcw mw(+
Shearon Harris Figure 15.2.13-2 Nain Steam Depressurization Safety injection Flowrate
P
~R )WAN'
" f~*L lk'
~ 2588.
LO 2888.
P) ( 88.
)QBB.
588.
188. 288. 588. 488. 588. 688. 788.
TINE ( SEC )
688.
LJ D
~ 588.
w 488
)
<Z m 388.
C) 288.
]88. 288. 588. 488. 588. 688. 788.
TINE (SEC)
Shearon Harris Figur e 15.2.13-3 Transient Response for a Steam Line Break Equivalent to 268 lb/sec at 1200 psia w/ Offsite Power Available
588.
p 488.
388.
Pz) 288.
C) 188.
-188.
8. 188. 288. 388. 488. 588. 688. 788.
TIME ( SEC )
3888.
2888.
]888.
I cc LJ
-1888.
-2888.
-3888.
8. 188. 288. 388. 488. 588. 688. 788.
T I(1E ( SEC )
Shearon Harris Figure 15.2.13-4 Transient Response for a Steam Line Break Equivalent to 268 lb/sec at 1200 psia '
w/ Offsite Power Available
p
1 ~ 4 I 1.2 M cC UJ X D X Q C)
,8 Cd cC 1L 4J O Cl O M CY ,4
,2 1.2 cC AOX C) .8 4 X X
CO LL I V 4A cC ,4 CC
.2
$ 2 658.
Lsl CC 625.
I 688.
UJ LL, 575.
UJ Lal C5 QJ cC cC CZ! Cg S58.
P cg ~
CI 4J 525.
CC 588.
158 288 +58 +88 zcQ 4&A ~
TIME ( SEC )
Shearon Harris Figure 15.2.14-1 Inadvertent Operation of the Emergency Core Cooling System at Power - Nuclear Power, Steam Flow, and Core Tavg versus Time
2688.
2488
'g
~
~~ 5 2288.
cc: 2888
'888.
1688.
1488.
1288.
gg. + 1C88.
888.
688.
7.
6.
4.
2.
8. 58. 189. 158. 288. 258. 388. 358. 488.
TIME ( SEC )
Shearon Harris Figure 15.2.14-2 Inadvertent operation of the Emergency Core Cooling System at Power - Pressurizer Mater, Volume, Pressurizer Pressure, and DNBR versus Time
0

ML18022A714

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