Text

Application for Amends to Licenses NPF-9 & NPF-17,revising Tech Specs Re Plant Operating Limitations Affected by Use of Optimized Fuel Assembly Designs.Fee paid.Marked-up Tech Specs Encl
	ML20099G748
Person / Time
Site:	McGuire, Mcguire
Issue date:	11/16/1984
From:	Tucker H; DUKE POWER CO.
To:	Adensam E, Harold Denton; Office of Nuclear Reactor Regulation
References
	TAC-53319, TAC-53320, TAC-56336, TAC-56337, TAC-56476, TAC-56477, TAC-56873, TAC-56874, NUDOCS 8411270423
	Download: ML20099G748 (139)
	v • d • e

{{#Wiki_filter:,D Dume POWER GOMPANY P.O. IBOX Wit *49 MN. N.c. 28g4;! HAL 3. TtJCMEN TER EP980ME mmaan November 16, 1984

====== - Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4

Subject:

McGhire Nuclear Station Docket Nos. 50-369 and 50-370 McGuire 2/ Cycle 2 0FA Reload

Dear Mr. Denton:

Mr. it. 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. Cormencing with the first refueling of each of the units, the standard fuel assemblies in use are to be replaced over the next four refuelings with optimized fuel assemblies (OFA). The letter transmitted a reference safety evaluation describing the safety impact of operation with a transition core and an all 0FA core, and indicated that since the transition to 0FA involves changes in operating limits, license amendmente will be required for operation of both Units 1 and 2 beyond the first cycle. McGuire Unit I has already begun this process with the NRC having approved the necessary license amendments via Ms. E. C. Adensam's April 20, 1984 letter to Mr. H. B. Tucker, and Unit 1/ Cycle 2 is currently operating with an OFA reload region. Attached are proposed license amendments to facility operating licenses NPF-9 and NPF-17 for McGuire Nuclear Station Units 1 and 2, respectively. The proposed amendments change plant operating limitations given in the Technical Specifica-tions affected by use of the OFA design for McGuire Unit 2/ Cycle 2 to ensure plant operation consistent with the design and safety evaluations. It should be noted that certain Unit 2 reload changes are applied to Unit I also (as opposed to affecting only Unit 2), but these involve only administrative type changes (corrections of minor otrors/ typos, clarifications, etc.) or are improvements incorporated for the Unit 2 specifications which are more conservative than the existing Unit I specifications. In addition. Technical Specification 3.5.1.2 is revised to delete the inadvertent application to Unit 1 of provisions which do not apply to the current core design. contains the proposed technical specification changes, and Attachment 2 discusses the Justification and Safety Analysis to support the proposed changes. Included in Attachment 2 ist A) the cycle-specific reload safety evaluation for McCuire Unit 1/ Cycle 2 including Fq surveillance and kAOC/ Base Load Technical Specifications. ThepeakingfactorlimitreportforMcGuireUnig2/ Cycle 2 D 0 v - - - a

4 _ Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation November 16, 1984 Page Two which is required in accordance with the proposed McGuire Unit 2 Technical Specification paragraph 6.9.1.9 (as given in Attachment 1) will be submitted by Decencer 14, 1984. Pursuant to 10 CFR 50.91, Attachment 3 provides an analysis performed in accordance with the standards contained in 10 CFR 50.92 which concludes that the proposed amendments do not involve a significant hazards consideration. The proposed amendments have been reviewed and determined to have no adverse safety or environmental impact. For Unit 2/ Cycle 2, the large break LOCA analysis applicable for transition and full 0FA core cycles of McGuire 1 and 2 was performed utilizing the OFA design consistent with the methodology given in the above-mentioned reference safety evaluation for the OFA transition. This analysis utilized the currently approved UHI large break ECCS evaluation model modified to incorporate BART core reflood heat transfer models. BART has been approved for use on non-UHI plants (WCAP-9561-P-A, "BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," Young, M. Y., et. al., March 1984). However, McGuire 1 and 2 are UHI plants. Addendum 1 to WCAP-9561 requesting approval for use of BART technology on UHI plants will be submitted by Westinghouse before the end of November 1984. Also, certain Technical Specification changes such as those involving limiting safety system settings changes (e.g. steam generator low-low level setpoint changes and updating of the lag time constants in the Delta-T and T channels) require plant modifications which are scheduled to be performed avg during the refueling outage. Since these changes are contingent upon NRC approval, any concerns with these should be resolved as expeditiously as possible so as not to impact the modification work. It is requested that the proposed amendments receive timely review and approval in view of the current McGuire Unit 2/ Cycle 2 startup schedule. Unit 2 first refueling chutdown is currently planned for late January with return to service planned for late March 1985. Any changes to this schedule will be provided to the NRC staff. Pursuant to 10 CFR 170.3(y),170.12(c), and 170.21, Duke Power proposes t. hat this application contains license amendments for McGuire Units 1 and 2 subject to fees based on the full cost of the review (to be calculated using the applicable professional staff rates shown in 10 CFR 170.20) and must be accompanied by an application fee of $150, with the NRC to bill Duke Power at six-month intervals for all accumulated costs for the application or when review Ia completed, which-ever is earlier. Accordingly, please find enclosed a check in the amount of $150. We will be pleased to ment with the NRC staff to discuss this matter at the staff's convenience. Very truly yours, & b.i e C tal B. Tucker PBNases Attachments

+ 4 .') r Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation November 16, l'384 'Page Three. cc: (w/all'attacheents) Mr. J. P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. Dayne Brown, Chief Radiation Protection Branch Division,of Facility Services Department of Human Resources P. O. Box 12200 Raleigh, North Carolina 27605 Mr. W. T. Orders Senior Resident Inspector McGuire Nuclear Station I 6 r h I f i i i l i i i

I' Mr. Harold R. Denton Office of Nuclear Reactor Regulation November 16, 1984 Page Four HAL B. TUCRER, being duly sworn, states that he is Vice President of Duke Power-Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this revision to the McGuire Nuclear Station-License Nos. NPF-9 and NPF-17 and that all statements and matters set forth there-in are true and correct to the best of his knowledge.

C Hal B. Tucker,' Vice President l

Subscribed and sworn to before me this 16th day of November, 1984. kb (._ Notary Pub *Jc" My Commission Expires: Septemb_5r 20, 1989 l { [ l

] r i_. o A n AC M I Proposed McGuire 11 nit 1 and 2 Technical S ceification Changes P

~ l 4 ,ey Yj pgfun&C W'WITH Tbtl F**t-Lowiuc FlkM h Y E N. s 665

g{

Flow Per Loop = 95,500 g;:m

h I

[ 65. Eggo * '* f 4 653i U"***** 645 2250o g, Wa on s ( 648-f 655< 653 20 c

/a c

igan ( O 623 f fa (, [ 615 - 612-b 635- \\ 623 .\\ 5*I" Acceptacle Oceration j

598,

[ 555" 553< c 575' (, 575 2 .1 .2 .5 4 .5 .5 .7 .3 9 1. 1. 1.2 ( ) PC1dCR t fr net. ion of nominea I l / \\ l e t ' i ? ? FIGURE 2.1-1b t( UNIT 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION ) ( -i McGUIRE - UNITS 1 and 2 2-2a Amendment rio. (Unit 1) Amendment No. (Unit 2) 1

O F /6 nA'E X 668 Flow Per Loop

97220 3pm 658 2

645 ' a ' Va* *** P ^ bl

t 648 E2Q 655-658-

$625 2% o 5 g ,'.s j 618 - 685 Accephble Operaboa j 680 l 595 - 598 585 8. .I .2 .5 4 .5 .6 .7 .8 .9 1. 1.1 1.3 POWER freaction or nominell ,l i 4 gsa-m2- ..- - - _m -,.y, I i M 1 _ _ _ _ _ _ _ _ _. _ _ _ _. _ _,, _. -. - l

,4 O O ( ) c LJ TABLE 2.2-1 4 M E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E

1. Manual Reactor Trip N.A.

N.A.

2. Power Range, Neutron Flux Low Setpoint - < 25% of RATED Low Sttpoint 1 26% of RATED

{ THERMAL POWER - THERMAL POWER o" High Setpoint - 1 109% of RATED High Setpoint - 1 110% of RATED THERMAL POWER THERMAL POWER 1

3. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with < 'i.5% of RATED THERMAL POWER High Positive Rate a time constant > 2 seconds Gith a time constant > 2 seconds

4. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER j High Negative Rate a time constant > 2 seconds Ulth a time constant > 2 seconds m un j

5. Intermediate Range, Neutron 1 25% of RATED THERMAL POWER 1 30% of RATED THERMAL POWER Flux 5

5

6. Source Range, Neutron Flux 1 10 counts per second i 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

3. Overpower AT See Note 2 See Note 3 g>

l gl

9. Pressurizer Pressure--Low 1 1945 psig

> 1935 psig 4 aa

10. Pressurizer Pressure--High 5 2385 psig i 2395 psig

'll. Pressurizer Water Level--High 5 92% of instrument span f 93% of instrument span g

12. Low Reactor Coolant Flow

> 90% of design flow per loop * > 89% of de ' _. loop

EE

) $$ i

Design flow is 98,400 gpm per loop for Unit I and M gpa per loop for Unit 2.

I j 9 7,11o 1 r \\

TABLE 2.2-1 (Continued) O REACTOR TRIP SYSTEM INSTRUMENTATION IRIP SETPOINTS E, [ FUNCTIONAL UNIT 1 RIP SETPOINT ALLOWABLE VALUES z 3

13. Steam Generator Water 1 12% of span from 0 to 30% of 1 11% of span from 0 to 30% of Level--Low-low RATED TilERMAL POWER, increasing RATED TilERMAL POWER, increasing linearly to 3 54.9% of span at to 53.9% of span at 100% of RA1ED TilERMAL PM(w~,r o,39.o <,(%,r t)

( 100% of RATED T L POWER. ts.wr n,80.0.t a,r 1)

14. Ilndervoltage-Reactor 1 5082 volts-each bus 1 5016 volts-each bus Coolant Pumps

15. Underfrequency-Reactor 1 56.4 ilz - each bus 1 55.9 Ilz - each bus Coolant Pumps

16. Turbine Trip

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

17. Safety Injection Input N.A.

N.A. from ESF 18. Reactor Trip System Interlocks -10 1 6 x 10'II j[ a. Intermediate Range Neutron Flux, P-6, 1 1 x 10 amps amps

4 Enable Block Source Range Reactor Trip is

rj$

b. Low Power Reactor Trips Block, P-7 h,b 1) P-10 Input 10% of RATED 1 9%, 5 11% of RATED TilERMAL POWER TilERMAL POWER f[ 2) P-13 Input 5 10% RIP Turbine . $ 11% RTP Turbine Impulse Pressure impulse Pressure C;c Equivalent Equivalent 1 O O O

f TABLE 2.2-1 (Continued) 2 h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 9 NOTATION ) k c* y NOTE 1: OVERTEMPERATURE AT vi 1+t S 1 8 { AT(y

3) (1 + TaS) I I

S)UIl + 1s5)- 3* 3(P-P') - f (AI)} T o 1 2 l+t y m Where: AT = Measured AT by RTD Manifold Instrumentation, 1 Lead-lag compensator on measured AT, = e It. I2 = Time constants utilized in the lead-lag controller for h m AT, 1 X 8 sec., T2 K 3 sec., 3 d 2 5 f 1 Lag compensator on measured AT, = y, 5 Ta Time constants utilized in the lag compensator for AT, 13 x 2 sec. 6<y.r s),(,su,c,<,) {[. = Indicated AT at RATED THERMAL POWER, U AT = o ( 4.100 K 5 W (Unit 2), 1.4060 (Unit 1), / y r I< yy K = CLZ^ 'Un. Q 0.0222 M jc 2 ,, n 3 3 1+T S ? 4 Nh The function generated by the lead-lag controller for T,yg dynamic compensation, = I+T 5 3 [ zz ( PP T 'gs Time constants utilized in the lead-lag controller for T g' = 4 ("a>> m, Is )l4 sec. avg, )< K

t. )( 28 sec,(uni t t

nngE T Average temperature, F, / = AA ( NH 1 l ag compensator on measured T,yg, = ) 1+1 5 8

O o/ o'v' v TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 7 NOTATION (Continued) d E NOTE 1: (Continued) ( --a 5 Time constant utilized in the measured T lag compensator, to X 2 sec(w.me), 1c = [ w 1 4 6$rcben.orn '9 i, a .\\ 1 T' < 588./*F Reference T, at RATED THERMAL POWER, / K = B _= L " d O.001095' W, l 3 Pressurizer pressure, psig, [ I P = P' 2235 psig (Nominal RCS operating pressure), = n' ( Laplace transform operator, sec"1 E S = 4 I and f (AI) 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 f instrument response during plant startup tests such that: -29 % + 5.o *4 (1) for q q between *36K and 18rW (Unit 2), - 41% and -4.0% (Unit 1); f (AI) = 0, l 3[ b y where q and q are percent RATED THERMAL POWER in the top and bottom t b Mlves of ~ the core respectively, and q d 9 is total THERMAL POWER in percent of RATED t b 38 THERMAL POWER; l 5E -as 4 .} j (ii) for each percent that the magnitude of q ~9 exceeds - M (Unit 2), -41% (Unit 1), h t b the AT Trip Setpoint shall be automatically reduced by ~w ^' 'LW 3.151% II2nHe:131 k - of its value at RATED THERMAL POWER; and - ( ) + 9.o '4 (iii) for each percent that the magnitude of q q exceeds f3rd (Unit 2), -4.0% f' 22 (Unit 1), the AT Trip Setpoint shall be kutomNtically reduced by BMHM% .507. 1 c " 3. (Unit 2), 1.447% (Unit 1) of its value at RATED THERMAL POWER. en UD

TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS o5 NOTATION (Continued)

o m

NOTE 2: DVERPOWER AT 2 E rS 1 1 I A T (1 * ' S1 + r,5) ()< T [K -K ( )( ) T -K [T( )- T"] - f (AI)} d 1-1 Q 2 1 + TsS o 4 5 1+1 S 1+1 S 6 1+ S D 7 8 Y ~ 6 Where: AT = As defined in Note 1, f = As defined in Note 1 [ f ri, T2 = As defined in Note 1 j[- 1 As defined in Note 1, = 1+T 5 3 j c ~ AT = As defined in Note 1, ~ o O K 5 1.090)( (Unit 2),1.0708 (Unit 1), )f 4 KS 0.02/ F for increasing average temperature and 0 for decreasing average = temperature, d c, r,S The function generated by the rate-lag controller for T,yg dynamic = y, 3 compensation, 1 3" RR ry = Time constant utilized in the rate-lag controller for Tavg, 17 5 5 sec h. 1 W, aa "5 "5 1 y, 3 As defined in Note 1, = 55 l rs As defirsed in Note 1, = Y c f 22 K ' U F (Lafi 0.00'169/ F Uh 4<:d for T > T" and l d 6

3. 3.

K = 0 for T 5 T", eo 6 y e G 9

~. \\. .v i TABLE 2.2-1 (Continued) x n REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

o
m NOTATION (Continued)

C .d 2 As defined in Note 1, 5 ' ' - l 4 T = j 1 J -<588./FReferenceT at RATED THERMAL POWER, l'. T" = avg i o, 3 a S As defined in Note 1, and = >~ N f (AI) 0 for all AI. = 2 Note 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2%. 'I 7 i 0 .i t l' I i e 4 i B3 l 3 aa te =a J l ZZ oO 'l CC -. 3 3 OC i ~

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have gea 3related to DNB tt.. a p, th: """-1 :: ve!sti:r. The t'""-1 DN" correlati'88das been ] developed to predict the DNB flux and the location of DNB for axially uniform gand nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR), f defined as the ratio of the heat flux that would cause ONB at a particular 4 kcore location to the local heat flux, is indicative of the margin to DNB. P -- O j The minimum value of the DNBR during steady-state operation,. normal m V opera nal transients, and anticipated transients is limited to 1.30 (base 7

upon W-3 rrelation).

This value corresponds to a 95% probability at a f, ' confidence vel that DNB will not occur and is chosen as an appropria ( piargin to DNB r all operating conditions. ?p The curves of ures 2.1-1 and 2.1-2 show the loci of po s of THERMAL j POWER, Reactor Coolant stem pressure and average temperat e for which the <minimum DNBR is no less n 1.30, or the average enthal at the vessel exit [is equal to the enthalpy o aturated liquid. I t l N These curves are based on an nthalpy hot nnel factor, F f 1.55 iand a reference cosine with a peak o 1.55 f axial power shape.H,An a N hallowance is included for an increase 1 F " at reduced power based on the 3 (gxpression: (

d F

= 1.55 [1+ 0.2 (1-P g k Where P is the fr ion of RATED THERMAL P0 ?/ These limiting h flux conditions are higher than tho calculated for Lthe range of all rol rods fully withdrawn to the maximum a wable control @odinsertion umir.g the axial power imbalance is within the li 'ts of the j f7 (AI) fu lon of the Overtemperature trip. When the axial power alance ' i5 not hin the tolerance, the axial power imbalance effect on the Ov pf temp ature AT trips will reduce the Setpoints to provide protection cons 1 ent c%i core Safety Limits. N i i McGUIRE - UNITS 1 and 2 B 2-1 Amendment No. (Unit 1) Amendment No. (Unit 2)

SAFETY LIMITS BASES 9 m i u The DNB design basis is as follows: there must be at least a 1,95% probability that the minimum DNBR of the limiting rod during Condition I and f JII events is greater than or equal to the DNBR limit of the ONB correlation being ,used (the WRB-1 correlation in this application). The correlation DNBR set such i that there is a 95% probability with 95% confidence that DNB will not occur when ? ?theminimumDNBRisattheDNBRlimit. f n In meeting this design basis, uncertainties in plant operating parameters, h J 3, nuclear and thermal parameters, and fuel fabrication parameters are considered y y '.atistically such that there is at least a 95% confidence that the minimum DNBR P (for the limiting rod is greater than or equal to the DNBR limit. The uncertainties e 1,in the above plant parameters are used to determine the plant DNBR uncertainty. ?This DNBR uncertainty, combined with the correlation DNBR limit, establ.ishes a j ' parameters without uncertainties. design DNBR value whicn must be met in pla S (l ( P The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor ( Coolant System pressure, and average temperature below which t5e calculated DNBR ) (lis no less than the design DNBR value or the average enthalpy at the vessel exit < is less than the enthalpy of saturated liquid. } c f The curves are based on a nuclear enthalpy rise hot channel factor, Ph, of l } 0 1.49 and a reference cosine with a peak of 1.55 for axial power shape. An allow-1, anceisincludedforanincreaseinFhatreducedpowerbasedontheexpression: Fh = 1.49 [1 + 0.3 (1-P)] h { Where P is the fraction of RATED THERMAL POWER. P ( These limiting heat flux conditions are higher than those calculated for the . range of all control rods fully withdrawn to the maximum allowable control rod 6 insertion assuming the axia1 power imbalance is within the limits of the ft (AI) ? function of the Overtemperature trip..When the axial power imbalance is not within 4 [(the tolerance, the axial power imbalance effect on the Overtemperature AT trips 4 . will reduce the setpoints to provide protection consistent with core safety limits. ? j<, 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radio-nuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3107 psig,.125% of design pressure, to demonstrate integrity prior to initial operation. McGUIRE - UNITS 1 and 2 8 2-2 Amendment No. (Unit 1) Amendment No. (Unit 2)

LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued) The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Range, Neutron Flux. High Rates The Poweb Range Positive Rate trip provides protection against rapid flux C increases which are characteristic of rod ejection events from any power level. 2 Specifically, this trip complements the Power Range Neutron Flux High and Low P trips to ensure that the criteria are met for rod ejection from partial power. The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power, a rod drop accident of a single or multiple rods could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from 1 occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those ccntrol rod drop accidents for which DNBR's wi',1 be greater than W mc usisu see.r ovag ums. l Intermediate and Source Range, Neutron Flux The Intermediate arid Source Range, Neutron Flux trips provide core protection dering reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from hubcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux c'ngnnels. The Source Range channels will initiate a Reactor trip at about 10 s counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. 1 l i O 1 McGUIRE - UNITS 1 and 2 8 2-4 ' Amendment No. (Unit 1) Amendment No (Unit 2)

.N b 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUT 00WN MARGIN - T >200 F LIMITING CONDITION FOR OPERATION f 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to mi ca--*4 ? .[IDndr-2C 1.3% delta k/k G1me<t] for four loop operation. ? APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than i n "::.. ~ m y " 9L 1.3% delta k/k j 'CIDz:4=tt immediately initiate and continue boration at greater than or equal a to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or 1 5 equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to ~~i C 2": i"n'r 7i-1.3% delta k/k T1frsk:C: a. Within 1 hour after detection of an inoperable control rod (s) and at least once per 12 hours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1.0 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; When in MODE 2 with K,ff less than 1.0, within 4 hours prior to c. achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;. d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.le., below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

See Special Test Exception 3.10.1.

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

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION P 3.1.1.3 The moderator temperature coefficient (MTC) shall be: a. W - % less positive than the limits shown in Figure 3.1-0,ANo L =b. I r ' nit 2, !ere petitife th:r 0 delt k/k/ I fe, t:c !' r:d: i +2d r:m, J -beginn<y ef-7. lifo rnni), het zer rugouqt ng;;3 ;;;3;;;;n, ;,yg 1

6. /.

F ""' 1 " L Less negative than -4.1 x 10 4 delta k/k/ F for.the j all rods with:frawn, end of cycle life (EOL), P.ATED THERMAL POWER condition 1 APPLICABILITY: Specifications 3.1.1.3a. N - MODES 1 and 2* only.# 1 Specification 3.1.1.3g. - MODES 1, 2, and 3 only.# 6 ACTION: With the MTC more positive than the limit of Specification / 3.1.1.3a. I a. W _.1 % above, operation in MODES 1 and 2 may proceed provided: j 1. Em wrk:1, Control rod withdrawal limits are established and f j maintained sufficient to restore the MTC to less positive than the limits shown in Figure 3.1-0 within 24 hours or be in HOT { STANDBY within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; /2 o Un't , c ntrol ro w' hd awal imi s are sta is e n in i d ffi en t or he to ess o ~i' th 1 /F it our r e H S D hi he I pxt ur. Te w dra im' s al be addit on to tha i, ett on mi c f <nor t f cation .l..6: m i i ?.. f. The control rods are maintained within the withdrawal limits } established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods q withdrawn condition; and 5 c

3. g.

A Special Report is prepared and submitted to the Commission l pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal I limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods i withdrawn condition. i

b i

b. With the MTC more negative than the limit of Specification 3.1.1.3g. l above, be in HOT SHUTDOWN within~12 hours.

With K,ff greater than or equal to 1.0.

See Special Test Exception 3.10.3.

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

1 (T REACTIVITY CONTROL SYSTEMS O SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: a. The MTC shall be measured and compared to the BOL limit of f Specification / 3.1.1.3a.W.1. M above, prior to initial I: operation above 5% of RATED THERMAL POWER, after each fuel loading; and b. The MTC shall be measured at any THERMAL POWER and compared to ~4 -3.2 x 10 delta k/k/*F (all r,ods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppe. In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 delta k/k/*F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specifica-tion 3.1.1.3g., at least once per 14 EFPD during the remainder of the l fuel cycle. 6 O 1 i i i i i / J i t McGUIRE - UNITS 1 and 2 3/4 1-5 Amendment No. (Unit 1) Amendment No. (Unit 2)

L M l {,, P k 0.5 2 3 k 4 e 0.4 5 E Acceptable Unaccectable ~ Operation Ooeration 0.3 ( 3 ( b s. g 0.2 5-5 ( ( e C{ 0.1 ( (> .g E I l c 0 10 20 30 40 50 60 70 80 90 100 { % of Rated Thermal Power (' ( FIGURE 3.1-0 ( MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL W (' e. McGUIRE - UNITS 1 and 2 3/4 1-Sa Amendment No. (Unit 1) Amendment No. (Unit 2) i i

O 3/4.2 POWER DISTRIBUTION LIMITS AFD 3/4.2.1 AXIAL FLUX DIFFERENCE (GitERI) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within: 6 the allowed operational space defined by Figure 3.2-1 for RAOC operation, a. e or 4 ( M WetI),5$4Hsf 1) - 4 p b. within a i 3" percent target band about the target flux difference during l 7 base load operation. / l APPLICA8ILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION: I For RAOC cperation with the indicated AFD outside of the Figure 3.2-1 a.

limits, 3

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 - O High Trip setpoints to less than or eq'ual to 55% of RATED THERMAL POWER within the next 4 hours. b. For Base Load operation above APLND** with the indicated AXIAL FLUX 3 DIFFERENCE outside of the applicable target band about the target 4 flux difference: ] 1. Either restore the indicated AFD to within the target band ?, limits within 15 minutes, or NO 2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes, j c. ' THERMAL POWER shall not be increased above 50% of RATED THERMAL-POWER unless the indicated AFD is within the Figure 3.2-1 ifmits.

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.

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

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by: a. Monitoring the indicated AFD for each OPERABLE excore channel: 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, 1 and 2. At least once per hour for the first 24 hours after restoring the AFD Monitoring Alarm to OPERABLE status. b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours 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 censidered outside of its limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits. 4.2.1.3 When in Base Load operation, the target axial flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specifiqation 4.0.4 are f not applicable. 4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining 3 the target flux difference y "m- ' '- ' 2.1.2 4 or by linear interpola-J y tion between the most recently easured value and Ttpopeent at the end of 1, cycle life. The provisions of Specification 4.0.4 are"not applicable. j r t ensc.,s,,rra vg e m cod 34NcnoH wo rst not s se Mos.sAmt seq.,ongngers op specopo wrey g/s z,p l O l l McGUIRE - UNITS 1 and 2 3/4 2-la Amendment No (Unit 1) t l Amendment No (Unit 2) l )

l$fl; WTl8E ?)9(rIE i g, __ % WER DISTRIBUTION LIMITS 's (V \\ 3/ N POWER DISTRIBUTION LIMITS 3/4.2 AXIAL FLUX DIFFERENCE (UNIT 2) LIMITING NDITION FOR OPERATION 1 3.2.1 The in cated AXIAL FLUX DIFFERENCE (AFD) shall be maintained with n h the following rget band (flux difference units) about the target flux difference: j a. i 5% for ore average accumulated burnup of less than or e al to ( 3000 MWD / , and 5 b. + 3% -12M fo core average accumulated burnup of greate than f 3000 MWD /MTU. APPLICABILITY: M0;E 1 abov 50% of RATED THERMAL POWER *. ACTION: ( a. With the indicated AFD utside of the above re ired target band 5 about the target flux di erence and with TH L POWER: l 1. Above 90% of RATED TH L POWER, wit n 15 minutes either: a) Restore the indicat d AFD to w hin the target band ( O' limits, or b) Reduce THERMAL POWER to le than 90% of RATED THERMAL POWER. ) 2. Between 50% and 90% of RATE H. MAL POWER: C a) POWER OPERATION ~may ntinue ovided: ( ( 1) The indicate AFD has not en outside of the above ( l required t get band for mor than 1 hour penalty deviatio cumulative during t previous 24 hours, and ( 2) The i icated AFD is within the imits shown on h Fig e 3.2-1. Othentise, reduce ERMAL POWER to less t n 50% of RATED THERMAL POWER wit n 30 minutes and I educe the Power Range Neutron Flux-H h Trip Setpoints to less than or equal to 55% of RATED ERMAL POWER within ti.e next 4 hours, b) urveillance testing of the Power Range Neutron lux channels may be performed pursuant to Specificat n 4.3.1.1 provided the indicated AFD is maintained within th limits of Figure 3.2-1. A total of 16 hours operation may e ( accumulated with the AFD outside of the target band d ing ( this testing without penalty deviation. [ b. THERMAL POWER shall not be increased above 90% of RATED THERMAL O [ / POWER unless the indicated AFD is within the above required target i band and ACTION a.2.a) 1), above has been satisfied. l ( See Special Test Exception 3.10.2. McGUIRE - UNITS 1 and 2 3/4 2-2 Amendment No. (Unit 1) Amendment No. (Unit 2)

NGCTE EWI)tcPMC w gy 7 0WER DISTRIBUTION LIMITS ~ L TING CONDITION FOR OPERATION x ACTIk(Continued) \\ s c. THERMAL POWER shall not be increased above 50% of RATED THE AL POWER unless the indicated AFD has not been outside of th above l required target band for more than I hour penalty deviat~on j umulative during the previous 24 hours. Power increa s above 50% o RATED THERMAL POWER do not require being within t target band pr vided the accumulative penalty deviation is not folated. SURVEILLANCE RE IREMENTS (: ( I 4.2.1.1 The indicat AFD shall be determined to e within its limits during d, POWER OPERATION above 5% of RATED THERMAL POWER y: ( a. Monitoring the indicated AFD for ea OPERABLE excore channel: 1) At least on e per 7 days whe the AFD Monitor Alarm is OPERABLE, ( and ( 2) At least once r hour f the first 24 hours after restoring ) the AFD Monitor arm t OPERABLE status. I ) b. Monitoring and logging t indicated AFD for each OPERABLE excore channel at least once pe ur for the first 24 hours and at least ( once per 30 minutes the eaft , when the AFD Monitor Alarm is ( inoperable. The logg value of the indicated AFD shall be assumed [ to exist during the terval pr eding each logging. b 4.2.1.2 The indicated AFD all be consider outside 3f its target band when two or more OPERABLE excor channels are indic ting the AFD to be outside the target band. Penalty dev ation outside of the t rget band shall be accumulated on a time basis of: a. One minute enalty deviation for each 1 min te of POWER OPERATION ( outside o the target band at THERMAL POWER vels equal to or above ( 50% of TED THERMAL POWER, and b. One-h f minute penalty deviation for each 1 minu of POWER OPERATION out de of the target band at THERMAL POWER levels tween 15% and i 50 of RATED THERMAL POWER. [ ( 4.2.1.3 e target flux difference of each OPERABLE excore channe shall be ( determi d by measurement at least once per 92 Effective Full Power ys. The provi ons of Specification 4.0.4 are not applicable. 4. 1.4 The target flux difference shall be updated at least once per ( O3 Effective Full Power Days by either determining the target flux differen ursuant to Specification 4.2.1.3 above or by linear interpolation between t most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification 4.0.4 are not applicable. McGUIRE - UNITS 1 and 2 3/4 2-3 Amendment No. (Unit 1) Amendment No. (Unit 2) mn-n -p ,-y-

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0g i::: c: # l'l W E -c J. j u.? I E 5 w w (-15,100) (6,100) /( 100 UNACCEPTABLE- ) UNACCEPTABLE OPERATION OPERATION 7 fl 80 ACCEPTABLE OPERATION .._. / _. (( (. 60 50 i } (-31,50 ) (17,50) .... - ~. -. ~. -.... - - -.... -. - - - - . - - - - - - - - ~ - - - - - h t> 20 i 0 (' -50 -40 ,-30 -20 -10 0 10 20 30 40 50 ( Flux Difference (aI)% c', C, FIGURE 3.2-1/ l AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER W O McGUIRE - UNITS 1 and 2 3/4 2-4 Amendment No. (Unit 1) Amendment No. (Unit 2)

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1 1_ l l 1 l t l 1 l 0-i ! ( 40 30 20 10 0 10 20 3 40 50 FLUX DIFFERENCE (Al) % t I FIGURE 3.2-lb ! h y IAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (UNIT 2) l r j \\ I McGUIRE - UNITS 1 and 2 3/4 2-5 Amendment No. (Unit 1) l Amendment No. (Unit 2) l..-

POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION e 3.2.2 F (Z) shall be limited by the following relationships: q 1.26 F (Z) 5 [ ] [K(Z)] for P > 0.5 (Unit 2) q F (Z) 5 [2 15] [K(Z)] for P > 0.5 (Unit 1) q p

i. 16 E

F (Z) 5 3 [K(Z)] for P 5 0.5 (Unit 2) q F (Z) 5 [2 ] [K(Z)] for P $ 0.5 (Unit 1) 9 9 P = THERMAL POWER l ere: 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: q Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit a. n within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K ) have been reduced at least 1% (in AT span) for each 1% F (Z) exceeds the limit; and q t. 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. 9 McGUIRE - UNITS 1 and 2 3/4 2-6 Amendment No. (Unit 1) Amendment No. (Unit 2)

oh SURVEILLANCE REQUIREMENTS UDN K Il l 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: q q O 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 increasina the value by 5% to account for measurement uncertainties, vcher noc 7tre onenc~n.* wceoo isaro.o-s.n.a. now casosrore. Satisfying the following relationship: l r c. F,":,) i z u, y,, g,4 P > c i(...r a) N S xW Fq (z) < W(z) for P > 0.5 N.r 0 ~' cEuv'z)e.rso.5p.cT)) N 2.15 x K(z) for P < 0.5 (u~.ro FQ (z) f( W(z) x 0.5 ' * *' ' M y g M whereF(z)isthemeasuredF(z)increasedbytheallowancegfo,r q g manufacturing tolerances and measurement uncertainty, 2.15"is the F l 4 limit, K(z) is given in Figure 3.2-2, P is the relative THERMAL POW R, and W(z) is the cycle dependent function that accounts 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 q 2. At least once per 31 Effective Full Power Days, whichever occurs first. i i

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.

lO l l McGUIRE - UNITS 1 and 2 3/4 2-7 AmendmentNo.Y(Unit 1) l Amendment No.R (Unit 2)

SURVEILLANCE REQUIREMENTS IID6H::6 (Continued) e. With measurements indicating maximum F (z) over z (K(z) { N has increased since the previous determination of Fq (z) either of the following actions shall be taken: 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 fFM (z)1 is not increasing. (K(z)/ over 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 expression: q [ maximum M ) Fn (z) x W(z) I _y x 100 for P > 0.5 N ~.r 0 ( over z 2.15 m P xK(z) J Q'.wHu!*ratros~wmmc'1's u,r,,,,,, z.. r, u 3. r t y c-(j'fmaximum M Q-K() () -1 x 100 for P < 0.5 (er ') L 'over z 2.15 L x K(z)d s '6 k h mus,- c.carme2 B,;1 0.5 .,.t,ym_,,, c 2) One of the following actions shall be taken: 'f_ a) Within 15 minutes, control the AFD to within new AFD 4 limits which are determined by reducing the AFD limits of 3.2-1 by 1% AFD for each percent F (z) exceeds its limits q as determined in Specification 4.2.2.2f.1). Within 8 hours, reset the AFD alarm setpoints to these modified limits, or b) Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated q above, or d l c) Verify that the requirements of Specification 4.2.2.3 for 6 l Base Load operation are satisfied and enter Base Load j i operation. AmendmentNok((Unit 1 McGUIRE - UNITS 1 and 2 3/4 2-8 Amendment No p Unit 2) l

O (Q. 'SI'RVEILLANCE REQUIREMENTS Lh"E(Continued) . 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. 4.2.2.3 Base Load operation is permitted at powers above APL if the l following conditions are satisfied: 7 ,g a. Prior to entering Base Load o eration, maintain THERMAL PohER above J NO APL and lets than or equal tb that allowed by Specification 4.2.2.2 for at least the previous 24 hours. Maintain Base Load operation I surveillance (AFD within i 3 fof target flux difference) during this l l time period.- Base Load operation is then permitted providing THERMAL NO POWER is reaintained between APL and APL or between APL and 100% (whichever is most limiting) and F surveillance is maintained q APL ' is defined as: O pursuant to Specification 4.2.2.4. 5 APLBL = minimumE (M(Z) x W(Z)gg2.15 x K(Z) ] x 105 W 0 ) over Z F g7p my, mm uu u0 ( N _ A,-sm,amm p q where: F (z) is the measured F (z) increased by the allowances for 'L g q manufagtuQngt{eganc,egandmeasurementuncertainty. The F limit [ q is 2.15. K(z) is given in Figure 3.2-2. W(z)BL is the cycle l f deper. dent 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 NO APL then the conditions of 4.2.2.3.a shall be satisfied before j re-entering Base Load operation. ) l ) 4.2.2.4 During Base Load Operation F (Z) shall be evaluated to determine if F (Z) is within its limit by: q q a. Using.the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APL h t b. Increasing the measured F (Z) component of the power distribution j q map by 3% to account for manufacturing r.olerances and further i increasing the value by 5% to account for measurement uncertainties. Hwr tu nuow, curs at sncorocar~ s.n nu unvote. f., o ~ l l McGUIRE - UNITS 1 and 2 3/4 2 9 Amer.dment No (Unit 1) Amendment No (Unit 2) i

SURVEILLANCE REQUIREMENTS BRN=27 (Continued) { c. Satisfying the following relationship: F (Z) < 2. Z) for P > APLND( rO j 4"y i V 1<, where: (Z) is the measured F (Z). The F limit is 2.13(er,)4w l( Q Q uu-,n) K(Z) is given in Figure 3.2-2. P is the relative THERMAL POWER. ~1W(Z)8L is the cycle dependent function that accounts for limited l ) power distribution transients encountered during normal operation. 3 This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9. d. Measuring (Z) in conjunction with target flux difference deter-mination according to the following schedule: 1. Prior to entering BASE LOAD operation after satisfying Sec. con b 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 prict to mapping, and ( ( 2. At least once per 31 affective full power days. [ With measurements indicating e. F (Z) maximum [ ] ( over Z ( has increased since the previous determination F (Z) either of the f following actions shall be taken: 1. F (Z) shall be increased by 2 percent over that specified in 4.2.2.4.c, or b 2. F (Z) shall be measured at least once per 7 EFPD until 2 7 successive maps indicate that F (Z) ) maximum [ ] is not increasing. ( over (, I f. With the relationship specified in 4.2.2.4.c above not being [ satisfied, either of the following actions shall be taken: ( l 1. Place the core in an equilbrium condition where the limit in M 4.2.2.2.c is satisfied, and remeasure F (Z), or q McGUIRE - UNITS 1 and 2 3/4 2-9a Amendment No. Unit 1) Amendment No. Unit 2)

b)) SURVEILLAhCE REQUIREMENTS IIlltN47(Continued) ) i 2. Comply with toe requirements of Specification 3.2.2 for F (Z) 7 q exceeding its limit by the percent calculated with one of the following expressions: 3 Fl(Z)xW(Z)BL] ) -1 ] x 100 for P > APLNO(w,ri)' r' t's$cknusu~ou [(max. over z of [ Y X K(2) ~ Q116 o~mu ur :..ng rnwar zy ~ 3 F"(Z) x W(Z) g <',2 ([-(max.overzof["._ j,i2a ;;; e-n_4 < P < APL }^*xK(Z) j l g. The limits specified in 4.2.2.4.c, 4.2.2.4.e. and 4.2.2.4.f above j are not applicable in the following core plan regions: 1. Lower core region 0 to 15 percent, inclusive. t 2. Upper core region 85 to 100 percent, inclusive. 4 4.2.2.5 When F (Z) is measured for reasons other than meeting the requirements 4 q of specification 4.2.2.2 an overall measured F (z) shall be obtained from a power q distribution n.ap and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. 4 i l McGUIRE - UNITS 1 and 2 3/4 2-9b Amendment No. Unit 1) Amendment N (Unit 2)

QF JTS MIPE PME ) s 00WER DISTRIBUTION LIMITS N-SU ILLANCE REQUIREMENTS (UNIT 2) / Y 4.2.2.1 The provisions of Specification 4.0.4 are not applicable, j 4.2.2.2 F shall be evaluated to determine if F (Z) is within ts limit by: p x q Usi,q'themovableincoredetectorstoobtainapo r distribution ) a. map a any THERMAL POWER greater than 5% of RATE THERMAL POWER, j b. Increasi the measured F component of the ower distribution map xy by 3% to ac unt for manufacturing toleran s and further increasing the value by to account for measureme uncertainties, } c. Comparing the F computed (F ) obta ed in Specification 4.2.2b., x above, to: g 1) The F limits f RATED THE L POWER (FRTP) for the appropriate xy x measured core plane given n Specifications 4.2.2.2e. and f., j < below, and 2) The relationship. [ F' =F [ 0.2(1 )], R xy l L Where F is th limit for frac ional THERMAL POWER operation k RTP f express as function of F a P is the fraction of RATED x THERMAL POW at which F was meas ed. xy d. Remeasuring F according to the following hedule: 'y y RTP 1) When F is greater than the F limit f the appropriate y x e meas red core plane but less snan the F r lationship, j xy ad itional power distribution maps shall be t en and F depareatoF RTP l and F either: x xy I a) Within 24 hours after exceeding by 20% of RAT THERMAL {' POWER or greater, the THERMAL PCWER at which F was last k determined, or xy ( b) At least once per 31 EFPD, whichever occurs first. (,' l l 9 l McGUIRE - UNITS 1 and 2 3/4 2-10 Amendment No. (Unit 1) Amendment No. (Unit 2)

O ffE E N ftM( P M g ' POWER DISTRIBUTION LIMITS I' }. S VEILLANCE REQUIREMENTS (UNIT 2) RTP 2) When the F is less than or equal to the F limi or the x appropriate measured core plane, additional powe distribution y, C l maps shall be taken and F*Y compared to F*RTP d F*Y at least j once per 31 EFPD. Y RTP e. The F imits for RATED THERML POWER (F shall be provided for xy x all core p nes containing Bank "D" contr rods and all unrodded core planes 'n a Radial Peaking Factor mit Report per Specifi-cation 6.9.1. / f. The F limits o pecificatio. 4 .2.2e., above, are not applicable xy in the following cor planes r ions as measured in percent of core height from the bottom f the uel: 4 L 1) Lower core region fr 0 to 15%, inclusive, 2) Upper core region rom 8 to 100%, inclusive, 3) Grid plane re 'ons at 17.8 + , 32.1 + 2%, 46.4 + 2%, 60.6 + 2% and 74.9 2%, inclusive!a f ~ ~ ~ O s i 4) Core pla regions within + 2% of e height (+ 2.88 inches) b about e bank demand position of th Bank "0" control rods. } C g. With F exceeding F , the effects of F n (Z) shall be xy eval ted to determine if F (Z) is within its li ts. 9 4.2.2.3 W n F (Z) is measured for other than F determinatio , an overall q xy measur F (Z) shall'be obtained from a power distribution map and ' creased ) q by 3 to account for manufacturing tolerances and further increased by % to ac ount for measurement uncertainty. I 1 I~ l l McGUIRE - UNITS 1 and 2 3/4 2-11 Amendment No. (Unit 1) Amendment No. (Unit 2)

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.s m POWER DISTRISUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR. LIMITING CONDITION FOR OPERATION 3.2.3 The: combination of indicated Reactor Coolant System (RCS) total flow rate and Rx lif shall be maintained within the region of allowable operation f shown on Figure '3.2-3 for four loop operation: Where: i L Fh 1 X

1.49 [1.0 + 0.3 (1.0 - P)] lR --

[1.0 + M y 1 J. .k A a (Unit n ; e 7 =- (Un;; W - Q "U)j - p THERMAL POWER __' RATED THERMAL POWER ', aus .. b. y. P g=MeasuredvaluesofFhobtainedbyusingthemovableincore f c, )(. F detectors to obtain a power distribution map. The measured values of F shall be used to calculate R since Figure 3.2-3 H includes penalties for undetected feedwater venturi fouling of 0.1% and for measurement uncertainties of 1.7% for flow and 4% for incore measurement of Fh,')m6 [ y-@D (RUl = Rod Bow Penalty as a. function of region p rgce buraep-st' shown in F+gure 3 2-4,whera a rq vo is defined as those assamh'f&ww. cne same loaaing ud-ds) or enrich-e (first carat (Anolies tn Unit 2 oniv). APPLICABILITY: MODE 1. ACTION: Witithe combination of RCS total flow rate and Rg M outside the region of ccceptable operation shown on Figure 3.2-3: l l L. Within 2 hours either: L 1. Restore the combination of RCS total flow rate and Rg W to within the.above limits, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. McGUIRE - UNITS.1 and 2 3/1 2-14 Amendment No. (Unit 1) Amendment No. (Unit 2)

~ POWER DISTRIBUTION LIMITS i LIMITING CONDITION FOR OPERATION <[S ACTION: '(Continued) ~ b.

Within 24 hours of initially being.outside the above limits, verify.

'through incora flux mapping and RCS total flow rate comparison that' the combination of Rx 5(, and RCS total flow rate are restored to j within the at,ove limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL. POWER within the next 2 hours. Identify and correct the cause of the out-of-limit condition prior c. to increasing THERMAL POWER above the reduced: THERMAL POWER limit 4 required by ACTION.a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of Rx ll( and indicated } RCS total _ flow rate are demonstrated, through incore flux mapping-and RCS total flow rate comparison, to be within the region of -l acceptable operation shown on Figure 3.2-3 prior to exceeding the i i following THERMAL POWER levels: 1. A nominal 50% of RATED THERMAL-POWER, 2. A nominal 75% of RATED THERMALPOWER,-and

f 4

3. Within 24 hours of attaining greater than or equal.to 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The combination of indicated RCS total flow rate determined by ^ process computer readings or digital voltmeter measurement and Rx itodJhr l shall be within the region of acceptable operation of Figure 3.2.3: a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and I b. At least once per 31 Effective Full Power Days. ^; 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per'12 hours when the most recently obtained value/ of RyitDd.4M obtained per Specification { 4.2.3.2, are assumed to exist. i 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. i 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. 1 4 McGUIRE - UNITS 1 and 2 3/4 2-15 Amendment No. ' (Unit 1) Amendment No. (Unit 2) i . _. _ _. _... _, _ _ _ _,,., _.,, _,. _.., _.. ~ _ _ _ _ - -. - - _. _ _ _ _.. _ _. -... _ _ _. _ _ _. -

O 1 ( Y E PENALTIES OF 0.1% FOR UNDETECTED FEED- ~ f 46 WATER VENTURI FOULING AND MEASUREMENT g UNCERTAINTIES OF.1.7% FOR FLOW AND 4% ~ e FOR INCORE MEASUREMENT OF F ARE AH c INCLUDED IN THIS FIGURE. h CEPT E 3 O ON 7 i l H a. RE FOR { w 0 SON 5. b } ~ ACCEPTABLE a: OPERATION 2 S 42 REGION W 9 M y i e UNACCEPTABLE m s F OPERATION ~ u) HEGION 7 O C ACCEPTABLE OPERATION REGION FOR 598% RTP (1.0, 38.888) a 596% RTP (1.0, 38.499) i 38 594% RTP (1.0, 38.110) YE $92% RTP (1.0, 37.7 2'11 l kk (1.0,37.33h) 590% RTP gg (1.0, 4) j gg - 3 6.9 v4 2z 36 P. 0.09 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 I 22 03 I 11.49(1.0 +M (1.0-P)] j 3, 3, Rg = FAH \\ y'Mn i Figure 3.2-3t' RCS FLOW RATE VERSUS Ry Jiid-( - FOUR LOOPS IN OPERATION (Unit 2)

l\\ s Il ( ( k, ,l ~~,s.a 8. Overpower AT $ 6.0" seconds~~

9.
. Pressurizer Pressure--Low 1 2.0 seconds 10. Pressurizer Pressure--High $ 2.0 seconds l 22 ig 11. Pressurizer Water Level--High N.A. ! !Z I a; !fg Neutron detectors are exempt from re'sponse time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel. j ' iT 3, !,
TABLE 3.3-2 (Continued) Ng REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES E m a FUNCTIONAL UNIT RESPONSE TIME Eq 12. Low Reactor Coolant flow m H a. Single Loop (Above P-8) $ 1.0 second E; b. Two Loops (Above P-7 and below P-8) 1 1.0 second g u~.r s), 3. C ( wo r 2.) m 13. Steam Generator Water Level--Low-Low 1 2.0" seconds 14. Undarvoltage-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 l N.A. w 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. IY rP $$?I ! f, 77 =t O O k ----- - - -
g EBLE 3.3-4 (Continued) n5 ENGINEERED SAFETY FEATURES AtliUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS A h FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 5 d 7. Auxiliary Feedwater ~ s a. Manual Initiation N.A. N.A. o, E b. Automatic Actuation Logic N.A. H.A. u and Actuation Relays c. Steam Generator Water Level--Low-Low N-e r e ), " c N"" O (w-e r o, 39 oMv-ir a) 1) Start Motor-Driven Pumps 2 12% ( span from 0 to 2 11% o span from 0 to 30%ofRpTEDTHERMALPOWER, 30% of R TED THERMAL POWER, w increasi increast linearly to 1 54.9%'pg linearly to ) of span at 100% of 2 53.9% of span at 100% of j w 4 RATED TilERMAL POWER. RATED THERMAL POWER. = 2) Start Turbine-Driven Pumps 1 12% of span from 0 to 2 11% of span from 0 to 30% of RATED THERMAL POWER, 30% of RATED THERMAL POWER, increasing linearly to increasing linearly to 2 54.9% f span at 100% of 2 53.9% f span at 100% of f RATED Tl RMAL POWER. RATED Til RMAL POWER.
~~~'r O, '1o. v.M, r i y k~.r a,11,o x (s.~st i) d.~~
Auxiliary feedwater 1 2 psig 1 1 psig Suction Pressure - Low (Suction Supply Automatic Realignment) sz Id e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints jj Start Motor-Driven Pumps and Allowable Values ri Motor-Driven Pumps and 8.5
~~0.5 second time~~
-> 3200 volts f. Station Blackout - Start 3464 i 173 volts with a Turbine-Driven Pump delay 7-g. Trip of Main Feedwater Pumps - N.A. ~ N.A. {j Start Motor-Driven Pumps 9 e e
n.'. A l INSTRUMENTATION MOVA8LE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERA 8LE with: a. At least 75% of the detector thimbles, j b. A minimum of two detector thimbles per core quadrant, and Sufficient movable detectors, drive, and readout equipment to map c. these thimbles. APPLICA8ILITY: When the Movable Incore Detection ' System is used for: a. Recalibration of the Excore Neutron Flux Detection System, b.- Monitoring the QUADRANT POWER TILT RATIO, or i o Measurement of Fh,w"F (Z)._% c. q l ACTION: With the Movable Incore Detection System inoperable, do not use the system for .i j the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. i l i i i SURVEILLANCE REQUIREMENTS \\ f 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERA 8LE at least once per 24 hours by normalizing each detector output when required for: i i a. Recalibration of the Excore Neutron Flux Detection System, or b. Monitoring the QUADRANT POWER TILT RATIO, or N *** c. Measurement of Fg, F (Z) K. q i I" ~'."r Q McGUIRE - UNITS 1 and 2 3/4 3-45 M8"""~'"A ) k~ Amnemnr wo. .-~,.--,,,,.---n.-. --,n,-n,,na.me.--.,,,mn .a...,,,,,,n.,,.,,,n-,,- _n ,e.--,,_,,,,,m,._.mn,,,n-
rO l -q # 3/4.5 EyRGENCYCORECOOLINGSYSTEMS ~ i 3/4.5.1 ACCUMULATORS COLD LEG INJECTION I LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERA 8LE with: The isolation v'alve open, a. b. Acontainedboratedwatervolumeofbetweeng N K 8022 and 8256 gallons /U^ 't Z p y -1 _.._ G% v.; ;.... s _..-.J.1 i, A ' boron concentration of between 1900 and 2100 ppe, 1 c. s d. A nitrogen cover pressure of between 430 and 484 psig IDn64dJ, au, j ~evu - = ;-- 5, G..:. 2), - - D e. A water *1evel and pressure' channel OPERA 8LE. APPLICABILITY: M00ES 1, 2, and 3*. ACTION: With one cold leg injection accumulator inoperable, except as a result a. of a closed isolation valve, restore the inoperable accumulator to i OPERA 8LE status within 1 hour or be in at least HOT STAN08Y within I the next 6 hours and in HOT SHUTOOWN within the following 6 hours. b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isola-tion valva or be in at least HOT STANOBY within 1 hour and in HOT SHUTOOWN within the following 12 hours. SURVEILLANCE REQUIREMENTS 4 4 4.5.1.1.1 Each c'old leg injection accumulator shall be demonstrated OPERABLE: - a. At least once per 12 hours by: t 1) Verifying the contained borated water volume and nitrogen i covei pressure in the tanks, and i 2) Verifying that each cold leg injection accumulator isolation valve is open.
~~Pressurizer pressure above 1000 psig.~~
McGUIRE - UNITS 1 and 2 3/4 5-1 - Amendment No. (Unit 1) ) Amendment No. (Unit 2) P
EMERGENCY CORE COOLING SYSTEMS UPPER HEAD INJECTION LIMITING CONDITION FOR OPERATION 3.5.1.2.Each Upper Heac. Injection Accumulator System shall be OPERt.8LE.with: a. The isolation valves open, b. The water-filled accumulator containing a minimum of 1850 cubic feet of borated water having a concentration of between 1900 and 2100 ppm of boron, and c. The nitrogen bearing accumulator pressurized to between 1206 and 1264 psig. APPLICABILITY: MODES 1, 2 and 3.* y, p,m ' ACTION: Whr @ Above 46% RATED THERMAL POWER: 'l th the Upper Head Injection Accumulator System inoperable, ept ) a. as sult of a closed isolation valve (s), restore the U) Head l Injectio ecumulator System to OPERABLE status withi Tour or be d at less than equal to 46% RATED THERMAL POWER close the isola-tion valves withi he next 6 hours. S b. With the Upper Head Inj on Accum r System inoperable due to the isolation valve (s) being , either immediately open the isolation valve (s) or be ess or equal to 46% RATED THERMAL POWER and close the aining isolat alves within 1 hour. g Less than or al to 46% RATED THERMAL POW s a. With t pper Head Injection Accumulator System in able, POWER OPE ON may continue provided the isolation valves ar losed hin 6 hours. } The provisions of Specification 3.0.4 are not applicable. p b SURVEILLANCE REQUIREMENTS 4.5.1.2 Each Upper Head Injection Accumulator System shall be demonstrated OPERABLF: a. At least once per 12 hours by: 1) Verifying the contained barated water volume and nitrogen pressure in the accumulators, and 2) Verifying that each accumulator isolation valve is open. " Pressurizer Pressure above 1900 psig. Amendment No. (Unit 2) McGUIRE - UNITS 1 and 2 3/4 5 3 Amendment No. (Unit 1) )
c- ~ f N h (/L T Y With the Upper Head Injection A.cc.umulator System inoperable, except a. as a result of a closed isolation valv2(s), restore the Upper Head Injection Accumulator System to OPERABLE status within 1 hour or be lh in at least HOT STAN08Y within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With the Upper Head Injection Accumulator System inoperable due to the isolation valve (s) being closed, either immediately open the isolation valve (s) or be in HOT STANOBY within 1 hour and be in HOT SHUTDOWN within the next 12 hours.
C(3 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1-BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTOOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN I,, MARGIN of 1ex - :;i.. m/L J " & 1.3% delta k/k (I!n4=0 is required to l4 control the reactivity transient. 5 Accordingly, the SHUT 00WN MARGIN requirement is based upon this limiting S condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection. 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting l condition assumed in the FSAR accident and transient analyses. i The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MOC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MDC associated with a core condition of all rods inserted (most positive MOC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.1 x 10 4 delta k/k/*F. The MTC value of -3.2 x 10 4 delta k/k/*F represents a conservative value (with O corrections for burnup and soluble boron) at a core condition of 300 ppm j V equilibrium boron concentration and is obtained by making these corrections to l the limiting MTC value of -4.1 x 10 4 k/k/*F. Amendment No. (Unit 1) McGUIRE - UNITS 1 and 2 B 3/4 1-1 Amendment No. (Unit 2)
REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS baron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature. NDT 3/4.1.2 80 RATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators, g With the RCS average temperature above 200*F, minimum of two boron injection flow paths are required to ensure singl functional capability in the event an assumed failure renders one of the ow paths inoperable. The boration capability of either flow path is suff cient to provide a SHUTDOWN MARGIN from expected operating conditions of 3>61 delta k/k af ter xenon decay l and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppm borated water from the boric acid storage tanks or 75,000 gallons of 2000 ppm borated water from the refueling water storage tank (RWST). With the RCS temperature below 200*F, one Boron Injection Systera is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable. The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 300*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. I McGUIRE - UNITS 1 and 2 B 3/4 1-2 1 (( ~ 2
3/4.2 POWER O!$TRIBUTION LIMITS j-BASES V I The specifications of this section provide assurance of fuel integrity during Condition I (Nomel 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 j the fission gas release, fuel pellet temperature, and cladding mechanical prop-erties to within assumed design criteria. In addition, limit' ng 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 i limit of 2200*F is not exceeded. I. 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 9 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 - l ances on fuel pellets and rods; i I Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of I the integral of linear power along the rod with the highest integrated i power to the average rod power; and "- d ; r., tar. is defined as the ratta af,---' ;nr 2. g y( [ i k m; 4 } to aver = -- -- 27. 2, in T,no nor m n w; ;?--- at care elevation Z ) de ) 3/4.2.1 AXIAL FLUX OIFFERENCE i 2.24, t Thelimitson[AXIALFLUXDIFFERENCE(AFD)assurethattheFQ(Z) upper bound envelope of W(Unit 2), 2.15 (Unit 1) times the normalized axial l I 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 i their respective insertion limits and should be inserted near their normal j position for steady-state operation at high power levsis. The value of the target flux difference obtained under these conditions divided by the fraction p l of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER P l for the associated core burnup conditions. Target flux differences for other ( l l THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value ) by the appropriate fractienal THERMAL POWER level. The periodic updating of i i the target flux difference value is necessary to reflect core burnup J t l considerations. i O j McGUIRE - UNITS 1 and 2 8 3/4 2-1 Amendment No (Unit 1) Amendment No (Unit 2) +
2 POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) \\~ Although it is intended that the plant will be operated with the AFD thin the target band required by Specification 3.2.1 about the target f x di rence, during rapid plant THERMAL POWER reductions, control rod on will use the AFD to deviate outside of the target band at reduced RMAL F0WER le 1s. This deviation will not affect the xenon redistrib on suffi-ciently to nge the envelope of peaking factors which may be ached on a subsequent ret to RATED THERMAL POWER (with the AFD with the target band) provided the time uration of the deviation is limited. cordingly, a 1 hour penalty deviation 1 t cumulative during the previou 4 hours is provided for operation outside of th target band but within th imits of Figure 3.2-1 while at THERMAL POWER lev s between 50%.and 9 of RATED THERMAL POWER. For THERMAL POWER levels bet 15% and 5 RATED THERMAL POWER, devia-g tions of the AFD outside of the get b are less significant. The penalty of 2 hours actual time reflects th e uced significance. Provisions for monitori e AFD on automatic basis are derived from the plant process comput3rd.pghrough the AFD Monr Alarm. The computer deter-mines the 1 minute aye(age of each of the OPERA 8L core detector outputs and provides an ala inessage immediately if the AFD for or more OPERABLE excore channel are outside the target band and the THE POWER is greater than 90% of TED THERMAL POWER. During operation at THERMA WER levels I between and 90% and between 15% and 50% RATED THERMAL POWER, computer output an alarm message when the penalty deviation accumulates beyo the lim of 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band, r Fiirhi&C 4t power levels below APLNO, the Ifmits on AFD are defined by S Figures 3.2-1, i.e. that defined by the RAOC operating procedure and limits, t,These limits were calculated in a manner such that expected operational s transients, e.g. Ioad follow operations, would not result in the AFD deviating i outside of those limits. However, in the event such a deviation occurs, the 1short period of time allowed outside of the limits at reduced power levels dwill not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to prevent operation in the vicinity 5 HO }oftheAPL power level. lAt power levels greater than APLND, two modes of operation are ermissible; j
~~51) RAOC, the AFD limit of which are defined by Figure 3.2-1, and 2) Bas, Load ioperation, which is defined as the maintenance of the AFD within a i3% band 6~~
Ljabout a target value. The RAOC operating procedure above APLND is the same as j (that defined for operation below APLND However, it is possible when j l [ following extended load following maneuvers that the AFD limits may result in i srestrictions in the maximum allowed power or AFD in order to guarantee poperationwithF(z)lessthanitslimitingvalue. To allow operation at the p ) g maximum permissible value, the Base Load operating procedure restricts the 1 McGUIRE - UNITS 1 and 2 8 3/4 2-2 Amendment No. (Unit 1) Amendment No (Unit 2)
(~] POWER DISTRIBUTION LIMITS V 8ASES AXIAL FLUX OIFFERENCE (Continued) w,r n,1 s%ts ~.n) Qndicat FD to relatively small target band and power swings (AFD target 5 iband of i,APLE < power < APL8L or 100% Rated Thermal Power, whichever is 1 3 f, lower). For Base Load operation, it is expected that the plant will operate ? within the target band. Operation outside of the target band for the short J [ time period allowed will not result in significant xenon redistribution such ) pthat the envelope of peaking factors would change sufficiently to prohibit ? dcontinued operation in the power region defined above. To assure there is no
~~> residual xenon redistribution impact from past operation on the Base Load doperation, a 24 hour waiting period at a power level above APLNO and allowed I,~~
j ty RAOC is necessary. During this time period load changes and rod motion are S ' restricted to that allowed by the Base Load procedure. After the waiting 5 i period extended Base Load operation is permissible. [ FJcWe+t'1, The computer determines the one minute average of each of the joPERA8LE excore detector outputs and provides an alarm message immediately if ? pthe AFD for at least 2 of 4 or 2 of 3 OPERA 8LE excore channels are:
~~1) outside c,'~~
~~i i the allowed AI power operating space (for RAOC operation), or 2) outside the Lallowed AI target band (for Base Load operation). These alarms a,e active ?" O1 when power is greater than:~~
~~1) 50% of RATED THERMAL POWER (fo 'RAOC operation),~~
~~j or 2) APLE (for Base Load operation). Penalty deviation.ainutes for Base Load k~~
operation are not accumulated based on the short per % d of time during which J

operation outside of the target band is allowed.
[ t 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT PHANM L FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The Ilmits 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 DN8R are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS accep-tance criteria Ifmit. 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 insure that the limits are maintained provided: I r i O l McGUIRE - UNITS 1 and 2 33/4 2-2a Amendment No. (Unit 1) AmendmentNo.g(Unit 2)
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-I i. i i, i x 0_-30% -20% -10% 0 +10% +20% +30% A INDICATED AXIAL. Ft.UX DIFFERENCE (v) FIGURE B 3/4 21. TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSU5i THERMAL POWER I @ b4 h s s,,e A,o. ( w ae a r g ) McGUIRE - UNITS 1 and 2 8 3/4 2-3 w %,, ~,, %,c o
c ,o POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE, HOT CHANNEL FACTOR (Continued) Control rods in a single group move together with no individual rod a. insertion differing by.more than + 13 steps from the group demand position; b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are. maintained; and d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. F will be maintained within its limits provided Conditions a. through H
~~d. above are maintained.~~
As noted on Figure / 3.2-3 WM, RCS flow rate r~rr and may be " traded of f" against,ys onen, (i.e., a low measured RCS flow one another ro-54 6-rate is acceptable if the m:= r; F g !: M: 1:w) to ensure that the calcu-lated DNBR will not be below the design DNBR value. The relaxation of F as gg a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. Rg as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts l for F less than or equal to 1.49. This value is used in the various accident g N analyses where F influences parameters other than DNBR, e.g., peak clad tem-perature, and thus is the maximum "as measured" value allowed. @ef4nsIC ~ llows for the inclusion of a penalty for Rod Bow on DNBR only. Thus, known the easured" values of F and RCS flow allows for " tradeoffs" i xcess of R equal to r the purpose of offsetting the Rod Bow DN enalty. Fuel rod bowing reduces t lue of DNB ratio. edit is available to partially offset this reduction. Th dit s from a generic or plant-specific design margin. For McGuire Uni margin used to partially I, offset rod bow penalties is 9.1%. s margin br down as follows 1) Design limit ON 1.6% 2) Grid spaci 2.9% i 3) The Offfusion Coefficient 1.2%
~~4) ) Pitch Reduction R Multiplier 1.7%~~
W 1.7% J McGUIRE - UNITS 1 and 2 8 3/4 2-4 Amendment No. (Unit 1) Amendment No (Unit 2)
~ POWER DISTRIBUTION LINITS BASES HEAT FLtd HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR EllTHALPY RISE HOT CHANNEL FACTOR (Continued) . ver, the margin used to partially offset rod bow penalties is with the 3.2% used to trade off against measured f1 s much p as 2% lower than the , flow plus uncerta penalties applied toFhtoaccountforrodbow(Fi as a function of burnup are consistent with t in Mr. John F. - ( * ) letter to T. M. Anderson ("r^.n.y.euse) dated April 5,1979 with the differe M t of margin each unit uses to partially offset rod bow cenalties. l ~ t N 6, Margin between the safety analysis limit DNBRs~(1.47 l .and 1.49 for th%1e and typical cells, respectively) and the design limit l DN8Rs (1.32 and 1.34 for thimble and typical cells, respectively) is maintained. A fraction of this margin'is utilized to accommodate the transition core DNBR-penalty (2%) and the appropriate fuel rod bow DNBR penalty (WCAP - 8691, Rev. 1) When an F measurement is taken, an allowance for both experimental error q and m9nufacturing tolerance must be made. An allowance of 5% is appropriate.- O for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. When RCS flow rate and F are measured, no additional allowances are necessary prior to comparison with the Ilmits of Figure / 3.2-3,]I8t:Aer0 l Measurement errors of 1.7% for RCS total flow rate and 4% for Fh have been allowed for in determination of the design DNOR value. The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate i indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-3. Any fouling which might i bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, i action shall be taken before performing subsequent precision heat balance i measurements, i.e., either the effect of the fouling shall be quantified and l compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. i The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3. P McGUIRE - UNITS 1 and 2 '8 3/4 2-5 Amendment No. (Unit 1) Amendment No. (Unit 2)
l ,q ADMINISTRATIVE CONTROLS V RADIAL PEAKING FACTOR LIMIT REPORT . 9.1. 9 The F limit for RATED THERMAL POWER (FRTP) shall be provide o xy x ,the nal Administrator of the NRC Regional Office, with a to the ? Director, r Reactor Regulation, Attention: Chie e Performance [ Branch, U. S. Nuc Regulatory Cormnission, W on, D.C. 20555 for all > core planes containing "D" control and all unrodded core planes at }wouldbesubmittedatso 1 east 60 days prior to cycle criticality. In the event that the limit er tim ing core life, it shall be submitted f, 60 days prior to the e the limit would effective unless otherwise exempted by the ission. RTP Any 1 ation needed to support F wil,1 be by request from t and x ed not be included in this report. j 7 c,q.1,9 4The W(z) functions for RAOC and Base Load operation and the value for APLND 7(as required) shall be provided to the Director, Nuclear Reactor Regulations, ( / Attention: Chief, Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20S55 at least 60 days prior to cycle initial criticality. lIn the event that these values would be submitted at some other time during pore life, it will be submitted 60 days prior to the date the values would [ become effective unless otherwise exempted by the Commission. ND Any information needed to support W(z), W(z)BL and APL will be by request j from the NRC and need not be included in this report. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office vrithin the time period specified for each report. McGUIRE - UNITS 1 and 2 6-21 Amendment No. Unit 1) Amendment No. Unit 2)
.m j.- 'ATTACHMgNT 2 JUSTIFICATION AND SAFITY ANALYSIS 4 Mr. H. B. Tucker's (DPC) November 14, 1983 letter to Mr. H. R. Denton (NRC/ONRR) K described planned changes in the fuel design for McGuire Nucleat' Station, Units 1 L and 2. McGuire Unit 2 has been operating with a Westinghouse _17x17 low-parasitic (STD) fueled core. It'is planned to refuel Unit 2 with Westinghouse 17x17 ( Reconstitutable Optimized Fuel Assembly'(OFA) regions.. As a result, future core-4 ' loadings would range from an approximately 1/3 0FA - 2/3 STD transition core to j. 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. The OFA fuel has similar design features compared to the STD fuel which has had substantial operating experience in a 5 number of nuclear plants. The major differences are the use of six intermediate (mixing vane),Zircaloy grids for the OFA fuel versus six intermediate (mixing vane) l Inconel grids for STD fuel and a reduction in fuel rod diameter. Major advantages j-for utilizing the OFA are (1) increased efficiency. of t.he core by raducing the amount of parasitic material and (2) reduced fuel cycle costs due.to an optimiza-tion of the water to uranium ratio. 1 The above letter provided a Reference Safety Evaluation Report summarizing the evaluation / analysis performed on the region-by-region reload transition from the l McGuire Units 1 and 2 STD fueled cores to cores with'all optimised fuel. The i 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 i core. The evaluation considered the standard reload design methods described in j WCAP-9272 and 9273, " Westinghouse Reload Safety Evaluation Methodology " and the i transition effects described for mixed cores in Chapter 18 of WCAF-9500-A, j " Reference Core Report - 17x17 Optimized Fuel Assembly." Consistent with the Westinghouse STD reload methodology for analyzing cycle specific reloads, 4 I parameters were chosen to maximize the applicability of the transition evalua-L tions for each reload cycle and to facilitate subsequent determination of the f applicability of 10 CFR 50.59. Subsequent cycle specific reload safety evalua-j tions will verify that applicable safety limits are satisfied based on the reference evaluation / analyses established in the reference report. A summary of { the mechanical, nuclear, thermal and hydraulic, and accident evaluations for the j 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 4emonstration i 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 /NIU respectively during the first quarter of 1984, and two others completed their second cycle of irradiation l with a burnup of 22,000 MWD /NIU and were operating in their third cycle. All demonstra-J tion 17x17 0FAs examined were in good or excellent condition. This provides l 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, Maanshan L Unit I was scheduled to begin irradiating a full core of 17x17 0"As during the first half of'1984, and McGuire Unit I has operated nearly a full cycle with an l OFA reload region (60 17x17 0FA assemblies). 5 f ) = /
m ^ c.; X + l i Pge 2 The results of evaluation / analysis and teste described in the Reference Safety Evaluation Report lead to the following conclusions: .a. The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are machen-ically' compatible with'the current STD design, control rods, and reactor-internals interfaces. Both fuel. assemblies satisfy the current design 4 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 j management effects. c. The reload 0FAs are hydraulically compatible with the current STD design. j I d. The accident analyses for the OFA transition core were shown to provide j acceptable results by meeting the applicable criteria, such as, minimus j DNBR, peak pressure..and peak clad temperature, as required. The previously reviewed and licensed safety limits are 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 i and 2 FSAR impacted by the proposed changes as determined in Section 6.0 of ) the Reference Safety Evaluation Report. The information contained within j was prepared using the NRC Standard Format and Content Guide, Regulatory i Guide 1.70, Revision 3 as it applies to McGuire Nuclear Station Units 1 and j 2). Plant operating limitations giver. in the Technical Specifications affected by e. } use of the OFA design and positive NFC will be satisfied with the proposed changes noted in Section 7.0 of the report. I f A is the cycle-specific Reload Safety Evaluation (RSE) for McGuire i Unit 2/ Cycle 2 including Fq surveillance and RAOC/ Base Load Technical Specifica- { tions. The RSE presents an evaluation for McGuire Unit 2. Cycle 2, 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, "Vestinghouse Reload Safety Evaluation Methodology." In addition, the NRC has previously approved a similar OFA reload for McGuire Unit i via Ms. E. G. Adensan's (NRC/0NRR) April 20, 1984 letter to H. B. Tucker (note that j base load operation technical specifications were previously approved for Unit 1 } by Ms. Adensan's letters dated June 21 and September 13, 1984). McGuire Unit 2 is operating in Cycle I with all Westinghouse 17x17 low parasitic [ (STD) fuel assemblies. For Cycle 2 and subsequent cycles, it is planned to refuel ~ the McGuire Unit 2 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 a a
.^ ,.s 7k -Q Y } p.. ' E
~~-Q Fage 3 submittal justified the compatibility of the OFA design with the STD design in~~
'a transition core as well as a fu11.0FA core.. The OFA transition licensing sub- . mittal contained mechanical, nuclear, thermal-hydraulic, and accident evaluations ' which are applicable to the Cycle 2 safety evaluation. ~ All of the accidents comprising the licensing bases which could. potentially be affected by the fuel reload have been reviewed for the Cycle 2 design described 4 herein.. The results of new analyses are included in the above-eentioned c licensing submittal and in the cycle-specific Reload Safety Evaluation, and the justification 'or the applicability of previous results for the rensining j analyses is prasented. 2 The McGuire Unit 2. Cycle 2 reactor core will be comprised of 193 fuel j assemblies arrassed in the core loading pattern configuration shown in Figure 1 j of the Cycle 2 Reload Safety Evaluation. During the cycle 1/2 refueling, 60 STD fuel assemblies will be replaced with 60 Region 4 optimized fuel assemblies. A i summary of the Cycle 2 fuel inventory is given in Table 1 of the Cycle 2 Reload i Safety Evaluation. l From the evaluation presented in the Cycle 2 Reload Safety Evaluation -it is con-l cluded that the Cycle 2 design does not cause the previously acceptable safety j limits to be exceeded. This conclusion is based on the following l 1. Cycle 1 burnup is between 14400 and 15400 IGfD/NTU. j j 2. Cycle 2 burnup is limited to 10700 IGiD/WrU including a coastdown. 1 l 3. There is adherence to all plant operating limitations given in the Technical Specifications as revised by the proposed changes submitted in support of the OFA transition licensing submittal and the changes given in Appendix A of the Cycle 2 RSE. l To ensure plant operation consistent with tho' design and safety evaluation con-4 I clusion statemenen made in the cycle 2 RSE and to ensure that these concluaions remain valid, several Technical Specifications changes will be needed for cycle l' 2. These changes are those outlined in Section 7.0 of the OFA transition licensing } submittal and the changes given in Appendix A of the cycle-specific RSE. Differences between the cycle-specific RSE Technical' Specification changes to those given in the 0FA transition licensing submittal are discussed in the cycle-specific RSE.'along with any necessary justifications. In addition to these changes. Technical j Specification 3.5.1.2 is revised to reflect the fact that the analysis performed to i allow operation at less than or equal to 46% rated theresi power with the upper head injection accumulator system inoperable which was the bases for a recent Technical Specification change (Amendment Nos. 37 and 18 to McGuire Nuclear Station ] Units 1 and 2 Facility Operating Licenses NFF-9 and NFF-17, respectively) is valid 1 only for the STD fuel design, and thus will not be applicable once the OFA reload occurs. Consequently, the specification is revised back to the way it was prior to Amendment Nos. 37/18. Note that Amendment Nos. 37/18 inadvertently revised the specification to be applicable to both Units 1 and 2 although McGuire Unit 1
3.. a Page 4 has already had an 0FA r(load, invalidating the change for Unit 1 (i.e. the specification should have indicated that the change applies to Unit 2 only). Therefore, the revision changes the application to Unit 1 also (this change is conservative). In ths interim the additional provisions of Amendment Nos. 37/18 will not be applied to McGuire Unit 1 through the use of administrative controls. Attachment 1 provides copies of these specifications as they presently appear in the McGuire Units 1 and 2 Technical Specifications with the appropriate changes noted. Certain changes are made such that they affect McGuire Unit I as well as Unit 2 (as opposed to indicating that they apply to Unit 2 only), but these constitute only administrative-type changes (correc-tions of minor errors / typos, clarifications, etc.) or are improvements incorporated for the Unit 2 specifications which are more conservative than the existing Unit 1 specifications. There are no changes which solely affect Unit 1. The Peaking Factor Limit Report for McGuire Unit 2/ Cycle 2 which will be submitted in accordance with the proposed Unit 2 Technical Specification 6.9.1.9 as given in Attachmont 1 provides the elevation dependent W(t) values that are to be used as inputs to define the appropriate fitting coefficients for W(z) and axial eleva-interpolations to be performed as a function of cycle burnup)iD. tion for RAOC and Base Load Operation, and the value for API The appropriate W(z) function is used to confirm that the Heat Flux Hot Channel Factor. FQ(z), will be limited to the values specified in the Technical Specifications. L
(_. _ _ _ 4J o \\, I r s / ATTACMMEIT 2A 4 RELOAD 5AFETY EVALUATION MCGUIRE NUCLEAR STATION UNIT 2 CYCLE 2 i Octobor, 1984 ) / Edited by: P. Schueron .Y i Approved: L. N'. Boman, Acting Manager Thermal Hydraulic Design Nuclear Fuel Dtvtsion 1734L:6/841105 f I
{.. ' 1 3 .t t-TABLE OF CONTENTS Title Pye
~~1.0 INTRODUCTION~~
~~AND~~
~~SUMMARY~~
1 1.1 Introduction 1 1.2 General Description 2 1.3 Conclusions 2 2.0 REACTOR DESIGN 3 2.1 Mechanical Design 3 f 2.2 Nuclear Design 3 2.3 Thermal and Hydraulic Design 5 3.0 POWER CAPA81LITY AND ACCIDENT EVALUATICN 6 1* l 3.1 Power Capability 6 i l' 3.2 Accident Evaluation 6 3.2.1 Kinetic Parameters 7 3.2.2 Control Rod Worths 7 ( 3.2.3 Core Peaking Factors 7 3.3 Reduced RCS Flow 8 j 3.4 LOCA Analysis 15 4.0 TECHNICAL SPECIFICATION CHANGES 18
~~5.0 REFERENCES~~
19 APPEN0!X A - Technical Specification Page Changes i 1734L:6/841105
I4 c. LIST OF TA8LES Table Tit 13 , Pag a 1 Fuel Assembly Design Parameters 20 2 Kinetic Characteristics 21 3 Shutdown Requirements and Margins 22 t 4 Control Rod Ejection Accident Parameters 23 L LI5T OF FIGURES Fj er, Title g 1 Core loading Pattern and Source and 24 Burnable Naserber Locations - I 11 1734L:6/441105 i i
G ;[ s 1
~~1.0 INTRODUCTION~~
~~AND~~
~~SUMMARY~~
~~1.1 INTRODUCTION~~
This report presents an evaluation for McGuire Unit 2, Cycle 2, which demonstrates that the core reload will not adversely affect the safety-of the plant. This evaluation was parformed utilizing the methodology described in WCAP-9273, " Westinghouse Reload Safety Evaluation Methodology"(I). McGuire Unit 2 is operating in Cycle I with all Westinghouse 17x17 low parasitic (STO) fuel assemblies. For Cycle 2 (expected startup early 1985) and subsequent cycles, it is planned to refuel the McGuire Unit 2 core with Westinghouse 17x17 optimized fuel assembly (OFA) regions. In the OFA transition licensing submittal (2) to the NRC,l approval was requested for the transition from the STO fuel design to the OFA design and the associated proposed changes to the McGuire Units 1 and 2 Technical Specifications. The licensing submittal justifies the compatibility of the OFA design with the STO design in a transition core as well as a full 0FA core. The OFA transition licensing submittal (2) contains mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 2 safety avaluation. All of the accidents comprising the licensing basesII'3) which could potentially be affected by the fuel reload have been reviewed for the Cycle 2 design described herein. The results of new analyses are included in the above mentioned licensing submittal and in this evaluation, and the justification for the applicability of previous results for the remaining analyses is presenttd. i 1734L:6/841105 1
s \\.. . { ;) a \\ y 4t 1.2, GENERAL DESCRIPTION w The McGuire Unit 2, Cycle 2 reactor core will be comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1. During the Cycle 1/2 refueling, 60 STD fuel assemblies will be re'placed with 60 Region 4 optimized fuel assemblies. A summary of the Cycle' 2 fuel inventory is given in Table 1. I INominal core design parameters utilized for. Cycle 2 are as follows: Core Power (MWt) 3411 System Pressure (psta) 2250 Core Inlet Temperature (*F) 558.5 Thermal Design Flow (gpm) 382,000 Average Linear Power Density (kw/ft) 5.43 (based on 144" active fuel length) !? )
~~1.3 CONCLUSION~~
S From the evaluation presented in this report, it is concluded that the Cycle 2 design does not cause the previously acceptable safety limits to be axceeded. This conclusion is based on the following: .g 1. Cycle 1 burnup is between 14400 and 15400 WD/MTU. 2. Cycle 2 burnup is limited to 10700 WD/MTU including a coastdown. 3. The analyses and proposed Technical Specification changes submitted i;rsupport of the OFA transition licensing submittal (2) are l i/ppr.sved by the NRC pejor to Cycle 2 startup. l \\ 4. inith the changes submitted in support of the OFA transition licensing submittal (2) and the Technical Specification changes given in Appendix A, there is adherence to all plant operating l limitations in the Technical Specification. ' !4 - p 1734L:6/841105 2
o 2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The new Region 4 fuel assemblies are Westinghouse OFAs. The mechanical descriptionandjustificationo{theircompatibilitywiththe Westinghouse STD design in a transition core is presented in the OFA transition licensing submittal.(2) The OFAs and Core Components are designed to be handled by existing handling tools. The control rods, thimble plugs, burnable absorber rods, and source rods are compatible with both the STD and 0FA designs. Table 1 presents a comparison of pertinent design parameters of the various fuel regions. The Region 4 fuel has been designed according to the fuel performance model(4) The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse clad flattening model(5) For all fuel regions, the fuel rod internal pressure design basis, which is discussed and shown acceptable in Reference 6, is satisfied. Westinghouse has had considerable experience with Zircaloy clad fuel. This experience is described in WCAp-8183, " Operational Experience with Westinghouse Cores."( ) Operating experience for Zircaloy grids has also been obtained from six demonstration 17x17 0FAs and four demonstration 14x14 0FAs. This experience is summarized in the OFA transition licensing submittal.(2) 2.2 NUCLEAR DESIGN The Cycle 2 core loading is designed to meet a F (z) x P ECCS limit of 9 l < 2.26 x K(z). I ( 1734L:6/841105 3 l l
q i J 4 Relaxed Axial Offset Control (RAOC) will be employed in Cycle 2 to enhance operational flexibility during non-steady state operation. RAOC makes use of available margin by expanding the allowable AI band, particularly at reduced power. The RAOC methodology and application is fully described in Reference 8. The analysis for Cycle 2 indicates that no change to_the safety parameters is required for RAOC operation. During operation at or near steady state equilibrius conditions, core peaking factors are significantly reduced due to'the limited amount of xenon skewing possible under these operating conditions. The Cycle 2 Technical Specifications recognize this reduction in core peaking i factors through the use of a Base Load Technical Specification. Adherence to the F limit is obtained by using the F Surveillance g g Technical Specification, also described in Reference 8. Fq surveillance replaces the previous F surve m ance h compaH ng a xy measured F, increased to account for expected plant maneuvers, to the g F limit. This provides a more convenient form of assuring plant g operation below the F limit while retaining the intent of using a g measured parameter to verify operation below Technical Specification limits. F surveillance is only a change to the plant's surveillance g requirements and as such has no impact on the results of the Cycle 2 analysis or safety parameters. Table 2 provides a summary of Cycle 2 kinetics character-1stics compared with the OFA transition current limits based on j previously submitted accident analyses. Table 3 provides the control rod worths and requirements at the most limiting condition during theicycle (end-of-life) for the standard burnable absorber design. The required shutdown margin is based on previously submitted accident analysis. The available shutdown margin I exceeds the minimum required. The loading pattern contains 64 burnable absorber (BA) rods located in 16 BA rod assemblies. Location of the BA rods are shown in Figure 1. i 1734L:6/841105 4 --n
~-. o-s 2.3 THERMAL AND HYDRAULIC DESIGN' 7 The thermal hydraulic methodology,.DNBR correlation and core DNB limits used for Cycle 2, are consistent with the OFA transition licensing submittal (2) The thermal hydraulic safety analyses used for Cycle 2 are based on la reduced design flow rate in comparison to Reference 2. No significant variations in thermal margins will result from the Cycle 2 reload. The therral-hydraulic methods used to analyze axial power distributions generated by the RAOC methodology are similar to those used in the Constant Axial Offset Control (CAOC) methodology. Normal operation power distributions are evaluated relative to the assumed limiting normal operation power distribution used in the accident analysis. Limits on allowable operating axial flux imbalance as a function of power level from these considerations were found to be less restrictive than those resulting from LOCA F considerations. g The Condition II analyses were evaluated relative to the axial power distribution assumptions used to generate ONB core limits and resultant Overtemperature Delta-T setpoints (including the f(AI) function). No changes in these limits are required for RAOC operation. l l 1734L:6/841105 5 l
~- [ I. 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION l l ' 3.1 POWER CAPABILITY The plant power. capability has been evaluated considering'the con - sequences of those incidents examined'in the FSAR(3) using the previously accepted design basis. It is concluded' that the core reload a will not adversely affect the ability to safely operate at the design power level (Section 1.0)Jduring' Cycle 2. For the overpower transient, the fuel centerline temperature limit of 4700*F can be accommodated with margin in the. Cycle 2 core. The time dependent densification-model(9) was used for fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining F (z) at or below 2.26 x K(7). q 3,2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the FSAR(3) were examined. In all cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in the previous applicable safety analysis, the safety analysis performed in support of the OFA transition licensing submittal (2), or reanalysis as described in Section 3.3. A core reload can typically affect accident analysis input parameters in the following areas: core kinetic characteristics, control rod worths, and core peaking factors. Cycle 2 parameters in each of these three areas were examined as discussed in the following subsections to i ascertain whether new accident analyses (in addition to the OFA i. analyses) were required. 1734L:6/841105 6
a 3.2.1 KINETICS PARAMETERS-Table 2 is a summary of the OFA transition kinetics parameters current limits along with the associated Cycle 2 calculated values. All of the~ kinetics values fall within the bounds of the OFA current limits. 3.2.2 CONTROL R00 WORTHS Changes in control rod worths may affect differential rod worths, shut-down margin, ejected rod worths, and trip reactivity. Table 2 shows that the maximum differential rod worth of twc RCCA control banks roving together in their highest worth region for Cycle 2 meets the OFA transition current limit. As noted in the OFA transition licensing submittal,(2) Table 3 shows that the Cycle 2 shutdown margin requirement has been changed from 1.6%Ap to 1.3%Ap. The reduced shutdown margin was shown to be acceptable by the results of the OFA transition safety analyses.(2) Table 4 is a summary of 0FA transition current limit control rod ejection analysis parameters and the corresponding Cycle 2 values. The ejected rod worths are within the OFA transition limits. 3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved dropped rod methodology described in Reference 10. Results show that DNB design basis is met for all dropped rod events initiated from full power. The peaking factors for steamline break and control rod ejection have been evaluated and are within the bounds of the limits of the OFA transition licensing submittal (2) analysis. l l l l 1734L:6/841105 7
./ ~ j \\ l 3.3 REDUCED RCS FLOW The safety analyses performed in support of the OFA transition licensing submittal (2) assumed a Thermal Design Flow of 386,000 gpm. For Cycle 2, the TDF. will be 382,000 gpm. This represents an approximate 1 -percent reduction in_the RCS flow used_for the OFA transition licensing submittal (2), 4 The following safety evaluation confirms the acceptability of operation at 100 parcent of rated thermal power and 99 percent of the RCS flow assumed in the OFA transition analyses. All of the affected FSAR-Chapter 15 accidents and protection system setpoints have been reviewed to determine the impact of the proposed reduction in flow requirement. In addition, Technical Specification changes required to support the reduced flow'are included in Section 4.0. 3.3.1 DN8 CONSIDERATIONS i The core DN8 limits have been verified to be unchanged from the OFA transition values, and the conclusion that the DNB basis is met for the following transients remains valid: 4 Excessive Heat Removal Due to Feedwater System Malfunction Excessive Load Increase Main Steamline Depressurization Main Steamline Rupture Loss of Load / Turbine Trip Partial-Loss of Forced Reactor Coolant Flow j Complete Loss of Forced Reactor Coolant Flow L Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition Uncontrolled RCCA Bank Withdrawal at Power Startup of an Inactive Reactor Coolant Loop Inadvertent ECCS Operation at Power _ Reactor Coolant System Depressurization 1734L:6/841105 8
r; 1,. 3.3.2 NON-DNB CONSIDERATIONS In addition to the DNB concern, the following evaluations.are presented for those acciden.s which are not DNB related or for which DNBR is not tha only safety criterion of interest. Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 15.2.1 of the FSAR). The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient. The power excursion caused a heatup of the moderator. However, since the power rise is rapid and is followed by an immediate reactor trip, the moderator temperature rise is small. Thus, nuclear power response is primarily a function of the Doppler temperature coefficient. An increase in temperature due to reduced RCS flow would result in more Doppler feedback, thus reducing the nuclear power excursion as calculated in the OFA transition analysis which partially compensates for the flow reduction. The OFA transition analysis shows that for a reactivity insertion rate -5 of 75 x 10 delta-K/sec, the pe&k hot spot beat flux achieved is 179.4 percent of nominal with a resultant peak fuel average temperature of 2242*F, and a peak clad temperature of 726*F. A 1 percent reduction of reactor coolant flow would degrade heat transfer from the fuel by a maximum of 1 percent. Thus, peak fuel and claa temperatures would also increase'by a maximum 1 percent, yielding maximum fuel and clad temperatures which are still significantly below fuel melt (4800*F) and zirconium-H O reaction (1800*F) limits. Therefore, the conclusions 2 presented in the OFA transition licensing submittal (2) are still valid. i i 1734L:6/841105 9 7 y .. ~ - g -'W
~~" * " " " " " ' ' +~~
^
.o. . o.. Baron Dilution The results of the baron dilution transient will remain unchanged for all modes of operation due to a reduction in reactor coolant flow. The maximum dilution f!ow rate, RCS active volumes, and RCS boron concentrations are not impacted by a reduction in flow. Since these parameters determine the amount of time available to the operator to terminate the dilution event, the results presented in the FSAR remain unchanged. Loss of Load The loss of load accident is presented in Section 15.2.7 of the FSAR and can result from either loss of external electrical load or a turbine trip. The result of a loss of load is an increase in core power which exceeds the secondary system power extraction, thus causing an increase in core water temperature. A reduction in RCS flow will result in a more rapid pressure rise than that calculated in the OFA transition analysis. The effect will be minor..however, since the reactor is tripped on high pressurizer pressure. Thus, the time to trip will be decreased, which will result in a lower total energy input to the coolant. The analysis shows a peak pressurizer pressure of 2567 psia. A 1 percent reduction in flow will lead to a conservative increase in system pressure to less than 2580 psia. The pressurizer will not fill, and the maximum pressures are within the design limits. Therefore, operation at reduced flow will not violate safety limits following a loss of load accident. Loss of Normal Feedwater/ Station Blackout This transient is analyzed to demonstrate that the peak RCS pressure I does not exceed allowable limits and that the core remains covered with water. These criteria are assured by applying the more stringent requirement that the pressurizer must not be filled with water. The effect of reducing initial core flow results in an initial more rapid heatup of the RCS. The resultant coolant density decrease increases the 1734L:6/841105 10
.c volume o'f water in the pressurizer. These transients'have been reanalyzed with the reduced flow assumption. In addition, the low-low-
~~steam generator' level setpoint'will be revised and a filter added to the~~
" channels to help prevent unnecessary reactor trips as a result of load rejections. These changes have been incorporated int'o the reanalysis,- and appropriate Technical Specification changes are identified in Section 4.0. The results show considerable margin ~to filling the ~ pressurizer. Therefore, all safety criteria are met for the events. Steamline Break The steamline break transient is analyzed at hot zero power, end-of-life conditions for the following cases: Inadvertent opening of a steam dump, safety, or relief valve (Section 15.2.13 of the FSAR) Main steam pipe rupture with and without offsite power available (Section 15.4.2 of the FSAR) A steamline break results in a rapid depressurization of the steam. generators and primary side cooldown. This causes a large reactivity insertion due to the presence of a negative moderator temperature coefficient. A reduction in reactor coolant flow will result in a reduction in heat transfer from the fuel to the coolant. : Therefore, the reactivityinsertionandreturn-topowerinthedouble-enifedrupture case for reduced flow conditions would be less limiting that the cases presented in the FSAR. For the double-ended rupture case, the time of safety injection actuation is unaffected by reduced coolant' flow. This, . coupled with a slower return to power would result in a significant reduction in peak average power from the FSAR results. The main steam depressurization case is bounded by the double-ended rupture. Since the return to power is less severe and the DN8 evaluations remain valid as previously stated, the conclusions presented in the OFA transition licensing submittal (2) are still valid for a 1 percent reduction in reactor coolant flow. l 1734L:6/841105 11 I
Rupture of a Main Feedwater Line This transient is analyzed to demonstrate that the peak RCS pressure does not exceed allowable limits and that the core remains covered with water. These criteria are' assured by applying the more stringent requirement that bulk voiding does not occur at the outlet of the core. The effect of reducing initial core flow results in an initial more rapid heatup of the reactor coolant system (RCS). This transient has been reanalyzed with the reduced flow assumption. In addition, the low-low steam' generator level setpoint will-be revised and a filter added to the channels to help prevent unnecessary reactor trips as a result of load rejections. These changes have been incorporated into the reanalysis, and appropriate Technical Specification changes are identified in Section 4.0. The results show considerable margin to hot leg saturation. Therefore, all safety criteria are met for the event. i Locked Rotor Following a locked rotor, reactor coolant system temperature rises until shortly after reactor trip. A reduction in RCS flow will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the transient. The flow reduction in the affected loop is j so rapid that the tjme of reactor trip on low flow does not change due to the 1 percent reduction in reactor coolant flow. Therefore, the nuclear power and heat flux transients will not change from those presented in the FSAR. However, the reduction in flow will result in slightly higher system pressures and clad temperatures. The peak RCS l pressure calculated in the OFA transition analysis was 2593 psia. A 1 percent reduction in reactor coolant flow would cause a conservative increase in pressure to less than 2620 psia, which is still signifi-i cantly below the pressure at which vessel stress limits are exceeded. The peak clad temperature calculated in the OFA transition analysis is 1964*F, well below the limit of 2700*F, and shows that-a slight increase ~ in this parameter due to reduced RCS flow can be easily accommodated. .Therefore, the' conclusions presented in the OFA transition licensing submittal (2) are still valid. l 1734L:6/841105 12
'N .I Control ' Rod Ejection The rod ejection transient is analyzed at full power and hot' standby for both beginning and end-of-life conditions (Sections'15.4.6 of the. FSAR). A reduction in core flow will result in a reduction in heat transfer to the coolant,.which will' increase peak clad and fuel temperatures and peak fuel stored energy. However; all cases have margin to fuel failure limits. The effect of reducing reactor coolant flow is to increase the peak clad temperatures. The analysis shows that,-for the worst case, there is sufficient conservatism in the analysis assumptions and margin in the results such that the peak clad temperature limit (2700*F) is not violate with the reduced flow. This was verified by a reanalysis of the limiting end-of-life zero power case. The peak clad temperature calculated for this case in the OFA transition analysis was 2685'F. The reanalysis of this iise assumed the reduced RCS flow, but used shorter time steps to remove some conservatism in the calculation of the nuclear power transient. The result was a peak : lad temperature of 2683*F. Thus, the limit is not violated. The fuel temperatures and peak fuel stored energy will also increase slightly due to the 1 percent decrease in reactor coolant flow. However, there is sufficient margin between the analysis results and the limits to accommodate the effects of the reduced flow.- Therefore, the conclusions presented in the OFA transition licensing submittal (2) are still valid. ~ LOCA Analysis A LOCA analysis has been performed for McGuire Unit 2 that uses the reduced Thermal Design Flow. Results of the analysis'are given in Section 3.4. i I 1734L:6/841105 13 . ~
Technical Specification Changes The.necessary revisions to the Technical Specifications to support operation at the reduced flow are included in Section 4.0. Each Technical Specification change from the OFA transition submittal (2) g, discussed below. 2.1 Safety Liinits A new reactor core safety limits curve is provided. As discussed above, the DNB limits of the figure are unchanged. However, the Vessel Exit Boiling limits become more restrictive since flow is reduced for a given power. 2.2 Limiting Safety System Settings The protection system setpoints have been reviewed for the reduced flow. The only setpoints which are impacted by the flow reduction are the Overtemperature Delta-T.and Overpower Delta-T functions. These setpoints are designed to protect the core by tripping the reactor before the core safety limits (Figure 2.1-1) are exceeded. The setpoint equations have been recalculated for the reduced flow with 0FA in addition to the introduction of ITDP acd steam generator low-low level setpoint changes. In addition, the time constants in the equations have been updated. Specifically, the lag time constants in the delta-T and Tavg channels have been increased from 2 to 6 seconds, to accommodate operational considerations. The effect of this change has been evaluated by reanalyzing the limiting events that rely on Overtemperature Delta-T and Overpower Delta-T protection. The limiting RCCA Withdrawal at Power cases from the OFA transition analyses have been reanalyzed with the increased time constants in the l Overtemperature Delta-T setpoint equation. The results show that the DNB design basis is met. 1734L:6/841105 14
r-The Overpower Delta-T trip is not relied upon for protection for any of the FSAR accident analyses. However, a spectrum of steamline breaks were analyzed at various power levels in Reference 11 to determine the limiting cases that are presented in the FSAR. Some of the small steamline breaks at power analyzed in this generic study rely on Overpower Delta-T for protection. A McGuire-specific analysis was performed that verifies that the DN8 design basis is met for small breaks at full power with the increased time constants in the Overpower Delta-T setpoint equation. Also, the lead-lag compensation on Tavg is changed from 33/4 to 28/4. The 28/4 compensation was used in the accident analyses and affords the plant more margin to an Overtemperature Delta-T trip on a load rejection. 3/4.2.3 RCS Flow Rate and F-delta-H, and Bases A new RCS flow vs. R figure is provided for Unit 2 to reflect the reduced flow, introduction of 0FA and ITDP, and removal of rod bow parameter, R ' 2 3.4 LOCA Analysis l The large break LOCA analysis applicable for transition and full 0FA core cycles of McGuire 1 and 2 was performed utilizing the OFA design. This is consistent with the methodology given in Reference 2 for the OFA transition. The currently approved UHI Large Break ECCS Evaluation Model modified to incorporate BART(12) core reflood heat transfer models was utilized for the analysis. BART(I3) has been approved for use on non-UHI plants. Four cold leg breaks were reanalyzed. Evaluation of hydraulic mismatches of less than 10% have shown an insignificant effect on blowdown cooling, such that the impact on reflood cooling alone needs to be considered. l l 1 1734L:6/841105 15
ii ~Since the overall resistance of the two types of fuel is essentially I identical, only the crossflows during core reflood due to the smaller -rod size and different grid designs need be evaluated. The maximum flow reduction due to crossflow calculated to occur-in the OFA is ~2.9%. Analyses have been performed which demonstrate that a 5% reduction leads to a maximum PCT increase of 19'F. Therefore, the PCT increase due to crossflow between adjacent OFA and STD assemblies'would be approximately 4 11*F. This effect can be offset in the McGuire 1 and 2 transition cores by considering the favorable UNI quench characteristics of the STD 9 j design. Quenching of fuel throughout the core during blowdown is ~ calculated using UNIPOWERREGIONS LOCTA, with computed parameters then being input to UNIWREFLOOO. If the STD design is modeled the quench j parameters significantly improve, leiding to a faster reflooding of the t core than is true for the OFA case. The magnitude of this benefit is several times the 11* penalty identified for transition cycles; because of this benefit no transition core penalty need be applied. Two further reasons why this method is indeed conservative for transition cores are: -1. The increase in core flow are associated with 0FA due to the y smaller rod diameter has an important impact on flooding rates during reflood. Full 0FA core representation decreases core flooding rates, which reduces heat transfer coefficients. i l i 2. The OFA design has a higher volumetric heat generation rate than STD design. The analysis assumes that the OFA has the hottest rod and maximus F which maximizes the calculated PCT. AH i For breaks up to and including the double-ended severance of a reactor .f coolant pipe, the emergency core cooling system will meet the acceptance j criteria as presented in 10 CFR 50.46..That is: ( 1: The calcu?ated peak fuel element clad temperature is below the require $ehtof2200*. i l t I t l- { 1734L:6/841105 16 ,,_,.,.._,.,.,--__. _ _.,.._,,..._..__.._ _ _-_,. _ ~_ _.
n 1 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of Zircaloy in the reactor. 3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limit of 17% is not exceeded during or after quenching. 4. The core remains amenable to cooling during and after the break. 5. The core temperature is reduced and decay heat is removed for an extended period of time as required by the long-lived radioactivity remaining in the core. The Large Break LOCA analysis for McGuire 1 and 2 utilizing the currently approved UHI Evaluation Models modified to incorporate BART technology resulted in a PCT of 2157'F at 2.26 F for the CD = 0.6 g (perfect mixing) DECLG break. The small impact for transition core cycles is offset by the presence of STD fuel in the core so that margin to 10 CFR 50.46 limits remains in transition cycles. i i i 1734L:6/841105 17
4.0 TECHNICAL SPECIFICATION CHANGES To ensure that plant operation is consistent with the design and safety evaluation conclusion statements made in this report and to ensure that .these conclusions remain valid,'several technical specifications changes will be needed for Cycle 2. These changes are summarized below. (1) Technical Specification changes outlined in the OFA transition licensing submittal.(2) (2) Technical Specification changes given in Appendix A. 4 e N i [ s i i 1734L:6/841105 18
} -b
~~5.0 REFERENCES~~
~~1. Bordelon,.F.M., et. al., " Westinghouse Reload Safety Evaluation~~
~~' Methodology", WCAP-9273, March 1978.~~
2. Duke Power Company Transmittal to NRC, " Safety Evaluation for McGuire Units 1 and 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies."

3. "McGuire Final Safety Analysis Report." ~

4. Miller, J.V., (Ed.), " Improved Analytical Model used in Westing-house Fuel Rod Design Computations", WCAP-8785, Dctober 1976.

5. George, R.A., (et. al.), " Revised Clad Flattening Model", WCAP-8381, July 1974.

6. Risher, D. H., (et. al.), " Safety Analysis for the Revised Fuel Rod Internal Pressure Design 8 asis," WCAP-8964, June 1977.

7. Skaritka, J., Iorii, J.A., " Operational Experience with Westinghouse Cores", WCAP-8183, Revision 13, September, 1984.

8. Miller, R. W., (et al.), " Relaxation of Constant Axial Offset Control-F Surveillance Technical Specification," WCAP-10217-A, g
~~June 1983.~~
9. Hellman, J.M. (Ed.), " Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.

10. Letter from NRC, C. O. Thomas to E. P. Rahe, Jr., Westinghouse,
~~" Acceptance for Referencing of Licensing Topical Report WCAP-10297-(P), WCAP-10298 (NS-EPR-2545) Entitled Dropped Rod Methodology for Negative Flux Rate Trip Plants", March 31, 1983.~~
11. Hollingsworth, S. D. and Wood, D. C., " Reactor Core Response to Excessive Secondary Steam Releases," WCAP-9226, Revision 1, (Proprietary), January,1978, and WCAP-9227. Ravlsion 1, (Non-Proprietary), January,1978.

12. Schwartz, W. R., " Addendum to BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients," WCAP-9561, Addendum 1, November 1984. (Westinghouse Proprietary) i

13. Young, M. Y., et al., "BART-A1: A Computer Code for the 8est Estimate Analysis of Reflood Transients," WCAP-9561-P-A, March 1984. (Westinghouse Proprietary) t I
1734L:6/841105 19
TABLE 1 MCGUIRE UNIT 2 - CYCLE 2 FUEL ASSEMBLY DESIGN PARAMETERS Region 1 2 3 4* Enrichment (w/o U-235)* 2.093 2.566 3.089 3.20 Density (% Theoretical)* 94.77 94.41 94.87 95.0 Number of Assemblies 5 64 64 60 Approximate Burnup at++ 14698 16750 11640 0 Beginnin.g of Cycle 2 (MWD /MTU) Approximate Burnup at++ 23793 26635 23250 10692 End of Cycle 2 (MWD /MTU)
~~Optimized Fuel - Zire grid~~
+ All fuel region values are as-built except Region 4 values which are nominal. ++ Based on EOC1 = 14900 MWO/MTU, EOC2 = 10700 MWO/MTU (coastdown included) 1734L:6/841105 20 ~ -.- -
TABLE 2 MCGUIRE UNIT 2 - CYCLE 2 KINETICS CHARACTERISTICS .OFA Transition Cycle 2 Current Limits (2) Desian L Minimum Moderator +5 < 70% of RTP +5 <70% of RTP Temperature Coefficient 0 1 70% of RTP 0 170% of RTP (pcm/ F)* Doppler Temperature -2.9 to -0.91 -2.9 to -0.91 Coefficient (pcm/ F)* Least Negative Doppler- -9.55 to -6.05 -9.55 to -6.05 Only Power Coefficient, Zero to Full Power, (pcm/% power)* Most Negative Doppler -19.4 to -12.6 -19.4 to -12.6 Only Power Coefficient, ~ Zero to Full Power (pcm/% power)* Minimum Delayed Neutron .44 >.44 Fraction S,ff, (%) Minimum Delayed Neutron .50 >.50 Fraction 6,ff, (%) [ Ejected Rod at BOL] Maximum Differential Rod 100 <100 Worth of Two Banks Moving Together (pcm/in)* -5
~~pem = 10 Ap i~~
p 1 l l 1734L:6/841105 21 l
TABLE 3 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MCGUIRE UNIT 2 - CYCLE 2 Control Rod Worth (%Ap) Cycle 1 Cycle 2 All Rods Inserted 7.86 7.11 All Rods Inserted Less Worst Stuck Rod 6.73 6.07 (1) Less 10% 6.06 5.46 Control Rod Requirements Reactivity Defects (Doppler, T,yg, 3.1 2.98 Void, Redistribution) Rod Insertion Allowance 0.50 0.50 (2) Total Requirements 3.60 3.48 Shutdown Margin [(1) - (2)] (%Ap) 2.46 1.98 Required Shutdown Margin (%Ap) 1.'60 1.30 l 1734L:6/841105 22 l
.c. TABLE 4 MCGUIRE UNIT 2 - CYCLE 2 CONTROL ROD EJECTION ACCIDENT PARAMETERS OFA Transition liZP-BOC - Current Limit Cycle 2 Maximum ejected rod 0.75 <0.75 worth, u p Maximus Fg (ejected) 11.5 <11.5 HFP-BOC Maximum ejected rod 0.23 <0.23 worth, %Ap Maximum Fg(ejected) 5.3 <5.3 HZP-EOC Maximum ejected rod 0.90 <0.90 worth, %Ap Maximum Fg (ejected) 20.0 <20.0 HFP-EOC Maximum ejected rod 0.23 <0.23 worth, u p Maximum Fg (ejected) 5.9 <5.9 t I r i i 1734L:6/841105 23 l
4 180* 'R P "N M L K J H G F E D 'C B A i 4 4 4 4 4 4 4 i 1 l l e 4 4 4 2 4 1 4 2 4 4 4 4 4 4 4 2 l 4 3 3 2 2 2 2 2 2 2 3 3 4 SS _3 4 3 3 3 3 2 2 2 3 _3 3 3 4 4 l. 4 4 2 3 3 2 3 2 3 2 3 3 2 4 ' 4 4 5 4 4 2 2 3 2~ 3 3 3 3 3 2 3 2 2 4 1-6 4 4 2 2 3 3 3 2 3 3 3 2 2, 4 ~4 4 7 4 90* 4 1 2 2 2 3 2 1 2 3 2 2 2 1 4 a 270* 4 4 2 2 3 3 3 2 3 3 3 2 2 4 4 4 1 4 8 4 2 2 3 2 3 3 3 3 3 2 3 2 2 4 ja 4 4 2 3 3 2 3 2 3 2 3 3 2l 4 4 4 l 4 11 4 3l3 3 3 2 2 2 3 3 3 3 4 1 I -12 4 3 3 2 2 2 2 2 2 2 3 3 4 SS 13 4 4 4 2 4 1 4 2 4 4 4 4 4 4 4 I4 4 4 4 4 4 4 4 15 0* l l X region number Y GA'S SS Secondary Source Figure L CORE LOADING PATTERM i MCGUIRE (MIT 2. CYCLE 2 24
APPENDIX A TECHNICAL SPECIFICATION PAGE CHANGES ,(In addition to proposed changes submitted in support of the OFA transition licensing submittal (2)) Delete Pages 3/4 2-2 3/4 2-3 3/4 2-5 (Figure 3.2-18) 3/4 2-10 3/4 2-11 (Reference to Amendment 32 (Unit 1), Amendment 13 (Unit 2)) i l 1734L:6/841105
4 e } O I l l e 4 e MODIFICATIONS TO 3/4.2.1 ' AXIAL F1.UX DIFFERENCE LIMITS 9 O l l l
e ~.: i 5 /4. 2 POWER DI5it*SUTICN LIMITS AFD 3/4.2.1 AXIAL FLUX DIFFERENCE _MNH-M LIMITING CON 0! TION FOR OPERATION i 3.2.1 The indicated AXIAL FLUX O!FFERENCE (AFD) shall be maintained within: a. the allowed operational space defined by Figure 3.2-1 for RACC operation, er 5 b. within a 2 2 percent target band about the target flux difference during base load operation. j APPLICA8ILITY: M00E 1 above 50K of RATED THERMAL POWER *. ] ACTION: t a. For RA0C operation with the indicated AFD outside of the Figure 3.2-1
~~limits, e~~
1. Either restore the fndicated AFD to'within the Figure 3.2-1 { limits within 15 minutes, or 2. Reduce THE M L POWER to less then 50K of RATED THEIMAL POWER within 30 minutes and reduce the Power Range Neutron Flum - High Trip setpoints to less than or equal to 55X of RATED THEIMAL POWER witnin the next 4 hours, b. For Base Lead operation above APL with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target flux difference: l 1. Either restore the indicated AFD to within the target band j limits within 15 minutes, or E 2. Reduce THEIMAL 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 50K of RATED THERMAL 1 POWER taloss the indicated AFD is within the Figure 3.2-1 limits. i l 1 asee special Test Exception 3.10.2. I
~~APL" is the minimum a11susble power level for base lead operation and will' be provided in the Peaking Facter Limit Report per Specification 6.9.1.9.~~
w w a .e
~ r 1 i l $"tvEILLANCE RECidt!MENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 505 of RATED THERMAL POWER by: a. Monitoring the indicated AFD for each OPERA 8LE excore channel: 1. At least once per 7 day's when,the AFD Monitor Alare is OPERABLE, and 2. At least once per hour for the first 24 hours after restoring the AFD Monitoring Alare to OPERA 8LE status. b. Monitor.ing .J logging the indicated AFD for each OPERABLE encore channel at least once per hour for the first 24 hours and at least once per 30 af.wtss thereafter, when the AFD Monitor Alare is inoperable. The logged values of the indicated AFD shall be assured j to exist during the interval preceding each logging. 4.2.1.2 The indicated AFD shall be considered outside of its limits when at least two OPERA 8LE encore channels are indicating the AFD to be outside the 1 limits. i 4.2.1.3 'idhen in Base Lead operation, the target axial flux difference of each OPERABLE excore channel shall be dotaruined by measurement at least once l per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 idhen in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining l the target flux difference p. .__.^. ;. ^.;.1.; above or t;y lincar interpola-i tion between the most recently% esured value and 0 e -_.-.; at the end of j cycle life. The provisions of Specification 4.0.4 arednot applicable. % calcaldd al2 JQ l 'i. c ;uk .at, L u niH e ray;<a,,J,,f spescak 6xP 5 i f l k*13 .=- - w, . e a esww E a 6 e,
~~.8 9~~
f-t l l I ) 100 6 100 ON TION l ON 1 i l t 1 60 3 1 50 l 1 50 (17,50) 40 i I 20 i O -50 -40 -30 -20 -10 0-10 20' 30 40 50 Flux Difference (AI)5 FIGuitE 3.2-1 AXIAL FLUX OIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER i l l l i l l
) e' l) . 3. ~ 'l e O s a MODIFICATIONS TO 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR. LIMITS e 9 9 0 0 m 9 _.7
~ POWER DISTRIBUTION LIMITS HEAT ILUX H0T CHANNEL FACTOR-FQ(Z) LIMITING CONDITION FOR OPERATION 3.2.2 F (z) shall be limited by the following relationships: q a.as F (z) 1 [ 4rM.1 (K(Z)] for'P > 0.5 ~ q r 2.26 ~ F (z) 1 [ M 1 [K(I)] for P 1.0.5 O o.5 where P = THDMAL POWER and K(z) is 'the function obt'ained from Figure 3.2-2 for a-given core height location. 1 APPLICA8ILITY: MODE 1 ACTION: With F (z) exceeding its limit: q 1. Reduce THERMAL POWER at least 1 percent for each 1 percent F (z). exceeds. the limit within 15 minutes and similarly 0 reduce the Pcwer Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided*the Overpower AT Trip Setpoints (value of K ) have been reduced at least 1 percent (in AT 4 span) for each I percent Fq(z) exceeds the lia,it. b. Identify and correct the cause of the out of limit condi-tion prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided Fn(z) is demonstrated through incore mapping to be within its limit. l e D 4
i.. 1 i SURVEILLANCE REQUIREMENTS JO::7 '; 4.2.2.1 The provisions of Specification 4.0.4 are not applica>,le, 4.2.2.2 For RA0C operation, F (z) shall he evaluated to determine if F (z) is within its limit by: q q a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than SE of RATED THERMAL POWER. b. Increasing the measured F (z) component of the power distribution q map by 35 to account for manufacturing tolerances and further increasin
~~ertainties~~
~~%rify %g the value by 55 to account for sensurementEisfid. rea;<<, e ts a staificatie 3.a.a ee s c. Satisfying the following relationship: N~~
~~"IZI Fq (z) i yg for P > 0.5 N~~
Fq (2) 1 for P i 0.5 ,o where (z) is the sensured F (z) increased by the allowances for q manufacturing tolerances and measurement uncertainty, N is the F g _. limit, K(2) is given in Figure 3.2-2, P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during noneal 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 10K or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined," or q 2. At least once per 31 Effective Full Powe[ Days, whichever i occurs first. t r
~~During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved j~~
and a power distribution sep obtained. m== =:n : = : m 2-7 e -r ,-m_<,--_,-~~~,3 -w -.v. .-,c.. m-- -,---4--e ,<,,.,--m,,e-
SU WEILL N E REC..sEMEC I J.:' 3 (Continued) e. With measurements' indicating maximus FM (z) over 2 ( z N has increased since the previous detemination.of Fq (z) either of ~ the following actions shall be taken: M 1) Fq (z) shall be increased by 25 over that specified in Specifi-cation 4.2.2.2c. or 2) F "(z) shall be measured at least once per 7 Effective Full q Power Days until two successive maps indicate that 4 maximum fFM (2) is not increasing. ( z/ over z f. Witin 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: q fI )4 x 100 ~ fmaximum M O (*)
~~NZ) xK(z)j-1 for P > 0.5~~
~~' wer z s }ffmaximum 0 (*)~~
~~NZ)T M~~
~~-1 x 100 for P < 0.5~~
~~l. L "*"~~
~~2.26j, Wg x K(z)j s 2)~~
One of the following actions shall ta taken: a) Within15 minutes,controltheAFD'EwithinnewAFD limits which are determined by reducing the AFD limits of 3.2-1 by 1% AFD for each percent F (z) exceeds its limits q i as determined in Specification 4.2.2.2f.1). Within 8 hours, reset the AFD alarm'setpoints to these modified limits, or b) Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated q above, or, c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation. ":C'J:"5 ""!T! ! rd 2 3/4 2-8 WNo3 "* M l + .,.,m ,w. ,,y.7.,-,.,.r._-_,_.., ,__,_,m,. ,,_~...,_,-,_...,-,-,--e--___-.--.-
7 ~ SUt.ERLANCE REC. *?.EMElCi. ":T " (continued) l s 'b ~ 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 155, inclusive. 2. Upper core region from 85 to 2005, inclusive. E 4.2.2.3 Base Lead operation is po mitt 2d at powers above APL if the following conditions are satisfied: a. Prior to entering Base Load operation, maintain THERMAL POWER above E 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 s'urveillance (AFD within 2 % of target flux difference) during this time perioo. Base Load operation is then permitted providing THERMAL i POWER is maintained between APL and APL er between APL and E l 1005 (whichever is most limiting) and F surveillance is maintained q E pursuant to Speciff a ion 4.2.2.4. APL is defined as: 3pgBL, minimum g x KfZ) ) x 1005 over I (Z) x W(Z)gt ~ here: (z) is.the measured F (z) increased by the allowances for w q f ring tolerances and measurement uncertainty. The F limit q is K(z) is given in Figure 3.2-2. W(z)BL I' *"" "YCI' dependent function that accounts for limited power distribution transients encountated during base load operation. The function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9. 4 b. During Base Load operation, if the THERMAL POWER is decreased below 1 E APL then the conditions of 4.2.2.3.a shall he satisfied before re-entering Base Load operation. 4.2.2.4 During. Base Load Operation F (Z) shall be evaluated to determine if q F (Z) is within its limit by: q Using the novable incore detectors to obta'in a power distribution a. map at any THERMAL POWr.R above APL ,i l b. Increasing the measured F (Z) component of the power distribution q asp by 35 to account for manufacturing tolerances and further increasing the value by 55 to account for measurement uncertainties. va;G +In nyin e,Js of 9eakh 3. 2. 2 ue.s.ksfist. NOC un T; 1.. 2 3/4 2-9 l s l
m. S'JR'!EILLA';CE REC '?.EFE* ~5 ""'F ) (Continued) Satisfying the following relationship: c. -F V F (Z) i p.i Z) f,7 p 3 gpg2 F"g(Z) is the measured F (2). 2.26 where: q The F limit is-art 5. q K(Z) is given in Figure 3.2-2. P is the relative THOMAL POWER. - ~ 9f(Z)gg is the cycle dependent function that accounts for Itaited power distribution transients encountered during nomal operation. This function is given in the Peaking Factor Limit Report as per ~ Specification 6.9.1.9. d. Measuring (1) in conjunction with target-flux difference deter-i aination according to the following schedule: 1. Prior to entering BASE LDAD operation after satisfying Sectiotr 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 E maintained above APL for the 24' hours prior to mapping, and 2. At least once per 31 effective full power days. ifith measurements indicating e. (2) maximum [ ] over Z has increased since the previous determination following actions shall be taken: (Z) either of the 1. (Z) shall be increased by 2 percent over that specified in 4.2.2.4.c, or / 2. (Z) shall be measured at least once per 7 EFPD until 2 i successive maps indicate that maximum [ ] is not increasing. over 1 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.cissatisfied,andremeasure((Z),or Wip 3i,. 2-. 4 O e 9 ,n,_.m_,, ,c-.,- -..,m,- .w w,,. ..,,_-,,--,_,.,,-,-,-ym.,, ,-n,-..,---ng,--.-,,.,-,,...,,.,,._,an., n,- .-w .-e-,
7_ 5t'Ev!Li'CE RECT.1REMEr2 5 f t'"I' M (Continued) 2. Comply with the requirements of Specification 3.2.2 for'F (Z) g exceeding its limit by the percent calculated with one of the following expressions: Fh(Z)xW(Z)BL ND ] ) -1 ] x 100 for P > APL ' ((max, over z of [ Y x K(Z) ~~ a.26 ~ x W(Z) g [(max. over z of [ ] x 100 for 0.5 i P < APL x K(Z) g. T,he 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, inc1,usive. 2. Upper core region 85 to 100 percent, inclusive. 4.2.2.5 When F (Z) is seasured 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 sensurement uncertainty. 3/4 2-9b l .m._ .m._, --.m.. - -...,, _ _ _.
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l e t ~ ?/a.2 F7gt 0:5T* ELTIO'. L!"ITS. BASES i The specifications of this section provide assurance of fuel integrity i - during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) ~ events by: (1) maintaining the calculated DNOR 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: i F (Z) Heat Flux Hot Channel Factor, is defined as the esximum local l 0 heat flux on the surface of a fuel rod at core elevation Z divided l by the average fuel rod heat flux, allowing for manufacturing toler-ances on fuel pellets and rods; F" Nuclear Enthalpy Rise Not Channel Factor, is defined as the ratio of g the integral of linear power along the rod with the highest integrated power to the average rod power; and a l 3/4.2.1 AXIAL FLUX OIFFERENCE ~ L24 ThelimitsonAXIAi. FLUX IFFERENCE (AFD) assure that the FQII) "PE*" bound envelope of 0. " 'S.P. 0,, frt5 '...M U times the normalized axial } peaking factor is not exceeded during either normal operation or in the event of menon redistribution following power changes. l Target flux difference is determined at equilibrium ~ xenon conditions. The full-length rods may be positioned within the core in accordance with l l 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 1evels are obtained by multiplying the RATED THERMAL POWER value t by the ap;,ropriate fractional THERMAL POWER 1evel. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations. e G w n m e .e-s-, - --m---<-c-p-(m,- --->,.-,-e=,w------x------e,w,-ww,,,,m,e-,,ar__ -sm.emm.,,--.n,-n._-e,---..n. m,,,e,mmn-,
FO.!ER DISTRIBUT:t'. LIMITS. BASES 1 AXIAL FLUX OIFFERENCE (Continued) Although it is intended that the plant will be operated with the AFD wi the target band required by Specification 3.2.1 about the target f1 diffe e, during rapid plant THERMAL POWER reductions, control rod sio n will cau the AFD to deviate outside of the target band at reduced RMAL POWER level. This deviation will not affect the xenon redistri on suffi-ciently to c the envelope of peaking factors which may ached on a subsequent return RATED THERMAL POWER (with the AFD wit the target band) provided the time du ion of the deviation is limited. cordingly, a 1 hour penalty deviation limit lative during the previou 4 hours is provided for operation outside of the t t band but within th imits of Figure 3.2-1 i while at THERMAL POWER levels tween 50% and 9 of RATED THERM L P;JER. 2 For THERMAL POWER levels between and 505 RATED THERMAL POWER, devia-4 ] tions of the AFD outside of the tar be are less significant. The penalty of 2 hours actual time reflects this ed significa,nce. Provisions for monitoring AFD on an ' tic basis are derived from j the plant process computer t the AFD Mont Alarm. The computer deter-eines the 1 minute avera f each of the OPERABLE ore detector outputs and provides an alare mess immediately if the AFD fort or more OPERA 8LE excore channels a tside the target band and the THE OlER is greater than 901-of RA HERMAL POWER. During operation at THERMA. R levels between 505 905 and between 155 and 50K RATED THERMAL POWER, computer outputs a are message when the penalty deviation accumulates bey the limits 1 hour and 2 hours, respectively. i i Figure 8 3/4 2-1 shows a typical monthly target band. T., L.:^.1, $ power levels below APL, the limits on AFD are defthed by E Figures 3.2-1, i.e. that defined by the RAOC operating procedure and limits. l These Ifmits were calculated in a manner such that expected operational transients, e.g. load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the - ~, short period of time allowed outside of the limits at reduced power levels will not result in significant xenon redistribution such that the envelope of j~ peaking factors would change sufficiently to prevent operation in the vicinity of the APL power level. At power levels greater than A.%, two modes of operation are permissible; j d
~~1) RAOC, the AFD limit of which are defined by Figure 3.2-1, and 2) ase Load operation, which is defined as'the maintenance of the AFD within a band f~~
N about a target value. The RAOC operating procedure above APL is same as E that defined for operation below APLE. However, it is possible when following extended load following maneuvers that the AFD limits may result in I restrictions in the maximum allowed power or AFD in order to guarantee operation with F (z) less than its ilmiting value. To allow operation at the q maximur. permissible value, the Base Lead operating procedure restricts the N J:": ...; ; ;., ~ B 3/4 2-2 h o W-I ~ ...,-,,,_,,-.-,--,,,,m_,.-_m m.,..,-- ,--4m,w,,,,v -.ww-,w-,e,--m,-.,--.---,,--,.--.--,.,.---..w. __,__2.
o j i r;.gc OISTR!!'.'TIU: lit'I' 5 i BASES i AXIAL FLUX DIFFERENCE (Continued) indicatedfDtorelativelysmalltargetbandandpowerswings(AFDtarget bandofitR,APLNO,,,,,,,gpgBL or 1005 Rated Thermal Power, whichever is lower). For Base Load operation, it is expected that the plant will operate within the target band. Operation outside of the target band for the short i time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would 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 NO operation, a 24 hour waiting period at a power level above APL and allowed ty RAOC is necessary. During this time period load changes and red motion are restricted to that allowed by the Base Loac procedure. After the waiting i j period extended Base Load operation is permissible. a T R. Ur.M 1, the computer determines the one minute average of each of the d OPERABLE encore detector outputs and provides an alare 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 RA0C operatian), or 2) outside the allowed AI target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RA0C operation), or 2) APLND (for Base Load operation). Penalty deviation minutes for lase Load operation are not accumulated based on the short period of time during which i operation outside of the target band is allowed. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux het channel factor, RCS flow rate, and nuclear enthalpy rise hot channe.1 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 accep-tance criteria limit. ~ i Each of these is measurable but will normally only be determined periodica113 as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided: 1 ) SUZ ..d !" 1 .m. 2 33/4 2-2a T.% a -* h.24(UnH 9 - g "--.^. Wo." (uni 2 6 --r-._, -m.. -...--.w. ,--.---.-,e e,, ,.-----,..,,v,--------.-,---v.
o i L t ! C:57R*E 71*.* LI"I?:- BASES i HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE HOT CHANNEL FACTOR (Continued) The hot channel factor F (z) is measured periodically and increased b cycle and height dependent power factor appropriate to either RAOC or Base Load operation, W(z) or W(z)8L, to provide assurance that the Ifmit on the hot channel factor, F (z), is set. W(z) accounts for the effects of normal q operation transients and was determined from expected over the full range of burnup conditions in the core. power control maneuvers W(z)BL accounts for the more re.strictive operating limits allowed by Base Lead operation which result in less severe transient values. is provided in the Peaking Factor Limit Report per Specification 6.9.1.9T 3/4.2.4 OUADRANT POWER TILT RATIO bution satisfies the design values used in the power c 1 Radial power distribution measurements are made during STARTUP test l periodically during power operation. l The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned rod. rect the tilt, the margin for uncertainty on fin the event such action does not cor-i q s reinstated by reducing the power by 3% from RATED THERMAL POWER for each percent of tilt in of 1.0. For purposes of monitoring QUADRANT POWER TILT RATIO when one ex the normalized symmetric pcwer distribution is consist POWER TILT RATIO. The incere detector monitorin l flux map or two sets of four symmetric thimbles.g is done with a full incore thimbles is a unique set of eight detector locations.The two sets of four symmetric l C-8, E-5, E-ll, H-3, H-13, L-5, L-11, N-8. These locations are 3/4.2.5 DNS PARAMETERS, The limits on the DNB-related parameters assure that each of the para-meters are maintained within the normal steady-state envelope of operaticn assumed in the transient and accident analyses. with the initial FSAR assumptions and have been analytically demonstratedTh adequate to maintain a design limit DNBR throughout each analyzed transient. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. 33/4 2-6 n __m_.-.my . _ - _ _. - -. - _. _ - - -.. - _. ~ - -
o THIS FIGURE DELETED ' e d Mg.are S 3/4 21 TYP1 cal. INDICATED AXtAL Flux ctFFERENCI VER3US TNEMMAl. FCWER 4
ADDITIONS TO 6.0 AIMINISTRATIVE CONTROL.S 9 s
17"!v!$'uT!'/t C Tr,5 RADIAL PEAKING FACTOR LIMIT REPORT ' 6. 9.1. 9 The F,y limit for RATED THERMAL POWER (F,RTP) shall be provide ional Administrator of the NRC Regional Office, with a to the Directo, lear Reacter Regulation, Attention: Chief Performance Branch, U. 5. Regulatory. Commission, Washi , D.C. 20555 for all 'n core planes containi "0" control rods a unrodded core planes at i l least 60 days prior to cycle I cr ity. In the event that the limit g would be submitted at some other core life, it shall be submitted g l 60 days prior to the date t it would ffective unless otherwise exempted by the Commi .Any info needed to support F,RTP will be by request from t nd y n et be includec in tr.is report. 'The W(2) functions for RAOC and Base Load operation and the value for APLND (as required) shall be provided to the Ofrector, Nuclear Reactor Regulations. Attention: Chief Core Performance Branch', U.S. Nuclear Regulatory Commission, Washington, D.C, 20555 at least 60 days prior to cycle initial criticality. In the event that these values would be submitted at some other time during g[. core life, it will be submitted 60 days prior to the date the values would become effective unless otherwise exempted by the Commission. i i Any information needed to support W(z), W(z)gg and M wH1 h by m uest ,from the NRC and need not be included in this report. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report. 4 t i t i 1 6 i I m,_.., 6 21 g -,---e-w-*,,-,-r.,,--,.,,w,m w -bw- - -,,-,,m,,.--~,-w, ,,g-m.u-,rm
FURTHER MODIFICATIONS DUE TO : Reduced RCS Flow Shutdown Margin Positive MTC New Rod Bow Methodology Increased Tin OTAT and OPAT Equations Improved Thermal Design Procedure Revised SG Low-Low Level Setpoint l t .-.r. ,,_.-,c. se.. - --w-yv. -e-., y-, -w., .,.-..v.. wr -n w- - - - - - - - '- --' -v-v-1 r- -e--v--S-nv'-= y--
I 642 I Sea 6 Flow Per Loon = .500 gun 1 655 i I Eton D 6536 tte 645 225 Unac:sstable
~~Oogf, Coeration 648 ggg.,~~
~~$38 ffe EIS-4.r 800 affe~~
~~823,~~
,sts - sis i SSE EC3 \\ 555 Ac:sotae g. Oo. ret- \\ E85 < 588<- 575< 9 \\1. .1 .2 .5 4 .s .7 J l.I 1.2 80MCR Ifr est. ion of nominee I \\ kaaee. ak t' \\ i 'O lOWIMC FIGURE 2.1-lb a i UNIT 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION COJ:F.: y!:'TC 1.... E 2-2a ? n.. "- : - t i:..;- tuin 6 4 .t..._.. no. u '.2* 2b i --m-r--, e-e e,~,., .-m--e _ - -, -,, - - - - - - - - ~ -,
? s48 Fl w b Laep 972IO ym a sss h/, Vas Isle Operabe s48 - 4o 855 s58 - ks25 k. n $W- \\ ~ $15 - /, j sis - ses. AccepMle. -Operaboa 5 45 sw- \\ ses S. I .2 5 4 .5 .8 .7 .8 9 1. 11 I3 l Powat Ireaction of nomineiI l l EsEAC, wit. CORE SAFETY LIMITS Fout Loops ig oPERATioM 4 e 4 -,,-y m,.,--.r-~ ,-v,-

~~.--.---m-----r--~~

--r-- - - - ~. - - - - - - ~ ~ - - - - - - - - = - -
=. _ _. i 3 I 4 i J TABLE 2.2-1 ) REACTOR TRIP SYSTEM INSTRt3ENTATION TRIP SETPOINTS FUNCTIONAL UNIT-TRIP SETPOINT ALLOW 4BLE VALUES
~~1. hanual Reactor Trip N.A.~~
~~' N.A. 1j~~~
~~2. Power Range, Neutron Flux Low Setpoint 1 255 of RATED Low Setpoint - 1 2 W of RATED i~~
~~THERMAL POWER THEMAL POWER High Setpoint - 1 109K of RATED High Setpoint - 1 1105 of RATED THERMAL POWER THE MAL POWER i i~~
~~3. Power Range,skutron Flux,~~
~~< SX of RATED THEMAL POWER with < 5.55 of RATED TIEMAL POWER l High Positive Rate a '.ime constant > 2 seconds with a time constant > 2 seconds l.~~
~~4. Power Range, Neutron Flux, 1 5% of RATED THERMAL POWER with 1 5.5% of RATED THElst4L POWER j~~
~~y High Negative Rate a time constant > 2 seconds with a time constant > 2 seconds j Ut~~
~~5. Intermediate Range, Neutron i 255 of RATED THERMAL POWER 1 30% of RATED THERMAL POWER Flux 5~~
5
6. Scurce Range, Neutron Flux 1 10 counts per second i 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 1

9. Pressurizer Pressure--!aw
~~) 1945 psig > 1935 psig l~~
~~10. Pressurizer Pressure--High 1 2385 psig i 2395 psig l~~
g
11. Pressurizer Water Level--High 1 92% of instrument span i 935 of instru met span I

12. Low Reactor Coolant Flow
~~> 905 of design flow per loop * > 895 of design flow per loop" t g~~
~~De ign flow is w -" M 77 % f:r-* tit 1 "Mgym per loop for Unit 2.~~
97,220 + v
~ t TA8LE 2.2-1 (Continued) l REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8LE VALUES E ]
~~13. Steam Generator Water~~
> 12% of span fron'0 to 30% of 3 11% of span from 0 to.30% 6f-Level--Low-Low RATED THERMAL POWER, increasing RATED THERMAL POWER, increasing linearly to >-64AE f span at to 49-415 of span at 100E of RAIED Q 100% of RATED THERMAL POWER.. THERMAt$ POWER. 40.0 % 39.0 % -)
14. Undervoltage-Reactor 1 5082 volts-each bus 1 5016 volts-each bus Coolant Pumps

15. Underfrequency-Reactor 1 56.4 Hz - each bus 1 55.9 Hz - each bus q
~~Coolant Pumps i~~
~~16. Turbine Trip~~
~~{ a. Low Trip System Pressure > 45 psig > 42 psig' b. Turbine Stop Valve Closure 1 1% open 1 1% open~~
~~17. Safety Injection Input N.A.~~
~~N.A. from ESF~~
~~18.. Reactor Trip System Interlocks~~
-10 ~11 a. Intermediate Range Neutron Flux, P-6, 1 1 x 10 amps 6 x 10 gs Enable Block Source Range Reactor Trip b. Low Power Reactor Trips Block, P-7 I 1) P-10 Input 10% of RATED > 95, < 11% of RATED THERMAL POWER . THERMAL POWER 1 2) P-13 Input < 10E RTP Turbine < 11% RTP Turbine Impulse Pressure Tapulse Pressure 1 Equivalent Equivalent 1
i 4 TABLE 2.2-1 (Continued)
~~c i~~
l _* 'l REACTOR TRIP SYSTEN INSTRtBENTATION TRIP SETPOINTS t } i ! I NOTATION l NOTE 1: OVERTEMPERATURE AT I L i j 1+tS 1 AT (I * ) (1 + r s) < ATo (K1 4 2 (1 + tsS)[T(1 + t.S)-T'] + K (P-P') - f (AI)) 1 + Ta5 -K 2 a 3 1 t Wherg: AT = i Measured AT by RTD Manifold Instrumentation, 1+rS g,,'$ Lead-lag compensator on measured AT. = J i 13, 1 = Time constants uttilzed in the lead-lag controller for i u, AT, is % 8 sec., t: 4 3 sec., i a 3 3, Lag compensator on measured AT', = b Time constants utlitzed in the lag compensator for AT. Ya 4/sec., Ta = 4. i AT* Indicated AT at RATED THERNAL POWER, = K 1.20'd 'i 't ), 1.t^'? (i !i 1), 1.200 3 j r K = 2 0.0 0; (:;.,;. 2), 0.0222 ' i'i 13, / l ~ F f 1 E 1 + 1.5 The function generated by the lead-lag controller for T,q dynam,1c compensation, = g,,,$ I !,I
~~g Time constants uttilzed in the lead-lag controller for T~~
= t. { t3 ; q -- :^. (L. ; ;). t. %.3F sec. 't't 2). ts 4 4sec.f8, 25 .T Average temperature 'F, = 1 a ne ~ " 1 Lag compensator on measu M T,,,, = g,t$ 1
I c 4 a TA8tE 2.2-1 (Continued) { i 2 REACTOR TRIP SYSTEM INSTatsIFNTATION TRIP SETPOINTS i r NOTATION (Continued) l NOTE 1: (Continued) i
~~i b.~~
~~Time constant utfilted in the esasured T,,8. leg coepensator, to 4/sec v. = 't. 1 0 ), 2 T' q <588./*FAeferenceT at RATED THERMAL POWER, ave x,~~
.== := :, 0.0040

a 1:,-
= A P = Pressurizer pressure, psig. .P' = 2235 psig (Nominal RCS operating pressure), 3 { y j in S = _g I Laplace transform operator, sec j and f (AI) is a function of the indicated difference between top and bottom detectors 'f g of the power-range nuclear ton chaebers; with gains to be selected based en measured i l instrument response during plant startup tests such that: { -2% +9.0% (1) for q q, between -94K and ". 0". f_ ;. O!, '' - ' ' " (" 't 1); f (AI) = 0, where*q aRd g are percent RATED THERMAL POWER in the top and botten g g j g,,,,g the core respectively, and gg '+ g is total MM NR in penent of RATED THERMAL POWER; i \\ -n% ~ j (ii) for each percent that the magnitude of g q exc gs.ggg su-as
==r -se n g b i j ?i the AT Trip Setpoint shall be automatically reduced by 1.1' r '" 't 0;,'3 151% '" 't 13 i of its value at RATED THERMAL POWER; and MI w' ) (fil) E ? for each percent that the magnitude of q, - exceeds +e:tK '".'t ;;, ,.76 '" 't 1), the AT Trip Setpoint shall be au tically reduced by9:9etti '! L '" 't ?), 1.'".5 '".'t 1) of its value at RATED THEMIAL POWER. t v i l.50 % w l r' ~ i 1
9 k TABLE 2.2-1 (Continued) REACIOR TRIP SYS1EN INSTRUNENTATION TRIP SETPOINTS a E NOTATION (Continued) E_ DVERPOWER AT NOTE 2: A T (I * S) ( 1 I iS 1 1 ) g,,5 1 + r s $ AT (K -K ( -) ( ) T -K [T( )- T"] - I)) 3 a o 4 5 1 + t,5 1 + t.S 6 1+tS l Where: AT = As defined in Note 1 I I L .t g, = As defined in Note 1 1. Ya = As defined in Note 1 3 3,,,$ As defined in Note 1, = AT = As defined in Note 1 o o K 1 1. % %. R G.. Z_, _.. G, 4 K = 0.02/*F for increasing average temperature and 0 for decreasing average 3 temperature, tS y The function generated by the rate-lag controller for T,,, dynaalc = 3,,3 compensation. t, Tlee constant ultitzed in the rate-lag controller for T,,,, r,& = i 5 sec (".!t.-- > = c), i,i 1 3,,g As defined in Note 1 = I fi f e .~ 1 T. As defined in Note 1 = j i t ! K - " 4*M'".' (i R 2h 0.#16W (M U for T > T" and
~~y }~~
6 K6 = 0 for T $ T", f / U i i
TA8tE 2.2-1 (Continued)
(

i REACTOR TRIP SYSTEM INSTRtBIENTATION TRIP SETPOINTS 5
NOTATION (Continued) E j T = As defined in Note 1 1 f 1588./*F Reference T,, at MD Mm mR, . T" = I S = As deft wd in Note 1, and f (AI) = y 0 for all AI. ' Note 3: The channel's maalaus Trip Setpoint shall not exceed its camputed Trip Setpoint by more than 2%. 7 M i I t e .i 3 j ii ! O T j 49
l I i 2.1 SAFETY LIMIT 5 i SA5ES l l 2.1.1 REACTOR CORE The restrictions of this safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the.reacter coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface i i temperature is slightly above the coolant saturation temperature. [ j Operation above the w e boundary of the nucleate belling regian could result in excessive cladding tamperatures because of the onset of departure from nucleate belling (OW) and the resultant sharp reduction in heat transfer ccefficient. 04 is not a directly seesurable parameter during operation and ' therefore THENL POWER and reacter caelant temperstare and pressure have been "related to ONS f.hrough the WW-1 correlation. The WAS-1 ONS correlation has'been i developed to pesdict the DNS flum and the location of'OW for axially unifem i and nonunifom neat flux distributions. The local ONS heat flux retis (OWR), { defined as the rette of the heat flux that would cause 04 at a particular core location to the local heat flux, is indicative of the margin to CNB. i t t l The minimum value of the ONOR during steady-state operation, nereal / tional transients, and anticipated transients is limited to 1.30 (base ( r i upon correlation). This value ca..;:;: 2 to a SSE probability at 4 55 j j conff level that DNS will not occur and is chosen as an apprope t l mergin to for all operating conditions. 4 j The curves e igures 2.1-1 and 2.1-2 show the leci of nts of THERMAL i POWER, Reactor Cael System pressure and average temper ure for which the j minimum DNBA is no less 1.30, or the average on py at the vessel exit j is equal to the enthalpy o aturated liquid. l .These curves are based on enthalpy channel facter, Fh,An of 1.55 ( q ) and a reference cosine with a peak 1.55 e axial power shape. } j allowance is included for an increase at reduced power based on the emeression: g j Fh=1.55(1+0.2 )) f Where P is the. action of RATED THE N L R. These lietti t flux conditions are higher than a calcalated for j the range of a entrol rods fully withdrawn to_ the maximum lewable control j red inserti swing the antal power imbalance is within the its of the l f3 (AI) f ten of the Overtamperature trip. When the antal powe abalance i is not thin the talerance, the antal power imbalance effect on the r-ature AT trips will reduce the Setpoints to provide protection con tent core Safety Limits. ~ s 6 8 2-1 e l l
l d $AFETY LIMITS 4 i ' RASE 5 4e M a h 4, lie M design basis is as follows: there must be at least a 905 probability that the minious OMA of the limiting red durig Condition I and t II events is yester then er aquel to the OWR limit of the M correlatten being used (the ife-1 correlation in this application). The correlatten OMR set such i that there is a M probability with 985 confidence that M'will not occur uhen i the sinteum M R is at the OW R limit. l i In meeting this desip basis, uncertainties in plant operating parameters, 1 ) nuclear and thereal parameters and fuel fabrication parameters are censidered l i statisticallysuchthatthereIsatleasta985confidencethattheminieueONBR for the limiting red is greater then er aquel to the ONBR limit. The uncertainties 1 in the above plant peranaters are used'te determine the plant OWR uncertainty. This OSR uncertainty, combined with the correlation DNBR 11eit establishes a desip OSA value which must be met in plant safety analyses usIng values of input j parameters without uncertainties. i i The curves of Figure 2.1-1 show the 1eci of points of THElutu. POWER, Reacter - Coelant System pressure and average temperature below which the calculated OM R is no less than the desIp OSR va' us or the average enthalpy at the vessel exit is less then the enthalpy of saturated liquid. j The curves are based en a nuclear enthalpy rise het channel factor, of i 1.4g and a reference cosine with a peak of 1.55 for amial power shape. An allow-l ance is included for an increase in at reduced power based on the expression: ~ ha1.4g(1+0.3(1-p)] [ ndhere p is the fraction of RhTED TMNHL POWER. These lietting heat flum conditions are higher then those calculated for the range of all control rods fully withdrawn to the maximum alleueble control rod t insertion assuming the axial power labelsace is within the limits of the ft (AI) l function of the Overtamperature trip. idhen the axial power imbalance is not within I the talerance, the asial power fatalance effect en the Overtamperature At trips will reduce the setpoints to provide protection consistent with core safety limits. j = r 2.1. 2 4 TACTOR COOLANT SYSTEM PRES $URE The restriction of this Safety Limit protects the integrity of the Reacter Coelant System from overpressurization and thereby prevents the release of radio-nuclides contained in the reacter coolant from reaching the containment atmosphere. The reacter versel and pressurizer are ' designed to Section !!! ef the ASME I Code for Nuclear Pouer plants uhich permits a emaisum transient pmssure of 115 (2735 peig) of design pressure. The Safety Limit of 2738 peig is therefore consistent with the desip criteria and associated code requirements. [ The entire teactor'Ceelant System is hydretested at 3107 peig, 1255 of design pressure, to demonstrete integrity prior to initial operation. 'O t-2 -. - - -., - - - - ~ - _ww,-_ ww - r .,'.m_ 8
4 LIMITING !AFETY SYSTEM SETTINGS BASES Power Ranee. Neutron Flux (Continued) The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 105 ef RATED THElWIAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Rance. Neutron Flux. High Rates The Power Range Pcsitive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specific. ally, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. The Power Range Negative Rate. trip provides protection for control rod drop accidents. At high power, a rod drop accident of a single or multiple rods could cause local flux peaking which could cause an unconservative local ONBR to exist. The Power Range Negative Rate trip will prevent this from i occ'Jrring by tripping the reactor. Its credit is taken for' operation of the l Power Range Megative Rate trip for those control rod drop accidents for which 7 i.;f DNSR ml.a. Dh4R's will be greater than'Det & Jesi l i Intermediate and Source Ranos. Neutron Flux _ i The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigste the consequences of an uncon-j trolled rod cluster control assembly bank withdrewal from a subcritical l condition. These trips provide redundant protection to the Low 5etpoint trip of the Power Range, Neutron Flux chgnnels. The Source Range channels will i initiate a Reactor trip at about 10 5 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25X of l RATED THERMAL POWER unless manually blocked when P-10 becomes active. l _-N .L:- _ -- 7 8 2-4 AsendmeaLNo." (u.4e-ry- .bmenenetw noMt *Q 6 9 e r w-- w-,, e -- --v-- .--o--- ~ ~ - -,v------- ,m-------mnr--,-w-n-,~,,,,r,e-w,,-,--rww..--r-mm-,w- - m ewe-og---~
) 3f4.1 REACTIVITY CO' TROL SYSTEMS. N 3/4.1.1 80 RATION CONTROL SHUTOOWN MARGIN - T_v_ >200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 4,45-deheeh- '" nit O,1.35 delta k/k (...:. l,' for four loop operation. APPLICA8ILITY: MDOES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MMtGIN less than 1."_ ti L'i '".it O,1.3% delta k/k '" nit L;, immediately initiate and continue boration at greater than or equal to 30 spa of a solution containing greater than or equal to 7000 ppe boron or equivalent untti the required SHUTDOWN MMIGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1. ~_1.'. ^ .,/L (... :t Ok 1.3% delta k/k (L".. O. 4 Within 1 houi* after detection of an inoperable control rod (s) and at a. least once per 12 nours thereafter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); i b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1.0 at least once per 12 hours by verifying that control bank withdrawai is within the limits of Specification 3.1.3.6; When in MODE 2 with-X,ff less than 1.0, within 4 hours prior to c. achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6; l l d. Prior to initial operation above 55 RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specification 4.1.1.1.le., below, with the control banks at the arximum insertion limit. of Specification. 3.1.3.6; and
~~See Special Test Exception 3.10.1.~~
~ Z : 6 T ; ; _... 3/4 1-1 _ ;.. n: M: '" U "--r- _ : 1.L, (L.. :10 e ~ _ -, -., - - - a --~,,,-,--,,y----,,we m-- ,w,,,,--,,--w.,---,-o,,e-w,_,,.. - -.. -,, -.,---,,,,,v- -,m.,,,---<--mm-,-,,,,,7 pe--. .y,,,,,
r REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION ~ ~ \\ 3.1.1.3 The moderator temperature coefficient (MTC) shall be: l a. P: "r.it ess positive than the Ifmits s,.hown in Figure 3.1-0, and .. o_4. i... __4.<__ o e_,+. of,.,... __m_ m ;<..; o,s.7,ie 34,. rane), 6.. _.. z z._ b.f. T;r"..;;.1::d",9essnega_tivethan-4.1x104-- : '"5"".ol "?"E" :::dit';;; and delta k/k/*F for the. all rods withdrawn, and of cycle life (EOL), RATED THERMAL POWER condit APPLICABILITY: Specification / 3.1.1.3a. : d 2.1.1.05. - MODES 1 and 2* only.# Specification 3.1.1.3(6 - MODES 1, 2, and 3 only.J ACTION: With the MTC more positive than the limit of Specificationg 3.1.1.3a. a. cr 2.1.1.35, above, operation in MODES 1 and 2 may proceed provided: 1. c^- "- 1, dontrol rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the Ifmits shown in Figure 3.1-0 within 24 hours or be in HOT STAN08Y within the next 6 hours. These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 9 ro. n.4+ g, ,..+.ei ..a m4+6a.._.i 34.4+, e...wie.u.a ..<-.a ._,,4.4 .u_ uve..,___ ___e..._ .m__ ~'riEElU4)Ae 555 55 bEEEE'Er'E5 2- ^ ~5Idi"EaE9'II55I.~55A , s.... vs_,_ _,.me_,.., ,3_3.. 3.,, 3;,_ ees.3 +k. 4. ..+4.. 34-4 5 -,c__gisag..{._ g 1 3 c; 7- )I. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3 /. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its ifmit' for the all rods withdrawn condition. b. With the MTC more negative than the limit of Specification 3.1.1.3/h above, be in HOT SHUTDOWN within 12 hours.
~~With K,77 greater than or equal to t.0.~~
?
~~See Special Test Exception 3.10.3.~~
":" :^- una.. 1 uu 2 3/4 1-4 ._..-- nu nu.ac tun.6 12 .___.._.-...nu... w....y I
REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE0VIREMENTS-4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: a. The MTC shall be measured and compared.to the BOL limit of Specification / 3.1.1.3a. rd 2.1.1.05., above, prior to initial operation above.5% of RATED THERMAL POWER, after each fuel loading; and b. The MTC shall be measured at any THERMAL POWER and compared to ~4 -3.2 x 10 delta k/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD aftef reaching an equilibrium baron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.2 x 10 4 delta k/k/*F., the MTC shall be remeasured, and compared to the EOL MTC limit of Specifica-tion 3.1.1.3/b, at least once per 14 EFPD during the remaihder of the fuel cycle. i I ":CU!"i ""!TS 1 :nd 2 .. ' 3/4 1-5 -f rdrnt 9:. :: ( U n ' 'e-1-)- f. nd ;nt M:.10 (L'ait 2)
.N t ~ R 0.5 a. T= W e 0.4 - 5 5 Acceptable UnacEactable Operation Oceration 5 0.3 - v ,= s 0.2 - E edg 0.1 .2
~~b I~~
~~0 10 . 20 30 40 50 60~~
~~70 80 90 100~~
/ % of Rated Themal powr FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT V5 POWER LEVEL "=" 'E e a 0 M Mtf7Bf _ f tM T*P P 9 a m 3/4 1-Sa rii 4. n ^-- t ^-^ ' U- $9E_ Og y. . - - -. - _ _ _ _....... gg, _.. G = e - _ - -., _ ~ -. - - -.., - - - - _ _, _ _ _ _, _ _ _ _ _ _ _, _. _ _ _, _. _ _
l? l l POWER DISTRIBUTION LINITS i 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RTSE HOT CHANNEL FACTOR LINITING' CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R W shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation: t Where:
k.
~~Y Y WW a~~
~~1.49 [1.0 + 0.3 (1.0 - P)]~~
,J, 4PTI. o + 0. 2 W a (g,,, ~Zf = } y ,o M _Rs (Unit li =,, s l i THERMAL POWER i b.)$ P = g g, F"g = Measured values of F" obtained by using the movable incore l g )( g l detectors to obtain a power distribution map. The measured l values of F" shall be used to calculate R since Figure 3.2-3 g includes penalties for undetected feedwater venturi fqpling of
~~0. N and for measurement uncertainties of 1. 3 for flow and 45 for incore measurement of F" M~~
g X-Penalty as a function of region aver =_a= M....,, as ' shown in s defined as those assemb1 ame ads) or enrich-first core). (Applies to Unit 2 only. -.. s APPLICA8ILITY: MDDE 1. f ACTION: With the combination of RCS. total flow rate and Rx M outside the region of acceptable operation shown on Figure 3.2-3: a. Within 2 hours either: 1. Restore the combination of RCS total flow rate and Rx J( to within the above limits, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to SSE of RATED THERMAL POWER within the next 4 hours. W :: TU: 1 -- - 3/4 2-14 1
POWER OISTRIBUTION LINITS LINI.T'ING CONDITION FOR OPERATION ACTION: (Continued) Within 24 hours of initially being ou) side the above [imits, verify h. through incore flux mapping and RCS total flow rate comparison that the combination of Re, M and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than SE of RATED THERMAL POWER within the next 2 hours. Identify and correct the cause of the out-of-limit condition prior c. to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2. and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of Ag X and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the reGon of acceptable operation shown on Figure 3.2-3 prior to excelding the following THERMAL POWER levels: i 1. A nominal 5GE of RATS THERMAL POWER, 2. A nominal 75% of RATED THERMALPOWER, and 3. Within 24 hours of attaining greater than or equal to 95% of i RATED THERMAL POWER. SURVEILLANCE REQUIREENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 1 4.2.3.2 The combination of indicated RCS total flow rate determined by process computer readings or digital voltmeter measurement and Rs. 2nda E shall be within the region of acceptable operation of Figure 3.2.3: a. Frior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b. At least once per 31 Effective Full Power Days. 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours when the most recently obtained values of Rar 2nd=E, obtained per Specification 4.2.3.2, are assuesd to exist. 4.2.3.4 The RCS total flow rate indicators shal[ be subjected to a CHANNEL CALIBRATION at least once per 18 months. l 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. 2Gt!!2D==WNWM - /_ 3/4 2-15 e
/ ~ f~ $ .s -th PENALTIES OF 0.1% FOR UNDETECTED FEED- ) E 4s WATER VENTURI FOULING AND MEASUREMENT t
~~l UNCERTAINTIES OF 1.7% FOR FLOW AND 4%~~
3 FOR INCORE MEASUREMENT OF F ARE g c: INCLUDED IN THIS FIGURE. H l s ----... / 0 -#L'ud ) i a 3 ( N ACCEPTABLE l I OPERAT 3 42 REGION g _r _ _ _ _ _ - t o "C" F id N 1 o i s e UNACCEPTABLE t OPERATION 'g REGION l m d ACCEPTABLE OPERATION Rl!GION FOR 690% RTP (1.0, My i $90% RTP (1.0, M.49W 38 3 ) 594% RTP (1.0,M.11h l I' $92% RTP '{1.0, 377III $90% RTP (1.0,37h f. 11.0,3me46 34.444 a O 38 o P 0.00 0.92. 0.s4 0.98 0.se 1.00 1.02 1.04 1.06 ,1 I I ,Ra = F 11.49113 + o.3 4&l1APil i 4 x== :: a. Figure 3.2-3b RCS FLOW RATE VERSUS Rx3nd Ag.- FOUR LOOPS IN OPERATION (Unit 2) i
i -~ l l a l g S \\. l ~ / a E g s t E p ~$ w i / a a a< g W c n a l Ld I 1 3 W \\ o .a ni E E e = N 4 M 4 E i mousvum nwmu me ocu L'... ;. 1 --- C 3/4 2-18 t... f-at. to.- Ameedumr o.13 G- * + 0,,

w

,---,e-, ,-,e-- ,ww r ,w-,~_n --n,,, w m_..--m_,_n__..---_,
t' 3 M. m \\ H Ii ! TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUNENTATION RESPONSE TIMES !E RESPONSE TIME {l FUNCTIONAL UNIT 1. Nanual Reactor Trip N.A. 'f. j f, 2. Power Range Neutron Flux 1 0.5 second* I 3. Power Range, Neutron Flux, N.A. High Positive Rate 4. Power Range, Neutron Flux, 1 0.5 second* High Negative Rate 1 R 5. ' Intermediate Range, Neutron Flux N.A. t N.A. l Y 6. Source Range, Neutron Flux l 7. Overtemperature AT seconds
8.
~~Overpower AT seconds~~
9.
Pressurizar Pressure--Low $ 2.0 seconds i l 10. Pressurizer Pressure--High 5 2.0 seconds 11. Pressurizer Water Level--High N.A. 1 Neutron detectors are exempt from response time testing. ' Response time of the neutron flux signal portion a of the channel shall be measured from detector output or input of first electronic component in channel. 9 9
TABLE 3.3-2 (Continued) I ' l l REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES i; r 1 FUNCTIONAL UNIT RESPONSE TIME ! i q 12. Low Reactor Coolant Flow n u a. Single Loop (Above P-8) $ 1.0 second l l b. Two Loops (Above P-7 and below P-8) 1 1.0 second d ' 3.5 l u 13. Steam Generator Water Level--Low-Low 1 4,4 seconds.
~~14. Undervoltage-Reactor Coolant Pumps~~
~~< 1.5 seconds~~
~~15. Underfrequency-Reactor Coolant Pumps~~
~~< 0.6 second~~
~~16. Turbine Trip.~~
w5 a. Low Fluid Oil Pressure N. A. ' w b. Turbine Stop Valve Closure N.A. l l 17. Safety Injection Input from ESF N.A. 18. Reactor Trip System Interlocks N.A. i 19. Reactor Trip Breakers N.A. 4 20. Automatic Trip and Interlock Logic N.A. i i
~~c TABLE 3.3-4 (Continued)~~
~~P l 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS i l FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES i i~~
(
7. Auxiliary Feedwater [ a. Manual I'nttiation N.A. N.A. U b. . Automatic Actuation Logic M.A. N.A. and Actuation Relays c. Steam Generator 51.8% Water Level--Low-Low > 12%'of span from 0 to 11% of span from 0 to 1) Start Motor-Driven Pumps 40.Gk>0% of RATED THERMAL POWER, J >0% of RATED THERMAL POWER, 3 3 m ncreasing linearly to reasing linearly to ) 44rilK of span at 100% of 6iMJN of span at 100% of NATED THERMAL POWER. NATED THERMAL POWEL j 4 t 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, i 40.02v1reasing linearly to increasing linearly to
~~7'4WJK of span at 100% cf of span at 100% of j~~
LATED THERMAL POWER. NATE RMAL POWER. i 31.0 7 > 2 psig > 1 psig. l d. Auxiliary Feedwater Suction Pressure - Low { Realignment) (Suction Supply Automatic j e. Safety Injection - See Item 1. above for all Safety Injection Trip Setpoints Start Motor-Driven Pumps ,and Allowable Values i f. Station Blackout - Start 3464
~~173 volts with a~~
> 3200 volts Motor-Driven Pumps and 0.5 i 0.5 second time i Turbine-Driven Pump delay l! g. Trip of Main Feedwater Pumps - N.A. N.A. Start Motor-Driven Pumps j O O.. 79
~ INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERA 8LE with: a. At least 755 of the detector thisibles, b. A minimum of two detector thimbles per core quadrant, and c. Sufficient movable detectors, drive, and readout equipment to map these thimbles. APPLICAEILITY: When the Movable Incore Detection System is used for: 4 a. Recalibration of the Excore Neutron Flux Detection System, b. Monitoring the QUADRANT POWER TILT RATIO, or N c. Measurement of Fg, p (7) =; q (% ACTION: With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 arg not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERA 8LE at least once per 24 hours by normalizing each detector output when required for: a. Recalibration of the Excore Neutron Flux Detection System, or b. Monitoring the QUADRANT POWER TILT RATIO, or Measurement of F", F (Z), rd e c. g q I l '. "d'J:": - U"IT'; 1..,4 2 3/4 3-45 f I
i l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION i 3.5.1.1 Each cold leg injection accumulator shall 'be OPERA 8LE with: a. The isolation valve open, - b. A contained borated water voTume of between; 8022 and 8256 gallons (U;tt 1),
)
"T. c.d 00^' ;"::: (U;ft *), A ' boron concentration of between 1900 and 2100 ppe, c. d. ~ A nitrogen cover pressure of ' etween 430 and 484 psig (Unit 1), b '^^ r.: '5'. ;;t;; ('O.it :), and e. A water level and pressure' channel OPERA 8LE. C APPLICABILITY: M00ES 1, 2, and 3*. ACTION: ~. With one cold leg injection accumulator inoperable, except as a result a of a closed isolation valve, restore the inoperable accumulator to OPERA 8LE status within I hour or be in at least HOT STAN08Y within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isola-tion valve or be in at least. HOT STAN08Y within 1 hour and in HOT SHUTDOWN within the following 12 hours. SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each cold. leg injection accumulator shall be demonstrated OPERA 8LE: a. At least once per 12 hours by: 1) Verifying the contained borated water volume and nitrogen cover pressure in the tanks, and 2) Verifying that each cold leg injection accumulator isolation l valve is open. l i l
~~Pressurizer pressure above 1000 psig.~~
l . ;^. ": U":T; 1 *. 4 2 3/4 5-1 -? -"5 3 23 5.5"." N
3/4.1 REACTIVITY CONTROL SYSTEMS ~ BASES ~ 3/4.1.1 80 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 GHUTDOWN MARGIN A sufficient SHUTDonel MMIGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent. criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,,g. The most restrictive-cordition occurs at E0L, with T,, at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown.' In the analysis of this accident, a minimum SHUTDOWN NARGIN of ' '" :' 2 i'i ('frit U,1.35 delta k/t ('frit O is required to - control the reactivity transient." Accordingly, the SHUTDOWN MMIGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200*F, the reactivity' transients resulting from a postulated steam Ifne break cooldown are minimal and a 1% delta k/k SHUTDOWN MARGIN provides adequate protection. 3/4.1.1.3 MODERATOR TEMPERATURE C0 EFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting i condition assumed in the FSAR accident and transient analyses. The MTC values of this specification are appifcable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other i than those expifcitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator density coefficient (M C), was obtained by incrementally correcting the M C used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MC associated with a core condition of all rods inserted (most positive MC) to an all rods withdrawn l candition and, a conversion for the rate of change of moderator density with i 1 temperature at RATED THERMAL POWER conditions. This value of the MC was then transformed into the Ifniting MTC value -4.1 x 10 4 delta k/k/*F. The MTC value of -3.2 x 10
~~delta k/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppe~~
. equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10 4 k/k/*F. m,s M ui umaA .- c 8 3/4 1-1 {"] [ 3 % { O I _,e,- -,,,,.-..w.. .,,,wn_e,,,.,,n -,,,___n, ,m,_____.w. w,.,,, ,,_w,,,,-
e i REACTIVITY CONTROL SYSTEMS SASES MDOERATOR TEMPERATURE COEFFICIENT (Continued) i The Surveillance Requirements for measurement of the MTC at the beginning i and near the end of the fuel cycle are adequate to confirm that the MTC. remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINDRM TDFERATURE FOR CRITICALITY This specification ensures that the reactor will not be inade critical with the Reactor Coolant System average temperature less than 551*F. This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressuriaar h capable of being in-an OPERA 8LE status with a steam bubble, and (4) the reactor vessel is above its minimum RTNOT**'P*"**" 3/4.1.2 80 RATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to i perform this function include: (1) borated water sources, (2) charging pumps, (3) separata flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systans, and (6) an emergency power supply from OPERABLE diesel generators. g3 4 With the RCS average temperature above 200*F, nimum of two boron injection flow paths are required to ensure single unctional capability in the event an assumed failure renders one of the f ow paths inoperable. The boration capability of either flow path is suff ient to provide a SHUTDOWN MARGIN from expected operating conditions of delta k/k after xenon decay j and cooldown to 200*F. The maximum expected boration capability requirement l occurs at EOL from full power equilibrium xenon conditions and requires 16,321 gallons of 7'J00 ppe borated water from the boric acid storage tanks or 75,000 gallons of 2000 ppe borated water from the refueling water storage tank l (RWST). With the RCS temperature below 200*F, one Boron Injection System is acceptable without single failure consideration on the basis of the. stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Soron Injection System becomes inoperable. The Ifnitation for a maximum of one centrifugal charging pump to be OPERA 8LE and the Surveillance Requirement to verify all charging pumps except the required OPERA 8LE pump to be inciperable below 300*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORY. ~ l " "I"' Z:1 4 ..- 4 8 3/4 1-2 l \\ __,,-r 7--m-m----P+-w-----w t"- '" --** " ' ' * ' ~ ' ^ " "
4 POWER DISTRIBUTION LIMITS BASES HEAT FLUX HDT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACTOR (Continued) .a. Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position; b. Control rod groups are sequenced with overlapping' groups as described in Specification 3.1.3.6; c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d. The axial power distribution, expressed in tems of AXIAL FLUX DIFFERENCE, is maintained within the limits. F" will be maintained within.its limits provided Conditions a. through g d above are maintained. As noted on Figurts; 3.2-3 6. RCS flow rate year u and p be " traded off" igsj g r} i.e., a low seasured RCS flow rate is acceptable if the z x n ".,.: :.:: ?=) to ensure that the calcu-lated Dh8R will not be below the design DNBR value. The relaxation of F as a function of THERMAL POWER allows changes in the radial power shape for all I pemissible rod insertion limits. Rg as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts for F" less than or equal to 1.49. This value is used in the various accident g analyses where F" influences parameters other than DN8R, e.g., peak clad tem-g perature, and thus is the maximum "as amasured" value allowed. T. ^ ' ' - ' for the inclusion of a penalty for Rod.8ow on DN8R only. Thus, know a the "as ured" values of F" and RCS flow allows for " tradeoffs" xcess g of R equal to 1. r the purpose of offsetting the Rod Bow penalty. Fuel rod bowing reduc he value of DNS ratio redit is available to partially offset this redt:ction. is credit s from a generic or plant-specific design margin. For McGuire , the margin used to partially offset rod bow penalties is 9.1%. T n breaks down as follows: 1) Design limit DNBR I.65 2) Grid spacing 2.95 3) The ffusion Coefficient 1.2% 4) D ltiplier 1.7% 5 itch Reduction 1.7% J e T .%:T: -- 7 B 3/4 2-4 9 ,,,,,,-.-.m- ,m.vve---m-,,.mm, e-,--,,
k e POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE NOT CHANNEL FACTOR (Continued) r, the margin used to partially offset rod bow penalties with the .2% issed to trade off against measured ng as much as 25 lower than the flow plus uncertaf he penalties applied to F to account for rod bow (Figu as a function of burnup are consistent with those in Mr. John F. Sto ) letter to T. M. Anderson (Wes e dated April 5, 1979 with the diffe due to the margin each unit uses to partially offset rod bow pena 9 i.. - - L E rgin between the safety analysis limit DM Rs (1.47 and 1.49 for thimble and typical cells, respectively) and the design limit DNBRs (1.32 and 1.34 for thiable and typical cells, respectively) is maintained. A fraction of this margin is utilized to accommodate the transition core DM R penalty (2%) and the appropriate fuel rod bow DNOR penalty (WCAP - 8691, Rev. 1) When an F measurement is taken, an allowance for both experimental error ~ q and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 35 allowance is appropriate for annufacturing tolerance. When RCS flow rate and F" are measured, no additional allowances are g necessary prior to comparison with the limits of Figurent 3.2-3 N l Measurement errors of 1.7% for RCS total flow rate and 4% for F have been allowed for in determination of the design DNOR value. The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater ves. turi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi.is included in Figure 3.2-3. Any fouling which might bias the RCS ficw rate measurement greater than 0.3% can be detected by monitoring and trending various plant perfonsance parameters. If detected, l action shall be taken before perfoneing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be l cleaned to eliminate the fouling. l The 12-hour periodic surveillance of indicated RCS flow is sufficient to l detect only flow degradation which could lead to operation outside the accept-l able region of operation shown on Figure 3.2-3. _N w a ._.----..---m,. - + -.. - - -, -. - -. - - .....,#+-4--~mmy-e+- -.ww
t' ) s ' Analysis of Significant Hazards Consideration As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10 CFR 50.92. Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety. The proposed amendments would change plant operating limitations given in the Technical Specifications affected by the use of the optimized fuel assembly design in McGuire Unit 2 Cycle 2. The reference safety evaluation report submitted by Mr. H. B. Tucker's Nov' ember 14, 1983 letter to Mr. H. R. Denton summarizes the evaluation performed on the region-by-region reload transition from the McGuire Units 1 and 2 standard (STD) fueled cores to cores with all optimized fuel (OFA). The report examines the differences between the Westinghouse STD design and 0FA design and evaluates the effects of these differences for the transition to an all 0FA core. The report justifies the compatibility of the OFA design with the STD design in a transition core as well as a full 0FA core. The report also contains summaries of the mechanical, nuclear, thermal-hydraulic, and accident evaluations. The McGuire Unit 2/ Cycle 2 reload safety evaluation (Attachment 2A) presents f an evaluation which demonstrates that the core reload will not adversely affect the safety of the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload were reviewed for the ? Unit 2 Cycle 2 design. The results of new analyses are included in the reference safety evaluation report and the Unit 2/ Cycle 2 RSE, and the justification for the applicability of previous results for the remaining analyses is presented. Ei The results of evaluation / analysis and tests lead to the following conclusions: a. The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mechani- ((- cally compatible with the current STD design, control rods, and reactor internals interfaces. Both fuel assemblies satisfy the current design bases for the ggy McGuire units, a=- 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 current STD design, d. The accident analyses for the OFA transition core were shown to provide acceptable results by meeting the applicable criteria, such as, minimum DNBR, peak pressure, and peak clad temperature, as required. The previously reviewed and licensed safety limits are met.
m p1 - };
~~e. _ Plant operating limitations given in the Technical. Specifications will.~~
~~'be satisfied with the proposed' changes. - s i From these evaluations, it is concluded that'the Unit 2 Cycle 2 design does~~
~~not cause the previously acceptable safety limits'to be exceeded.~~
The commission has provided examples of amendments likely to involve no signi-ficant hazards considerations (48 FR 14870). One example of this type is (vi), "A change which either may result in some increase to the probability or conse-quences of a previously analyzed accident or may reduce in some way a safety i margin, but where results of the change are clearly within all' acceptable criteria with respect to the system or component specified in the standard review plan: for example, a change resulting from the application of a small refinement of a previously used calculational model or design method". Because the evaluations previously discussed show that all of the accidents comprising the licensing bases which could potentially be affected by the fuel reload were reviewed for the Unit 2 Cycle 2 design and conclude-that the reload design does not cause the previously acceptable safety limits to be exceeded, the above example.can be applied to this situation. In addition, the NRC has previously concluded that similar changes'for McGuire Unit 1 Cycle 2 did not involve significant hazards considerations-(these changes were subsequently approved via Ms. E. G. Adensam's (NRC/ONRR) letters to Mr. H. B. Tucker dated April 20, June 21, and September 13, 1984). Based upon the preceding analyses. Duke Power Company concludes that the proposed amendments do not involve a significant hazards, consideration. L.- ..m._, ,.,..,. - -. - - -}}

v t e Letter Sequence Request
TAC:53319, (Open) TAC:53320, (Open) TAC:56336, (Open) TAC:56337, (Open) TAC:56476, (Approved, Closed) TAC:56477, (Approved, Closed) TAC:56873, (Approved, Closed) TAC:56874, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML20082D639, ML20082D657, ML20084C989, ML20086P477
MONTHYEAR1984-02-20 [Table View]

ML20099G748

Text

Subject:

Dear Mr. Denton:

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

1.1 INTRODUCTION

1.3 CONCLUSION

5.0 REFERENCES

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