ML20113H622

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Proposed Tech Specs Re Replacement of Std Fuel Assemblies W/Optimized Fuel Assemblies
ML20113H622
Person / Time
Site: McGuire, Mcguire  Duke Energy icon.png
Issue date: 01/11/1985
From:
DUKE POWER CO.
To:
Shared Package
ML20113H604 List:
References
TAC-56746, TAC-56747, NUDOCS 8501250236
Download: ML20113H622 (26)
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Text

{{#Wiki_filter:l 2 TABLE 2.2-1 E3 55 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES C 35 1.. Manual Reactor Trip N.A. N.A. Gl

2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low Setpoint 126% of RATED THERMAL POWER THERMAL POWER S

High Setpoint 5 109% of RATED High Setpoint 1 110% of RATED n3 THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with < 5.5% of RATED THERMAL POWER High Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds

4. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with < 5.5% of RATED THERMAL POWER High Negative Rate-a time constant > 2 seconds with a time constant 1 2 seconds

5. Intermediate Range, Neutron 1 25% of RATED THERMAL POWER 1 30% of RATED THERMAL POWER h) o' Flux 5

5

6. Source Range, Neutron Flux i 10 counts per second 1 1.3 x 10 counts per second
7. Overtemperature AT See Note 1 See Note 3
8. Overpower AT See Note 2 See Note 3
9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig RF gg
10. Pressurizer Pressure--High 1 2385 psig 5 2395 psig gy
11. Pressurizer Water Level--High 5 92% of instrument span 5 93% of instrument span ne gigF
12. Low Reactor Coolant Flow 1 90% of design flow per loop *

> 89% of design flow per loop

  • CC
  • Design flow is 98 400 gp: pcr !00p for Unit 1 :nd 97,220 gpm per loop,fsr 'Jnit 2, j2 '

7 ^ ^) ~ 8501250236 850111 '~ PDR ADOCK 05000369 P PDR L .. l

a TABLE 2.2-1(Continuef E E -REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS W m x ..) FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 2 3 13{ Steam Generator Water > 12% of? span from 0 to 30% of > 11% of span from 0 to 30% of' Level--Lowf ow-RATED THERMAL POWER, increasing RATED THERMAL POWER, increasing l linearly to ?1(4Jnit_1)r g. 40.0% W ;>.i of span at 100% to M '"r.! M 39.0% [Dn 4:21 of span at 100% of RATED THERMAL of RATEDiT,HERMAL POWER. POWER. m s /

14. Undervoltage-Reactor

> 5082.vol's-each bus t Coolant Pumps ~ > 5016 vots-each bus

15. Underfrequency-Reactor

> 56.4 Hz 'each bus > 55.9 Hz each bus Coolant Pumps' I' j

16. Turbine Trip m

a. Low Trip System Pressure > 45 psig > 42 psig b. Turbine Stop Valve Closure > 1% open > 1% cpen

17. Safety Injection Input N.A.

N.A. from ESF 18. Reactor Trip System' Interlocks kk -10 a. Intermediate Range Neutron Flux, P-6, > 1 x 10 -11 ga Enable Block Source Range Reactor Trip amps > 6 x 10 amps =a a kk b. Low Power Reactor Trips Block, P-7 22PP 1) P-10 Input 10% of RATED > 9%, i 11% of RATED X THERMAL POWER THERMAL POWER _gg 2) P-13 Input i 10% RTP Turbine $ 11% RTP Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent my

= TABLE 2.2-1 (Continued) C5 REACTOR TRIP-SYSTEM INSTRUMENTATION TRIP SETPOINTS A NOTATION E .3 NOTE 1: OVERTEMPERATURE AT AT (f I ) (y,1Ts3) 1 AT, {K -K2( )[T(y 3)-T'] + K (P-P') - f (AI)} y 3 y m Where: -AT = Measured AT by RTD Manifold Instrumentation, l 'b y 3 Lead-lag compensator on measured AT, = Time constants utilized in the lead-lag controller for

11. T2

= AT, tg 2 8 sec., T2 1 3 sec., N 1 y, Lag compensator on measured AT, = Time constants utilized in the lag compensator for AT, 131 X l T3 = (IIrri4 6 sec. (noit ftT, AT, Indicsted AT at RATED THERMAL POWER, = K 1 1.200 QIntM),1.40GG-(linit I), y FV K 0.0222 = @d 2 1+TS4 kk 1 + TsS The function generated by the lead-lag controller for T,yg dynamic compensation, = 22PP Time constants utilized in the lead-lag controller for T 14, 13 = y 14 1 28 sec, is 1 4 sec.,

avg, 22 T.

Average temperature, *F, = nY NO 1 y, Tc3 Lag compensator on measured T,yg, =

I TABLE 2.2-1 (Continued) a5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS M 4 NOTATION (Continued) c-5 NOTE 1: (Continued) GI Time constant utilized in the measured T lag compensator, Ts < N ' l Ts = 111n4cl1. 6 sec. (tht4L2_), avg o, i j m T' 588.2*F Reference T,yg at RATED THERMAL POWER, K3 0.001095, = P Pressurizer pressure, psig, = P' 2235 psig (Nominal RCS operating pressure), = m S = Laplace transform operator, sec-1 a and f (al) is a function of the indicated difference between top and bottom detectors y of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for qt ~9h between -29% and +9.0%.(valish--=eh-d -4:qqner=33; f}(61) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt

  • 9 is total THERMAL POWER in percent of RATED b

THERMAL POWER; waN e we aR (ii) for each percent that the magnitude of qt b exceeds -29% imut 4;,Mid~init U, f ~9 the AT Trip Setpoint shall be automatically reduced by 3.151% of its value at RATED THERMAL POWER; and 55 (iii) for each percent that the magnitude of q q exceeds +9.0% N IX Ilfrdt<Q,theATTripSetpointshallbeAutomIticallyreducedby1.50% gg (unn. 27;-1-4W%.411n111) of its value at RATED THERMAL POWER.

3. 3.

A vv I

) TABLE 2.2-1 (Continued) [ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS S, NOTATION (Continued) m a NOTE 2: OVERPOWER AT E ] AT(f ) (y,ITs3) 1 AT {K -K IS 1 1 3 (y, t 3) (y, 7c3) T -K U(1 + r S)- T"] - f f0IN g 4 7 6 2 s m Where: AT As defined in Note 1, = ro 1 As defined in Note 1 = T2. T2 As defined in Note 1 = 1+TS As defined in Note 1, = 3 AT, As defined in Note 1, = g K 5 1.0900 S Q r l. = -(U d t, 4 } K = 0.02/ F for increasing average temperature and 0 for decreasing average 5 temperature, T5 I 7 The function generated by the rate-lag controller for T,yg = 1+1S7 compensation, dynamic pg T7 = Time constant utilized in the rate-lag controller for T,yg, 1 em 7 > 5 sec, b 1+rS As defined in Note 1, = s nn As defined in Note 1, ,22 Ts = .o K = 0.00169/ F for T > T" and K6 = 0 for T 5 T", 6 22 11 ne

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within: the allowed operational space defined by Figure 3.2-1 for RAOC operation, a. or b. within a i M dt:11, 5 CI!n4<27 percent target band about the target l flux difference during base load operation. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION: For RAOC operation with the indicated AFD outside of the Figure 3.2-1 a.

limits, 1.

Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux - High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. ND** b. For Base Load operation above APL with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target flux difference: 1. Either restore the indicated AFD to within the target band limits within 15 minutes, or ND 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 50% of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 limits.

  • See Special Test Exception 3.10.2.

ND

    • APL is the minimum allowable power level for base load operation and will be provided in the Peaking Factor Limit Report per Specification 6.9.1.9.

McGUIRE - UNITS 1 and 2 3/4 2-1 Amendment No. (Unit 1) Amendment No. (Unit 2) /

g POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships: q F (Z) [2.26] [K(Z)] for P > 0.5 W q t=( U ] ',. T : " s (Unit J F (Z) [2 ] [K(Z)] for P < 0.5 lun k:ZT q EE'ZF f^P M, n Where: P _ THERMAL POWER - RATED THERMAL POWER ' and K(Z) is the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: MODE 1. ACTION: With F (Z) exceeding its limit: 9 a. Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit o within 15 minutes and similarly reduce the Power Range Neutron I Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K ) have been reduced at least 1% (in AT span) for each 4 1% F (Z) exceeds the limit; and q b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. q l i McGUIRE - UNITS 1 and 2 3/4 2-6 Amendment ilo. (Unit 1) Amendment No. (Unit 2) t

F SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 For RAOC operation, F (z) shall be evaluated to determine if F (z) is within its limit by: 9 9 Using the movable incore detectors to obtain a power distribution a. map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. b. Increasing the measured F (z) component of the power distribution q map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of specification 3.2.2 are satisfied. Satisfying the following relationship: c. 1) .v. M Fq (z) 2.26 x K(z) for P > 0.5 @t2k2) (z)

  • 1'r
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U N 2 x )frP 5 0.5 DI!!it.JJ Fq (z) )x0 where F"(z) is the measured F (z) increased by the allowances for q manufacturing tolerances and measurement uncertainty, 7.'IS-(tinMcI) .and 2.26 *CI!nM 4t) is the F limit, K(z) is given in Figure 3.2-2, P q i is the relative THERMAL POWER, and W(z) is the cycle dependent function that a:: counts for power distribution transients encountered during normal operation. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9. N d. Measuring Fq (z) according to the following schedule: 1. Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or 9 2. At least once per 31 Effective Full Power Days, whichever occurs first.

  • 0uring 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.

McGUIRE - UNITS 1 and 2 3/4 2-7 AmendmentNo.y(Unit 2) Unit 1) Amendment No.f,(

c SURVEILLANCE REQUIREMENTS (Continued) With measurements indicating e. maximum (FM(z)} k K(z) / over z has increased since the previous determination of F "(z)'either of the following actions shall be taken: q N 1) Fq (z) shall be increased by 2% over that specified in Specifi-cation 4.2.2.2c. or N 2) Fq (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum [FN (z) is not increasing. over z 1 K(z) f. 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 cxpression: q h # f 2.15 for P > 0.5 (Unit 1) P x K(z)- J J (maximum FM(z)xW(z$ p over z xg gp>u 2.26 xK(If - -1 x 130 for P < 0.5 (Unit 1) { 2 Sg xK(z)J ~ .I (maximum Mq(z)xW(z)h-1{x100 (over z for P < 0.5 UIDW4) 2.26 xK(z)b 0.5 J / 1 2) One 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 of 3.2-1 by 1% AFD for each percent F (z) exceeds its limits as deter q mined in Specification 4.2.2.2f.1). Within 8 hours, reset the AFD alarm setpoints to these modified limits, or McGUIRE - UNITS 1 and 2 3/4 2-8 Amendment No. '(Unit 1) Amendment No. (Unit 2)

~ SURVEILLANCE REQUIREMENTS (Continued) b) Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above, or q c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load oper-ation. g. The limits specified in Specifications 4.2.2.2c, 4.2.2.2e., and 4.2.2.2f. above are not applicable in the following core plane regions: 1. Lower core region from 0 to 15%, inclusive. 2. Upper core region from 85 to 100%, inclusive. ND 4.2.2.3 Base Load operation is permitted at powers above APL if the following conditions are satisfied: a. Prior to entering Base Load operation, maintain THERMAL POWER above ND APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours. Maintain Base Load operation surveillance (AFD within F :

^. 1) 4 15% Bra k dQ of target

[ flux difference) during this time period. Base Load operation is ND then permitted providing THERMAL POWER is maintained between APL OL ND and APL or between APL and 100% (whichever is most limiting) and F surveillance is maintained pursuant to Specification 4.2.2.4. q BL APL is defined as: .-1 Q L,mi..i gZ.15xK(Z)]T1'05)U$it1), i over M p (Z) e j APL ' = *i"I*"" [ (2.26 x K(Z) ] x 100% (}Ib4:0 O ver M p (Z) x W(Z)BL q where: F (z) is the measured F (z) increased by the allowances for q manufacturing tolerances and measurement uncertainty. The F limit q is E (m.M 2.26 (Un4::0 K(z) is given in Figure 3.2-2. } W(z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during base load operation. The function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9. b. During Base Load operation, if the THERMAL POWER is decreased below ND APL then the conditions of 4.2.2.3.a shall be satisfied before re-entering Base Load operation. McGUIRE - UNITS 1 and 2 3/4 2-9 Amendment No. '(Unit 1) Amendment No (Unit 2)

SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.4 During Base Load Operation F (Z) shall be evaluated to determine if F (Z) is within its limit by: 9 q Using the movable incore detectors to obtain a power distribution a. map at any THERMAL POWER above APLND b. Increasing the measured F (Z) component of the power distribution q map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of specification 3.2.2 are satisfied. Satisfying the following relationship: c. f nit F (Z) 2.26 Z) for P > APLND yp,ggg) N A where: F (Z) is the measured F (Z). The F limit is.h 4T 9 q W 2.26 D:rpte::ZJ. K(Z) is given in Figure 3.2-2. P. is the relative THERE\\L POWER. W(Z) is the cycle dependent function that accounts for limited power dis-tribution transients encountered during normal operation. This function is given in tae Peaking Factor Limit Report as per Specification 6.9.1.9. d. Measuring F (Z) in conjunction with target flux difference deter-mination according to the following 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 ND maintained above APL for the 24 hours prior to mapping, and 2. At least once per 31 effective full power days. With measurements indicating e. F (Z) maximum [ g 7) ] over Z has increased since the previous determination F (Z) either of the N following actions shall be taken: 1. F (Z) shall be increased by 2 percent over that specified in 4.2.2.4.c, or McGUIRE - UNITS 1 and 2 3/4 2-9a Amendment No. (Unit 1) Amendment No. (Unit 2)

SURVEILLANCE REQUIREMENTS (Continued) M 2. f (Z) shall be measured at least once per 7 EFPD until 2 successive map:; indicate that F (Z) maximum [ ] 1s not increasing. over Z f. With the relationship specified in 4.2.2.4.c above not being satisfied, either of the following actions shall be taken: 1. Place the core in an equilbrium condition where the limit in 4.2.2.2.c is satisfied, and remeasure F"(Z), or 2. Comply with the requirements of Specification 3.2.2 for F (Z) 9 exceeding its limit by the percent calculated with one of the following expressions: I~ = x W(Z)o, \\ N j [(max. ove z g 2~.15 J APL (Unit 12 /y x K(Z) p M F (Z) x W(Z) L ND [(max. over z of [ 2.26 ] ) -1 ] x 100 for P > APL x K(Z) p g. The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plan 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 for reasons other than meeting the requirements q of specification 4.2.2.2 an overall measured F (z) shall be obtained from a power q distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. McGUIRE - UNITS 1 and 2 3/4 2-9b Amendment No. '(Unit 1) Amendment No. (Unit 2)

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i *i-i;i :-L A 0. 0 2 4 6 8 10 1' BOTTOM CORE HEIGHT ( FEET ) TOP FIGURE 3.2-2/ K(Z)- NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT W q McGUIRE - UNITS 1 and 2 3/4 2-12 Amendment No. (Unit 1) Amendment No. (Unit 2)

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0.0 FF / M 'I d / e '2 &i a n a n a a a 9 9 5c, m o 33 4 4 J J B 2 2 -EE CORE HEIGHT (FT) ? C 22 FIGURE 3.2-2b ((- K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT (UNIT 2) q ro N

50 I PENALTIES OF 0.1% FOR UNDETECTED FEED-o a g) + WATER VENTURI FOULING AND MEASUREMENT REGION 2 ~ 4g CERTAINTIES OF 1.7% FOR FLOW AND 4% FOR CORE MEASUREMENT OF F ARE (1.04. .SSI E INCLU IN THIS FIGURE. C 5 4s x / b b N C E Y }A ACCEPTABLE Y o OPERATION d

a REGION a4 42

/ UNACCEPTABLE R o OPERATION a b-REGION = / 4. / f (1.0\\. 39 ACCtFTA,LE OPERA' WN REGIO fFOR 555M.nur I / <SS% RTP (1.0, 38.97 g / 294% RTP H.0, M.57) ,' \\ 33 38 kS2% RT" ~1 I 11.0, 37.79) N 'b (Ts @g / $50% n a r (1.0, 37.39) -l I kk / cp #*- En aa / M

== x ? ?? 0.06 / 0.32 0.S4 0.96 0.98 1.00 1.02 1.04 1. 1.08 { K \\ / R. = F 11.4 5 ~.311.OsP) 4 s UC / q FIGURE S TOTAL OW RATE VERSUS R (UNIT ) (lrl n O

E E E ' PENALTIES OF 0.1% FOR UNDETECTED' FEED- ~h 46 WATER VENTURI FOULING AND MEASUREMENT = UNCERTAINTIES OF 1.7% FOR FLOW AND 4% j d FOR INCORE MEASUREMENT OF F " ARE ~ g l INCLUDED IN THIS FIGURE. I = m c. { M m O b: us l Q ACCEPTABLE e OPERATION S 42 REGION O d a i ta 4 1 UNACCEPTABLE m t-OPERATION <a REGION m l I I ACCEPTABLE OPERATION REGION FOR 598% RTP (1.0, 38.y) <S6% RTP (1.0, 38.499) 38 (1.0,38.11hl 594% RTP- ! R 8. i 3a 592% RTP (1.0, 37.721)

gg 590% RTP (1.0, 37.332')

) <o r+ 11.0, 36.944) ! 72 gg 36 X 0 09 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 ? 22 , oo R=F )M AH 11.4911.0 + 0.311.0 P)] j Figure 3.2-3/ RCS FLOW RATE VERSUS R - FUUR LOOPS IN OPERATION,(llrrRJt)

2 E. TABLE 3.3-2 A REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES EZ v5 FUNCTIONAL UNIT RESPONSE TIME g 1. Manual Reactor Trip N.A. a. ^2 2. Power Range, Neutron Flux < 0.5 second* 3. Power Range, Neutron Flux, High Positive Rate N.A. 4. Power Range, Neutron Flux, High Negative Rate < 0.5 second* 5. Intermediate Range, Neutron Flux N.A. 6. Source Range, Neutron Flux N.A. y 7. Overtemperature AT 1614 8.0 Iti,4tdQ seconds

  • f.

ua 8. Overpower AT < TXTthdt:II, 8.0 (tiniQ) seconds

  • i 9.

Pressurizer Pressure--Low _ 2.0 seconds 10. Pressurizer Pressure--High _ 2.0 seconds 11. Pressurizer Water Level--High N.A. $N dE A yg Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion gg cf the channel shall be measured from detector output or input of first electronic component in channel. T et 22 kA "r

\\ y TABLE 3.3-2 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES e g FUNCTIONAL UNIT RESPONSE TIME U 12. Low Reactor Coolant flow g a. Single Loop (Above P-8) i 1.0 second o. b. Two Loops (Above P-7 and below P-8) i 1.0 second na 13. Steam Generator Water Level--Low-Low 1 Nit-IL 3.5 (Unib2J seconds 14. Undervoltage-Reactor Coolant Pumps < 1.5 seconds 15. Underfrequency-Reactor Coolant Pumps < 0.6 second 16. Turbine Trip a. Low Fluid Oil Pressure N.A. y b. Turbine Stop Valve Closure N.A. 17. Safety Injection Input from ESF N.A. 18. Reactor Trip System Interlocks N.A. 19. Reactor Trip Breakers N.A. 20. Automatic Trip and Interlock Logic N.A. IN

Y ti 8

,n 22 bb a n,

.) pc TABLE 3.3-4 (Continued) c, 55 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E! c3 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES = v! 7. Auxiliary Feedsater g a. Manual Initiation N.A. N.A. a. no b. Automatic Actuation Logic N.A. N.A. and Actuation Relays c. Steam Generator Water Level--Low-Low 1) Start Motor-Driven Pumps > 12% of span from 0 to > 11% of span from 0 to 30% of RATED THERMAL POWER, 30% of RATED THERMAL POWER, increasing linearly to increasing linearly to ca 'E > S4 d M, 40.0% > 53;1iW-(Wit 31, 39.0% Tthete21 of span at 100% of Tunic 21 of span at 100% of ca A3 RATED THERMAL POWER. RATED THERMAL POWER. o> 2) Start Turbine-Driven Pumps > 12% of span from 0 to > 11% of span from 0 to 30% of RATED THERMAL POWER, 30% of RATED THERMAL POWER, increasing linearly to increasing linearly to > Scs&4snTCIL 40.0% > Yts2FfW4CIL 39.0% fikst 23 of span at 100% of (thMc2) of span at 100% of RATED THERMAL POWER. RATED THERMAL POWER. 3m>gp d. Auxiliary Feedwater > 2 psig > 1 psig Suction Pressure - Low E ip' (Suction Supply Automatic pgg Realignment) r+ r, z== e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints ?? Start Motor-Driven Pumps and Allowable Values f. Station Blackout - Start 3464 i 173 volts with a > 3200 volts 5FEi Motor-Driven Pumps and 8.5 1 0.5 second time 77 Turbine-Driven Pump delay v- ~~~' g. Trip of Main Feedwater Pumps - N.A. N.A. Start Motor-Driven Pumps

e l 1 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop-erties 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 are 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 as the maximum local 0 heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing toler-ances on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper q bound envelope of 2.26 (Untr ;). 2.1 3 - il times the normalized axial l peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full-length 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. The-value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value 'by the appropriate fractional THERMAL POWER level. The periodic updating of .the target flux difference value is necessary to reflect core burnup considerations. McGUIRE - UNITS 1 and'2 8 3/4 2-1 Amendment No. (Unit 1) Amendment No. (Unit 2)

m POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) At power levels below APLND 3.2-1, i.e. that defined by the RAOC operating procedure and limits., the limits were calculated in a manner such that expected operational transients, These

e. g'. load follow operations, would not result in the AFD deviating outside of

~ those limits. of time allowed outside of the limits at reduced power levels j in significant xenon redistribution such that the envelope of peaking factors i would change sufficiently to prevent operation in the vicinity of the APL D + power level. At power levels greater than APLND

1) RAOC, the AFD limit of which are defined by Figure 3.2-1, and 2) Ba i

25% (Dame <W band abouc a carget value.. The RAOC o ND APL is the same as that defined for operation below APL i ND However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with F (z) less than its limiting.value. To allow q I operation at the maximum permissible value, the Base Load operating procedure restricts the indicated AFD to relatively small target band and power swings S (AFD target band of 53Fr (men 11,.15% {LWt $ APL 1 power 1 APL ' or 100% l ND O Rated Thermal Power, whichever is lower). For Base Load operation, it is expected that the plant will operate within the target band. Operation out - ' side of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would 4 change sufficiently to prohibit continued operation in the power region defined above. To assure there is no residual xenon. redistribution impact from past-operation on the Base Load operation, a 24 hour waiting period at a power level i 0 'above APL and allowed by RAOC is necessary. During this time period load-changes and rod motion are restricted to that allowed by the Base Load procedu ] After.the waiting period extended Base ' Load operation is permissible. The computer determines the one minute average of each of the OPERA 8LE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are:

1) outside the allowed AI power operating space (for RAOC operation), or 2) outside the

' allowed AI target band (for Base Load operation). These alarms are active when power is greater than: ND (for Base Load operation).1) 50% of RATED THERMA or 2) APL Penalty deviation minutes for Base-which operation outside of the target band is allowed. Load op McGUIRE - UNITS 1 and 2 B 3/4 2-2 Amendment No.\\/(Unit 1) AmendmentNo.A(Unit 2)

~ ATTACHMENT 2 JUSTIFICATION AND SAFETY ANALYSIS Mr. H. B. Tucker's (DPC) November 14, 1983 letter to Mr. H. R. Denton (NRC/0NRR) described planned changes in the fuel design for McGuire Nuclear Station, Units 1 and 2. McGuire Unit 1 had been operating with a Westinghouse 17x17 low-parasitic (SID) fueled core. It was planned to refuel Unit 1 with Westinghouse 17x17 Reconstitutable Optimized Fuel Assembly (OFA) regions. As a result, future core loadings would range from an approximately 1/3 0FA - 2/3 STD transition core to eventually an all 0FA fueled core. McGuire Unit 1 is currently operating with the first such 0FA reload region (Cycle 2), with the second 0FA region scheduled for the upcoming Cycle 3 refueling. (McGuire Unit 2's first OFA reload region is scheduled for the upcoming cycle 2 refueling - Ref. Mr. H. B. Tucker's November 16, 1984 McGuire 2/ Cycle 2 0FA reload submittal). The OFA fuel has similar design features compared to the STD fuel which has had substantial operating experience in a number of nuclear plants. The major differences are the use of six intermediate (mixing vane) Zircaloy grids for the OFA fuel versus six inter-mediate (mixing vane) Inconel grids for STD fuel and a reduction in fuel rod diameter. Major advantages for utilizing the OFA are: (1) increased efficiency of the core by reducing the amount of parasitic material and (2) reduced fuel cycle costs due to an optimization of the water to uranium ratio. The above letter provided a Reference Safety Evaluation Report summarizing the evaluation / analysis performed on the region-by-region reload transition from the McGuire Units 1 and 2 STD fueled cores to cores with all optimized fuel. The report examined the differences between the Westinghouse OFA and STD designs and evaluated the effects of these differences for the transition to an all 0FA core. The evaluation considered the standard reload design methods described in WCAP-9272 and 9273, " Westinghouse Reload Safety Evaluation Methodology," and the transition effects described for mixed cores in Chapter 18 of WCAP-9500-A, " Reference Core Report - 17x17 Optimized Fuel Assembly." Consistent with the Westinghouse STD reload methodology for analyzing cycle specific reloads, parameters were chosen to maximize the applicability of the transition evaluations for each reload cycle and to facilitate subsequent determination of the applicability of 10 CFR 50.59. Subsequent cycle specific reload safety evaluations will verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in the reference report. A summary of the mechanical, nucicar, thermal and hydraulic, and accident evaluations for the McGuire Units 1 and 2 transitions to an all 0FA core are given in the reference report. WCAP-8183, " Operational Experience with Westinghouse Cores," presents the operating experience through December 31, 1983 of six 17x17 0FA demonstration assemblies (two in each of three reactors) which have the McGuire 1 and 2 design features. During 1983 four assemblies operated in their fourth cycle and were expected to achieve burnups of 39,000 and 35,000 MWD /MTU respectively during the first quarter of 1984, and two others completed their second cycle of irradiation with a burnup of 22,000 MWD /MTU and were operating in their third cycle. All demonstration 17x17 0FAs examined were in good or excellent condition. This provides evidence of favorable operation of Zircaloy grids and reduced fuel rod diameters which are the major new design features of the 17x17 0FA. In addition, Mannshan Unit 1 was scheduled to begin irradiating a full core of 17x17 0FAs during the first half of 1984, and McGuire Unit I has operated nearly a full cycle with an OFA reload region (60 17x17 0FA assemblies).

r Page 2 The results of evaluation / analysis and tests described in the Reference Safety Evaluation Report led to the following conclusions: a. The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mechan-ically compatible with the STD design, control rods, and reactor internals interfaces. Both fuel assemblies satisfy the design bases for the McGuire units. b. Changes in the nuclear characteristics due to the transition from STD to 0FA fuel will be within the range normally seen from cycle to cycle due to fuel management effects. c. The reload 0FAs are hydraulically compatible with the STD design. d. The accident analyses for the OFA transition core were shown to provide acceptable results by meeting the applicabic criteria, such as, minimum ,DNBR, peak pressure, and peak clad temperature, as required. The previously reviewed and licensed safety limits were met. Analyses in support of this safety evaluation establish a reference design on which subsequent reload safety evaluations involving 0FA reloads can be based. (Attachment 2A of H. B. Tucker's December 12, 1983 Unit 1/ Cycle 2 0FA reload submittal presents those detailed non-LOCA and LOCA accident analyses of the McGuire Units 1 and 2 FSAR impacted by the changes as determined in S(ction 6.0 of the Reference Safety Evaluation Report. The information contained within was prepared using the NRC Standard Format and Content Guide, Regulatory Guide 1.70, Revision 3 as it applies to McGuire Nuclear Station Units 1 and 2). e. Plant operating limitations giver in the Technical Specifications affected by use of the OFA design and positive MTC would be satisfied with the changes noted in Section 7.0 of the report. A is the cycle-specific Reload Safety Evaluation (RSE) for McGuire Unit 1/ Cycle 3. The RSE presents an evaluation for McGuire Unit 1, Cycle 3, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was performed utilizing the methodology, described in WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology". In addition, the NRC has previously approved a similar OFA reload for McGuire Unit i via Ms. E. G. Adensam's (NRC/0NRR) April 20, 1984 letter to H. B. Tucker, and a planned 0FA reload for McGuire Unit 2 is currently under review (Ref. Mr. H. B. Tucker's November 16, 1984 submittal). McGuire Unit 1 is operating in Cycle 2 with Westinghouse 17x17 low parasitic (STD) fuel assemblies and optimized fuel assemblies (OFA). For Cycle 3 and subsequent cycles, it is planned to refuel the McGuire Unit I core with Westinghouse 17x17 optimized fuel assembly (OFA) regions. In the OFA transition licensing submittal to the NRC (Reference Safety Evaluation, November 14, 1983 letter) an analyses of the safety aspects of the transition from STD fuel design to 0FA design was provided. This licensing submittal (which has been approved by the NRC) justifies the compatibility of the OFA design with the STD design in a transition core as well as a full 0FA core. The OFA transition licensing submittal contained mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicabic to the Cycle 3 safety evaluation.

r* Page 3 All of the accidents comprising the licensing basec which could potentially be affected by the fuel reload have been reviewed for the Cycle 3 design described herein. The results of new analyses and the Justification for the applicability of previous results for the temaining analyses is presented in the cycle specific reload safety evaluation. The McGuire Unit 1. Cycle 3 reactor core will be comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1 of the Cycle 3 Reload Safety Evaluation. During the Cycle 2/3 refueling, 60 STD fuel assemblies will be replaced with 60 Region 5 optimized fuel assemblies. A summary of the Cycle 3 fuel inventory is given in Table 1 of the Cycle 3 Reload Safety Evaluation. As in Cycle 2 this cycle will contain one region 4 demonstration assembly of an intermediate flow mixer grid fuel assembly design. This demonstration assembly has been previously discussed in Mr. H. B. Tucker's February 20, 1984 letter to Mr. H. R. Denton. This assembly will be loaded into the core in a manner which satisfies the requirements of the " Safety Evaluation for the intermediate flow mixer grid (IFM) demonstration fuel assembly in McGuire Unit 1" (Davidson, S. L. (Ed.), February 1984). From the evaluation presented in the Cycle 3 Reload Safety Evaluation, it is concluded that the Cycle 3 design does not cause the previously acceptable safety limits to be exceeded. This conclusion is based on the following: 1. Cycle 2 burnup is between 10200 and 10577 FS'D/MTU. 2. Cycle 3 burnup is limited to 11700 FN'D/MTU including a coastdown. 3. There is adherence to all plant operating limitations given in the Technical Specifications as revised by the proposed changes given in Appendix A of the Cycle 3 RSE. To ensure plant operation consistent with the design and safety evaluation conclusion statements made in the Cycic 3 RSE and to ensure that these conclusions ~ remain valid, several Technical Specifications changes will be needed for Cycle 3. These changes are discussed in Section 4.0 and given in Appendix A of the cycle-specific RSE. Differences between the cycle-specific RSE Technical Specification changes to those given in the OFA transition licensing submittal are discussed in the cycle-specific RSE, along with any necessary justifications. Attachment 1 provides copf.es of these specificatio m as they should appear in the McGuire Units 1 and 2 Technical Specifications following approval of the proposed McGuire Unit 2/ Cycle 2 0FA reload license amendments (Ref. Mr. H. B. Tucker's November 16, 1984 submittal) with the appropriate Unit 1/ Cycle 3 changes indicated. It should be noted that certain changes given in Appendix A of the Unit 1/ Cycle 3 RSE are not indicated as needing to be made in Attachment 1 due to their having previously been submitted (for both Units) in the Unit 2/ Cycle 2 OTA reload submittal (these constituted only administrative-type changes (corrections of minor errors / typos, clarifications, etc.) or were improvements incorporated for the Unit 2 specifications which were more conservative than existing Unit i specifications, or are changes to specifications which have no meaning with respect to Unit 1 - these are already reficcted in the typed pages of Attachment 1]. Also, some McGuire Unit 2 specifications are adminis-tratively af fseted in that they are combined into one specification applying to both McGuire Units 1 and 2, but there are no changes to the content of Unit 2 specifications.

P Page 4 Bis the Peaking Factor Limit Report for McGuire Unit 1/ Cycle 3 { whic'1 is provided in accordance with Technical Specification 6.9.1.9. This report provides the k(z) functions that are to be used for RAOC and Base Load l Operation during Cycle 3, and the value for APLND. For both RAOC and Base Load operation, a set of data covering three specific burnup steps is provided which permits the determination of W(Z) at any cycle burnup through the use of three point interpolation. The informatien for Base Load operation has been obtained using a 15 percent AFD about a measured target in the power interval between 80% and 100% of rated thermal power. Figures 1-3 are.the W(Z) functions appropriate for RAOC operation and Figures 4-6 are the W(Z) functions appropriate for Base Load operation. The appropriate W(z) function is used to confirm that the Heat Flux Hot Channel Factor, F (z), will be Q limited to the values specified in the Technical Specifications. l I l i t -}}

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