Text

Proposed Tech Specs 2.0, Safety Limits, TS 3.3.1, Reactor Trip Sys Instrumentation & TS 3.4.1, RCS Pressure,Temp & Flow Departure from Nucleate Boiling (DNB) Limits
	ML20196H755
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Site:	Mcguire, Catawba, McGuire
Issue date:	06/24/1999
From:	; DUKE POWER CO.
To:	;
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ML20196H752	List: ML20196H747; ML20196H755;
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	NUDOCS 9907060465
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r M8 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 1

In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1 for four loop operation.

2.1.2 RCS Pressure SL in MODES 1,2,3,4, and 5, the RCS pressure shall be maintained s 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated: )

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3,4, or 5, restore compliance within 5 minutes.

McGuire Units 1 and 2 2.0-1 Amendment Nos. 184/166

~

9907060465 990624 PDR ADOCK 05000369 p PM

SLs 2.0 j 6 70 DO NOT OPER INTHIS AREA

\

I 650 -

osio -

640 -

2400 pelo 1

)

be 22 I

h620

2100 psio 610 -

45 psio 600 -

1 590 ACCEPTABLE OPERATION f f f f 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Roted Thermal Power gc. M p psd w VW Figure 2.1.1-1 Reactor Core Safety Limits -

Four Loops in Operation McGuire Units 1 and 2 2.0-2 Amendment Nos.

y

.< i SLs 2.0 670 DO NOT OPERATE IN THIS AREA 660 650 640 g 2400 psia P 630 cc 2280 psia h 620 2100 psia 610 600 1945 psia 590 ACCEPTABLE OPERATION 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits -

Four Loops in Operation ;

McGuire Units 1 and 2 2.0-2 Amendment Nos.

RTS InStrum:ntition 3.3.1 Table 3.3.1-1 (page 2 of 7)

Reactor Trip System Instrumentation APPUCABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

6. Overtemperature AT 1.2 4 E SR 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 1 (Page Note 1 SR 3.3.1.6 3.3.1 18) -(Page SR 3.3.1.7 3.3.1-18)

SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.17

7. Overpower AT 1.2 4 E SR 3.3.1.1 Refer to Referto SR 3.3.1.3 Note 2 (Page Note 2 SR 3.3.1.6 3.3.1-19) (Page SR 3.3.1.7 3.3.1 19)

SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.17

8. Pressurizer Pressure

a. Low 1(f) 4 M SR 3.3.1.1 2 1935 psig g 1945 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. High 1.2 4 E SR 3.3.1.1 s 2395 psig s 2385 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

9. Pressurizer Water j(f) 3 M SR 3.3.1.1 s 93 % s92%

Level- High SR 3.3.1.7 SR 3.3.1.10

10. Reactor Coolant Flow-

a. Single Loop 1(9) 3 per loop N SR 3.3.1.1 g f

z %

SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 g g

b. Two Loops 1(h) 3 per loop M SR 3.3.1.1 a . m /*

SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

11. Undervoltage RCPs 3(f) 1 per bus M SR 3.3.1.9 y 5016 V z 5082 V SR 3.3.1.10 SR 3.3.1.16

)

(continued)

(f) Above the P 7 (Low Power Reactor Trips Block) interlock.

(g) Above the P-8 (Power Range Neutron Flux)interiock.

J (h) . Above the P 7 (Low Power Reactor Tnps Block) interlock and below the P-8 (Power Range Neutron Flux) interlock, j I

McGuire Units 1 and 2 3.3.1-15 Amendment NoS.N i

I RTS instrumentation l 3.3.1 Table 3.3.1-1 (page 5 of 7)

Reactor Trip System Instrumentation ;

Note 1: Overtemoerature AT l

The Overtemperature AT Function Allowable Value shall not exceed the followng Trp Setpoint by more than 4.4% of RTP.

' ~

1 AT (1 + ris) ' 1 (1 +r,s) ,1 + ros , g 7, Jg, ,, g, ((1 1 + r, s) ,+

T r, s) - T' (1 + r, s) ,

+ K, (P - P') - f. ( AI)

Where: AT is measured RCS AT by loop narrow range RTDs, 'F.

ATo is the indicated AT at RTP, 'F.

s is the Laplace transform operator, sec-1.

T is the measured RCS avera erature *F.

T'is the nominal Tavg at RT ,4 585

F.

l P is the measured pressurizer ressure, psig P' is the nominal RCS operating pressure. = 2235 psig j

Ki l

= - Overtemperature AT reactor trip setpoint, as presented in t e COLR. I K2 = Overtemperature AT reactor trip heatup setpoint penalty coefficient. as l

presented in the COLR, K3 = Overtemperature AT reactor trip depressurization setpoint er.at i coefficient, as presented in the COLR, T ta = Time constants utilized in the lead-lag controller for aT. as :rese-ted h t e i

COLR, ta = Time constants utilized in the lag compensator for AT. as p ese :ed in tne COLR, 4

t4, is = Time constants utilized in the lead-lag controller for T,g. as crese .ted in the COLR, to = Time constants utilized in the measured T,y ag l compensa:or. as presented in the COLR, and, fi(Al) = a function of the indicated difference between top and bottern de: ecto s of the power-range nuclear ion chambers; with gains to be seec:e:: cased on measured instrument response during plant startup tests s ch tMr (i) for qi- go between the ' positive" and ' negative

f,(au brea< points as presented in the COLR; fi(AI) = 0, where qi and q- a e oe ent RATED THERMAL POWER in the top and bottorn Mlves of tr.e core respectively, and qi + go is total THERMAL POWER , :>e cent of RATED THERMAL POWER;

ent ued 1

McGuire Units 1 and 2 3.3.1-18 Amendmer: Nos.

RCS Pr ssura, Temp:raturs, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.

APPLICABILITY: MODE 1.

-NOTE- --- ---

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL. POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure or A.1 Restore DNB parameter (s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average to within limit.

temperature DNB parameters not within limits.

B. B.1 Reduce the P er Range 6 urs RCS tota %w rate in the region restricted Neutron FI - High Trip oper on of Figure Setpoin elow the nomin

3. -1. setp by the same i a unt as the pow imitation in Figur 3.4.1-1.

l (con ' ued) fu &SW A on uex f ye 1

s McGuire Units 1 and 2 3.4.1 -1 Amendment Nos. 8 66

n-Insert A B. RCS total flow rate B.1 Reduce THERMAL POWER 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

< 390,000 gpm but to < 98% RTP.

> 386,100 gym.

AND B.2 Reduce the Power 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Range Neutron Flux -

High Trip Setpoint below the nominal setpoint by 2% RTP.

(continued) r i

RCS Pr:ssura, Temp ratura, and Flow DNB Limits 3.4.1

. ACTIONS (continued)-

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rateW C.1 Restore RCS total flow rate 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months reg!en of prcht t;d to ;;;th r the res:Or. of opereH cf r;ggia rastricted ep;ratica, q^<< y, \te, to o qf ~ .

< ) fs, t o o it* - 98 C.2.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to < 50% RTP.

AND C.2.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux - High Trip Setpoint to s 55% RTP.

AND I Rekrs C.2.3 VerifpRCS total flow rate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

!e -"W the region of

. parm!rst!c or rcotrictod erarotter - {

b >f Vf,100 7' D. Required Action and D.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

l 1

/

McGuire Units 1 and 2 3.4.1-2 Amendment Nos. 8 166

i RCS Pr:s:urs, Tcmper tura, cnd Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS

' SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.2 Verify RCS average temperature is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.3 Verify RCS total flow rate is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l SR 3.4.1.4 Perform CHANNEL CAllBRATION for each RCS total 18 months j flow indicator.

]

I J

l l

1 l

i l

McGuire Units 1 and 2 3.4.1-3 Amendment Nos. 184/166

s :

RCS Pr:ssura, Temp::ratura, cnd Flow DNB Limits 3.4.1 l

i Table 3.4.1-1 (page 1 of 1) l RCS DNB Parameters l l

1

)

PARAMETER 1NDICATION No. OPERABLE LIMITS CHANNELS

1. Indicated RCS Average meter 4 s 500,54E Q 7.2 */3 Temperature meter 3 s 500.2 T TVs.1 *F computer 4 < g3,9 er. pgy,7 'f:

computer 3 _590.9 as rg 1.T #:

2 /

2. Indicated Pressurizer meter 4 2 2220.5F air LtIT S 7)

Pressure meter 3 2 2220.S peig 1.t.t.t..f Nig computer 4 2 2324d' pois t1.t r.5 [O computer 3 2 9224.2 psig ti.r7.fp 4:3

3. RCS Total Flow Rate -

b>)F:;;w 0.4.1-F3 90,o s a l

l l

McGuire Units 1 and 2 3.4.1 -4 Amendment No 166

RCS Pr:ssura, Temper:.tura, t Flow DNB Limits 3.4.1 PO R DISTRIBUTION LIMITS 385,820N

\, A penap y of 0.1% for undetected feedwater Permissible enturi fouling and a measurement uncertainty Operation o 1.7% for flow are included in this figure.

(98, 382,000) "'9'"

382,000 - - - --- ----------

378,180 e

g. Restri d

= Operati 5 negion (94, 374,360)

E 374,360 h

lii m 370, 40) Prohibited (9 operation e 370,540 3 Region 8

o

$ (90, 366,720) g 366,720

. 362,900 359,080 86 88 90 92 94 96 98 100 102 Percent of Rated Therma Power Figure 3.4.1-1 RCS OTAL FLOW RATE VERSUS RATED THERMAL POWER - FOUR LO S IN OPERATION McGuire nits 1 and 2 3.4.1-5 Amendment Nos.184/ 66

Rnictor Cors SLs B 2.1.1 i

B 2.0 SAFETY LIMITS (SLs) /)o B 2.1.1 Reactor Core SLs BASES

BACKGROUND GDC 10 (Ref.1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational

' transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

1 l

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding perforation, that would result in the release of fission products to the reactor coolant. Overheating of the fuelis prevented by maintaining the transient peak linear heat rate (LHR) below f the level at which fuel centerline melting occurs. Overheating of the fuel i cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the i cladding surface temperature is slightly above the coolant saturation temperature.

J Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor ;

coolant.

Operation above the boundary of the nucleate boiling regime could result

.in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water

- (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of

' activity to the reactor coolant.

The proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of the reactor core SLs.

1 i

McGuire Units 1 and 2 B 2.1.1-1 Revision No. 0 l

~

y R:c.ctor Cora SLs B 2.1.1 E

BASES //d APPLICABLE The fuel cladding must not sustain damage as a result of normal

- SAFETY ANALYSES operation and AOOs. The reactor core SLs are established to preclude -

violation of the following fuel design criteria:

a.' There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b. - The hot fuel pellet in the core must not experience centerline fuel

, melting.

l The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient.

conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided.by the following functions:

a. High pressurizer pressure trip;

b. Low pressurizer pressure trip;

c. Overtemperature AT trip;

d. Overpower AT trip;

e. Power Range Neutron Flux trip; and

f. Steam generator safety valves.

The limitation that the average enthalpy in the hot leg be less than or-equal to the enthalpy of saturated liquid also ensures that the AT measured by instrumentation, used in the RPS design as a measure of core power, is proportional to core power.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, 'RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed ,

initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

McGuire Units 1 and 2 - B 2.1.1-2 Revision No. O

r R: actor Cors SLs '

B 2.1.1 BASES O SAFETY LIMITS The curves provided in Figure B 2.1.1-1 show the loci of points of AT('F),

RCS Pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.

The curves in Figure 2.1.1-1 are based on a reference nuclear enthalpy rise hot channel factor (Fm), a reference axial power shape (Fz, x/L), the approved CHF correlation and the Technical Specification minimum flow rate. Therefore, these curves provide limits for which the analyses analyzed at the above reference values will be bounded. The curves in Figure B 2.1.1-1 illustrate the various RPS functions that are designed to -

prevent the unit from reaching the limit.

The SL is' higher than the limit calculated when tne AFD is within the limits of the Fi(Al) function of the Overtemperature AT reactor trip. When the AFD is not within the tolerance, the AFD effect on the Overtemperature AT reactor trips will reduce the setpoints to provide protection consistent with the reactor core SLs (Ref. 3).

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because those are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, " Reactor Trip System (RTS) Instrumentation." In MODES 3,4,5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 VIOLATIONS places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of ~ operation where this SL is not applicable, and reduces the probability of fuel damage.

REFERENCES. 1. 10 CFR 50, Appendix A, GDC 10.

2. UFSAR, Section 7.2.

3. DPC-NE-2011PA, March 1990.

McGuire Units 1 and 2 B 2.1.1-3 Revision No. O

g Reactor Cora SLs B 2.1.1 kp e1/ 4-i

- s s s s . .

. . s s . .

s s .

,. _____.s -

s s

..~~~. s **~ .

s Law Pummme . . s ~ . OP . g7 '

Ranemer Tdy s s s .

. . s s High Pesense

. s s Rammer' Dip s .

ws s p ', '

s .s s a t s g a .

. s s W . s %

s 4 %

~ s s- . s ..

s. s . s

. s s s s s i s s s s s %

s s %

s . s

% s %

seam acaener s

\ '.

sateer wives open s .

s s s %

, . \. .

. 73 ,9 (*f) 1

- - oraT CORE I.IMIn _.

Figure B 2.1.1-1 Illustration of Overtemperature and Overpower AT Protection

, McGuire Units 1 and 2 B 2.1.1-4 Revision No. O j

RCS Pressura SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES l BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization.' In the event of fuel claddirg failure, fission products L are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC.15, " Reactor Coolant System Design" (Ref.1), the reactor coolant

~

. pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and anticipated operational occurrences (AOOs).

Also, in accordance with GDC 28, " Reactivity Limits" (Ref.1), reactivity accidents, including rod ejection, do not result in damage to the RCPB !

greater than limited local yielding.

The design pressure of the RCS is 2500 psia. During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section 111 of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydroste. tic @r tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concems relative to limits on radioactive releases specified in 10 CFR 100,

" Reactor Site Criteria" (Ref. 4).

APPLICABLE The RCS pressurizer safety valves, the main steam safety valves SAFETY ANALYSES (MSSVs), and the reactor high pressure trip have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in

! Section 111 of the ASME Code for Nuclear Power Plant Components McGuire Units 1 and 2 B 2.1.2-1 Revision No. 0 '

4

RCS Pr:s:ura SL B 2.1.2 BASES M* T

j APPLICABLE SAFETY ANALYSES .(continued) l (Ref. 2), for anticipated operational occurrences. During the transient, no control actions are assumed, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the ' secondary plant safety valve settings, and nominal feedwater supply 1 is maintained. .j The Reactor Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs.

The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). . The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices.

More specifically, no credit is taken for operation of the following:

a. Pressurizer power operated relief valves (PORVs);

b. Steam Generator (SG) PORVs;

c. Steam Dump System;

d. Rod Control System;

e. Pressurizer Level Control System; or

f. Pressurizer spray valves.

SAFFTV LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code, Section lil, is 110% of design pressure. The maximum transient pressure allowed in the RCS piping, valves, and fittings under ASME Code Section 111 (Ref. 2) is 120% of design pressure.

The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig.

. 1 APPLICABILITY SL 2.1.2 applies in MODES 1,2,3,4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization i events. The SL is not applicable in MODE 6 because the reactor vessel i head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

McGuire Units 1 and 2 B 2.1.2-2 Revision No. 0

=

r.

RCS Prcssurs SL B 2.1.2 BASES A/a t-

/

SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, VIOLATIONS the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100,

" Reactor Site Criteria," limits (Ref. 4).

The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

If the RCS pressure SL is exceeded in MODE 3,4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the -

RCS pressure SL in MODE 3,4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code, Section Ill, 1971 Edition, Winter 1971 Addenda.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Article IWB-5000.

4. 10 CFR 100.

5. UFSAR, Section 7.2.

McGuire Units 1 and 2 B 2.1.2-3 Revision No. O

p i

RTS Instrumentation B 3.3.1

. BASES

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure' protection when below MODE 4.

9. Pressurizer Water Level-Hioh The Pressurizer Water Level-High trip Function provides a backup signal for the Pressurizer Pressure--High trip and also provides o

protection against water relief through the pressurizer safety valves. These valves are designed to pass steam in order to achieve their design energy removal rate. ' A reactor trip is actuated prior to the pressurizer becoming water solid. The - ,

setpoints are based on percent of instrument span. The LCO l requires three channels of Pressurizer Water Level--High to be OPERABLE. The pressurizer level channels are used as input to .

the Pressurizer Level Control System. A fourth channel is not '

required to address control / protection interaction concems. The level channels do not actuate the safety valves, and the high l pressure reactor trip is set below the safety valve setting.

. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip. j in MODE 1, when there is a potential for overfilling the pressurizer,

. the Pressurizer Water Level--High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow

, and the operator will have sufficient time to evaluate unit conditions and take corrective actions.

10. Reactor Coolant Flow-Low

a. Reactor Coolant Flow-Low (Sinale Looo)

The Reactor Coolant Flow-Low (Single Loop) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow.

Above the P-8 setpoint, which is approximately 48% RTP, a loss of flow in any RCS loop will actuate a reactor trip. The setpoints are based on a minimum measured flow of.95,50v -

[00 McGuire Units 1 and 2 B 3.3.1-16 Revision No I

I

RTS Instrum!ntation B 3.3.1

~ BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) gpm. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels !

per loop to be OPERABLE in MODE 1 above P-8.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is -

required to actuate a reactor trip (Function 10.b) because of !

the lower power level and the greater margin to the design limit DNBR.

b. Reactor Coolant Flow-Low (Two Looos)

! The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-8 setpoin'. a loss of flow in two or more loops will initiete a reactor trip. The setpoints are based on a minimum measured flow of W gpm. Each loop has three flow detectors to monitor flow. % TPd The flow signals are not used for any control system input.

i The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE.

l in MODE 1 above the P-7 setpoint and below the P-8 l setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern et this low j

> power level are unlikely. Above the P-7 setpoint, the reactor ,

trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power j level and the reduced margin to the design limit DNBR. ;

51cGuire Units 1 and 2 B 3.3.1-17 Revision No.[

RCS Pr ssura, Temp:ratura, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref.1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed.

The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will cause the reactor core to approach DNB limits.

The RCS coolant averags temperature limit is consistent with full power operation within the nominal operational envelope. Indications of temperature are averaged to determine a value for comparison to the limit. A higher average temperature will cause the core to approach DNB limits.

The RCS volumetric flow rate normally remains constant during an operational fuel cycle with all pumps running. Flow rate indications are yh averaged within a loop and then summed among the four loops to come ced

' f,.I.MX up with a value for comparison to the limitYRCS flow rate =d THERM ^L \

r f*DM j

...mm,u ... a-amu -,,r_.-__-- m _ _ _ u . . ._ i- r-:,. . . _ _ e 4 NE Ekto ensu're that h$caicu!EteEbbbillb$elelSw3e 'desEDbR ak* M"'M value(A lower HUS flow will cause the core to approach DNB limits.

Operation outside these DNB limits increases the likelihood of a fuel ,

cladding failure in a DNB limited event.

APPLICABLE The requirements of this LCO represent the initial conditions for SAFETY ANALYSES transients analyzed in the plant safety analyses (Rei.1). The safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the acceptance criteria, including the DNBR criterion. This is the acceptance limit for the RCS DNB parameters.

Changes to the unit that could impact these parameters must be I

McGuire Units 1 and 2 B 3.4.1-1 Revision No. l l

/ 1

iL 1 RCS Pr:ssura, Tamperitura, and Flow DNB Limits B 3.4.1 BASES j APPLICABLE SAFETY ANALYSES '(continued)

assessed for their impact on the acceptance criteria. A key assumption for the analysis of these events is that the core power distribution is within

. the limits of LCO 3.1.6, " Control Bank Insertion Limits"; LCO 3.2.3,

' AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, " QUADRANT POWER TILT RATIO (OPTR)."

gypf The pressurizer pressure limits and the S average temperature limits correspond to analyticallimits of ooo 79:nd 5^*.~" used in the safety analyses, with allowance for measurement un

, . ,g

' The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO, This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses. Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.

RCS total flow rate contains a measurement error of 1.7% based on the pedormance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow meas ement. Potential fouling of the feedwater venturi, ,

which might not haw been detected, could have biased the result from '

these past precision heat balances in a nonconservative manner.

Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance to 1.8% for no fouling.

]

The LCO numerical values in Table 3.4.1-1 for pressure andM- dare given for the measurement location with adjustments for the indication instruments.

~ APPLICABILITYJ in MODE 1, the limits on pressurizer pressure, RCS coolant average

' temperature, and RCS flow rate must be maintained during steady state !

operation in order to ensure DNBR criteria will be met in the event of an !

unplanned loss of forced coolant flow or other DNB limited transient. In i all other MODES, the power level is low enough that DNB is not a l concern.

I McGuire Units 1 and 2 B 3.4.1-2 Revision No.

,. RCS Pressure, Temp:r:tura, end Flow DNB Limits j B 3.4.1 )

i BASES I APPLICABILITY (continued)

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

Another set of limits on DNB related parameters is provided in SL 2.1.1,

" Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severc Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

ACTIONS .A_j, Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter (s).

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides ,

sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

]

%1pul 6!L E(lie,M oed Tf% RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is

>f N,10 0 SP* -asin ,ho rgien e r ee+,te,ori an=ratinn in Finura '4 a 1 5, then THERMAL OWER may not exceed *e H=a chama fa 'he ?!gurfin addition, the ;

q84 gn Power Range Neutron Flux - High Trip Setpoint must be reduced from the nomina setpoint by? am^"n' eqM'!O 'he THEP"^.L POWEP "m!!

l

em TP ""m- 5 Q;.,ie 0.4.M within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ' cr :=; ple, lb r

i 2% RTP m : = t p o w ; p 2 m e e .c e t .ne r .nc g : p-.p e: m _ ; n e rete:d by 'he $ffereaa= he'--^a 'he ma u aad o'm -h!^" : 4 'a. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and

there is no urgent need to reduce the trip setpoints. This is a sensitive l operation that may inadvertently trip the Reactor Protection System. ,

THtwM PentsQ *-usf 4 e- r1cd*'d 4 % 'L k* *L4 - '

% Gwls% %< eG 2 %s h tes,s.sind w:lt %res Ad;,, h,\.)

McGuire Units 1 and 2 B 3.4.1-3 Revision No.K

RCS Pressurs, Temp;rature, and Flow DNB Limits B 3.4.1 BASES l

ACTIONS (continued) hf 3 f(, l' O 9fC C.1 C.2.1. C.2.2, and C.2.3 (V4 goo gp%

If the indicated RCS total flow rate is wlthln th; reg lon of prohlbited epcra!!cn - gurc 3.1 * *, then clthcr 'he comb!a.2t!cn of RCS total flow )

cnd THEn".AL POWER must be restored tcWhc F"! chewn ;n the ;,gure

_ fnr rnetrintna nnoret!cn within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or power must be reduced to less g,;gn.f wI% than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with

), t,'. m 5.3.. Required Action A.1. If THERMAL POWER is reduced to less than 50%

( _ RTP, the Power Range Neutron Flux - High Trip Setpoint must also be reduced t.o s; 55% RTP, The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpointstecga'zer 'ho+ with nefer reduccd, th; safcty ana'ysl3 ;

- cccumpt!cne ara caCr'!ed and tScre ! nc urgent need to reduce th; trlp-

- ca'p^tn4. This is a sensitive operation that may inadvertently trip the g Reactor Protection System. Operation is permitted tu continue provided

( WN8W2 S the RCS total flow is #Ced ic bc *^ 'Sc reg!cn of perm!ccib!c or i >f 374 soo 9t^ j socenctad nnare60nhithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is

( _f reasonable cor sidering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient l are limited by the reduced trip setpoint.

.D_:.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not !

apply. To achieve this status, the plant must be brought to at least l MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable I to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify opcration is within safety analysis assumptions.

l McGuire Units 1 and 2 B 3.4.1-4 l

Revision No.k l

RCS Pr:ssura, Temperatura, cnd Flow DNB Limits B 3.4.1 BASES Mo ^ ---j &

SURVEILLANCE REQUIREMENTS - (continued)

SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

' SR 3.4.1.3 This surveillance demonstrates that the RCS total flow rate remains

. within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the' other indications available to the operator in the control room for monitoring the RCS flow rate and related equipment status (e.g. RCP voltage and frequency, etc.). The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify operation within safety analysis assumptions.

SR 3.4.1.4 Calibration of the installed RCS flow instrumentation permits verification tha the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.

The Frequency of 18 months is consistent with operating experience.

REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

McGuire Units 1 and 2 B 3.4.1-5 Revision No. 0 l

Attachment 1b Mdrked Copy of the Current Catawba Technical Specifications l

l l

i l

l l

4 I

4

hd

~

p, [L y

, SLs 2.0 2.0 SAFETY LIMITS (SLs)-

' 2.1 SLs 2.1,1 Reactor Core SLs in MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the SLs specified in Figure 2.1.1-1 for four loop operation.

2.1.2 : RCS Pressure SL in MODES 1,2,3,4, and 5, the RCS pressure shall be maintained s 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2,' restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3,4, or 5, restore compliance within 5 minutes.

Catawba Units 1 and 2 2.0-1 Amendment Nos. 173/165

SLs 2.0 g .s 670 00 NOTOPE ATE N THIS Ar4EA t i

660

)

i l

650 _ i 2455 1

640 2400 l

Em .

I h620 j 2100 pelo 610 -

1945 paio 600 -

590 -

ACCEPT TION 580 e i ao u 4 0 0.8 1.0 1.2 Fro tion of Roted ermal Power ,

os<A d Figure 2.1.1-1 (UNIT 1 ONLY)

Reactor Core Safety Limits -

Four Loops in Operation Catawba Units 1 and 2 2.0-2 mendment Nos.173/165

r SLs 2.0 670 DO NO OPERATE IN THIS AREA. >

660 -

l l

l e - 2

64 -

240 psia 2280 paio b630 -

t I 620 -

l 2100 psia 610 -

1945 pelo l 600 -

590 -

ACCEPT OPERATION l 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 l

Froction of R ed ThermalPower

{< t.- kexImeffd 9YO DFeA Figure 2.1.1-I['pt .

(UNIT 2 ONLY)

Reactor Core Safety Limi -

Four Loops in Operation Catawba Units and 2 2.0-3 Amendment Nos. 173/165 g

r:

SLs 2.0 4

670 DO NOT OPERATE IN THIS AREA

'660 650 640

- 2400 psia LL v 630 c)

$ 2280 psia l-h 620 2100 psla 610 1

600 1945 psia i

590 ACCEPTABLE OPERATION l 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1.1-1 l Reactor Core Safety Limits -

~ Four Loops in Operation Catawba Units'1 and 2 2.0-2 Amendment Nos.

(-

L -- -

V A Afn (4 y RTS Instrum:nt: tion 3.3.1 L Table 3.3.1-1 (page 5 of 7)

Lg Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 4.5% of RTP.

'1 AT 1+r,s

((1 + r, s) ' , s aT. K, - K, II + 8) T - T' + K3 (P - P')- f, (AI)

,1+r,s (1 + r, s) , (1 + r s)

Where: AT is the measured RCS AT by loop narrow range RTDs, 'F.

ATo is the indicated AT at RTP, 'F.

s is the Laplace transform operator, sec".

T,is the measured RCS average temperature, *F.

T is the nominal T vg at RTP (allowed by Safety Analysis), s 585.1*F (Unit 1) s 590.8'F (Unit 2).

P,is the measured pressurizer pressure, psig P is the nominal RCS operating pressure, = 2235 psig K i

= Overtemperature AT reactor trip setpoint, as presented in the COLR, K2 = Overtemperature AT reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K3 = Overtemperature AT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR, ti, t, = Time constants utilized in the lead-lag compensator for AT, as presented in the COLR, t, = Time constant utilized in the lag compensator for AT, as presented in the COLR, t , t. = Time constants utilized in the lead-lag compensator for T.vg, as presented in the COLR, T. = Time constant utilized in the measured T.vs lag compensator, as presented in the COLR, and fi(A!) = a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for qi - go between the " positive" and " negative" f 3(AI). breakpoints as presented in the COLR; fi(AI) = 0, where qi and qo are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and gi + go is total THERMAL POWER in percent

,. of RATED THERMAL POWER; (ii) for each percent Al that the magnitude of qi- go is more negative than the f i(AI) " negative" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the fi(AI) " negative" l slope presented in the COLR; and (continued) l Catawba Units 1 and 2 3.3.1-18 Amendment Nos. 173/165 1 i

I l'

l

RCS Pressurs, Temp:raturo, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling

. (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.

APPLICABILITY: MODE 1.

.....N OT E--- - ---

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or b .' THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure or A.1 Restore DNB parameter (s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average to within limit.

temperature DNB parameters not within limits.

. RCS otal flow rate the B Reduce t Power Ran 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months re on of restrict Neutro lux - High T '

eration of F ure Setp nt below the minal 3.4.1 -1. se oint by the s e ount as th ower limitation in igure 3.4.1-1.

/

(cn ras n ~ mk &

I Catawba Units 1 and 2 3.4.1-1 Amendment Nos.' 7

l Insert A B. RCS total flow rate B.1 Reduce THERMAL POWER 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

< 390,000 gpm but to < 98% RTP.

> 386,100 gpm.

AND B.2 Reduce the Power 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Range Neutron Flux - i High Trip Setpoint below the nominal setpoint by 2% RTP.

(continued) 1

)

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p ,

l' RCS Prrssura, Temp:raturo, and Flow DNB Limits 3.4.1 I

ACTIONS (continued) 4 CONDITION REQUIRED ACTION COMPLETION TIME C; RCS total flow rate &4he C.1 ~ Restore RCS total flow rate 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

.taguxwf givinviiud to ?"* the regien of nnarotien Of Oggia rnntrir tod Oper;;lgg,

.3.4rM-- ), 44, Io o ya -

< %s, to o y* - 08 C.2.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months I POWER to < 50% RTP. f j

AND ~

C.2.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux - High Trip Setpoint to s 55% RTP.

i AND l Reshrt. b C.2.3 Vertfy ROS total flow rate 4e- 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

"'ithin 'hc "cglGa Of ,

-pe""!:ct!c cr rectr:cted np~2 tion. '/u

>s 366, /o O 9P""*

D. Required Action and D.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

l

/

Catawba Units 1 and 2 3.4.1-2 Amendment NosI-L

=

p RCS Pr:ssura, Tempsrature, and Flow DNB Limits 3.4.1 ACTIONS (continued) /*- A=

SURVEILLANCE REQUIREMENTS r/ o SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.2 Verify RCS average temperature is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.3 Verify RCS total flow rate is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.4 Perform CHANNEL CAllBRATION for each RCS total 18 months flow indicator.

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Catawba Units 1 and 2 3.4.1-3 Amendment Nos. 173/165

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)

RCS DNB Parameters No. OPERABLE PARAMETER INDICATION CHANNELS LIMITS

1. Indicated RCS Average meter 4 s 592.0 T ft7 2 *F Temperature - Af I meter 3 s 500.0 ^F ras .9 'F computer 4 s v70.0 T TS7 7 *F computer 3 s 500.0 T Tl7.T~*t*

3- >

2. Indicated Pressurizer meter - 4 2 2227.0 pc!g 2119.S /flg Pressure meter 3 2 9?0.0 pe!; ptz.,e pog computer 4 2 2222.0p;lgIter.tff) computer 3 2 000+0psig t,z s7.y p5:g

3. - RCS Total Flow Rate F!;=c 0.4.1 l g >/19o,8*o pm

\

LA.S ACf % us.b 4- is f92.9 F 7 A v a + 2- *b 3 6 r9t,e 'F c.g ev 4 6 T91. * *F

> s y,, . .e l

Catawba Units 1 and 2 3.4.1 -4 Amendment Nos.

RCS Pressurs, Tcmparatura, cnd Flow NB Limits 3.4.1 POWER DlS BUTION LIMITS l

385,820 A pena of 0.1% for undetected ieedwator .

venturlfo ' g and a rnsasurement uncertainty j of 2.1% for w are hcluded in this figure.

)

382,000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

378,180 Restricted Operation 374,360 * ' '

l lj (92, 370,540) Rohbited g 370,540 operaton Region u

8 (90, 366,720) t 366,720 3

cc 362,900 1

359,080 86 88 90 92 94 96 98 100 102 P rcent of Rated Thermal Power Figure 3.4.1-1 )

(UNIT 1 ONLY) l RCS TOT FLOW RATE VERSUS RATED THERMAL POWER - FOUR LOOPS IN i OPERATION I

Catawba its 1 and 2 3.4.1 -5 Amendment Nos.173/165 I

l

m

\

RCS Pr:ssurs, Temp:ratura, and Flow DNB Li its

.4.1 OWER DISTRIBUTION LIMITS

'388,

. A patally of 0.1% for undetected feedw ater ras &

venturl f ouling and a measurement uncertainty ,,,

of 2.1% for flow are included in this figure. .

385,000 - - - - - - - - - - - - - - - . - - - - - - - - - - - - - -

/

E (96, 381,150) g 381,150 Restricted 3 (94, 377,300)

{ 377,300 I

h (92, 373,450) Prohlt*ed operaton y 373,450 e Region b

o B (90, 369,600) ti 369,600 8

m 365,750 361,900 '

86 88 90 92 9 96 98 100 102 l

Percent of Rated T tmal Power l

Figure 3.4.1-1

. (UNIT 2 ONLY)

RC OTAL FLOW RATE VERSUS RATED THERMAL POWER - F R LOOPS IN OPERATION ;

1 tawba Units 1 and 2 3.4.1 -6 Amendment os.173/165

Reactor Cora SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs) /Js. "y 7

8 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that DNB will not occur and by requiring that fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding, as well as possible cladding pen' oration, that would result in the release of fission products to the reactor coolant. Overheating of the fuelis prevented by maintaining the transient peak linear heat rate (LHR) below the level at which fuel centerline melting occurs. 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.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuelis high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam !

film, high cladding temperatures are reached, and a cladding water !

(zirconium water) reaction may take place. This chemical reaction results I in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.

I The proper functioning of the Reactor Protection System (RPS) and steam generator safety valves prevents violation of ' 7 actor core SLs. j Catawba Units 1 and 2 B 2.1.1-1 Revision No. O

R: actor Core SLs B 2.1.1 BASES 0

APPUCABLE The fuel cladding must not sustain damage as a result of normal SAFE FY ANALYSES operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b. The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling-ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the following functions:

a. High pressurizer pressure trip;

b. Low pressurizer pressure trip;

c. Overtemperature AT trip;

d. Overpower AT trip;

e. Power Range Neutron Flux trip; and

f. Steam generator safety valves.

The limitation that the average enthalpy in the hot leg be less than or equal to the enthalpy of saturated liquid also ensures that the AT measured by instrumentation, used in the RPS design as a measure of core power, is proportional to core power. ,

l The SLs represent a design requirement for establishing the RPS trip l setpoints identified previously. LCO 3.4.1, "RCS Pressure, Temperature, '

and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

Catawba Units 1 and 2 B 2.1.1-2 Revision No. O

R:: actor Core SLs B 2.1.1 :

$] o Ow- t/

i SAFETY LIMITS The curves provided in Figure B 2.1.1-1 show the loci of points of AT( F),

RCS Pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.

- The curves in Figure 2.1.1-1 are based on a reference nuclear enthalpy rise hot channel factor (Fm), a reference axial power shape (Fz, x/L), the approved CHF correlation and the Technical Specification minimum flow ra's Therefore, these curves provide limits for which the analyses analyzed at the above reference values will be bounded. The curves in Figure B 2.1.1-1 illustrate the various RPS functions that are designed to prevent the unit from reaching the limit.

The SL is higher than the limit calculated when the AFD is within the limits of the Fi(AI) function of the Overtemperature AT reactor trip. When the AFD is not within the tolerance, the AFD effect on the Overtemperature AT reactor trips will reduce the setpoints to provide protection consistent with the reactor core SLs (Ref. 3).

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, " Reactor Trip System (RTS) Instrumentation." In MODES 3,4,5, and 6, Applicability is not required since the reactor is not i generating significant THERMAL POWER.

SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 places VIOLATIONS the unit in a MODE in which this SL is not applicable.

l The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

I REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. UFSAR, Section 7.2.

3. DPC-NE-2011 PA, March 1990.

Catawba Units 1 and 2 B 2.1.1-3 Revision No. O

o R=ctor Cors SLs B 2.1.1

s s s

s s s s s s

' s s . s K ,',

. . s s

s s ~- ; Psi Low Pnumme s , s ,

Ramamme Tsip s s s .

s i s s s IGeh Puunnus s s s Rasmus " Dip s i p.m ,,1 s

Opsames/ '

[

s i

E'"'

N ('g s

= g s

. . s ./ N ssg 4

y s

s s

/ s s s %

s ,

s s s s % %

s, s % %

s s s s s s ' s s s s s %

s s %

s s

s s *

.. s ,

i Saam Gescreant

', S\ .

Saferv Valves Open .

i s

s

\ , s s .s r ., c o

- org "0RE L*MITI i

1 i

Figure B 2.1.1-1 l

lllustration of Overtemperature

~

and Overpower AT Protection Catawba Units 1 and 2 B 2,1.1-4 Revision No. O

y- y i

RCS Prrssura SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)- M -

~B 2.1.2 - Reactor Coolant' System (RCS) Pressure SL

~ BASES' BACKGROUND' The SL on RCS pressure protects the integrity of the RCS'against l overpressurization. In the event of fuel cladding failure, fission products

.'are released into the reactor coolant. The RCS tnen serves as the

- primary barrier in preventing the release of fission products into the

- atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and ,

GDC 15, " Reactor Coolant System Design" (Ref.1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded '

during normal operation and anticipated operational occurrences (AOOs).

Also, in accordance with GDC 28, " Reactivity Limits" (Ref.1), reactivity accidents, including rod ejection, do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psla. During normal operation and AOOs, RCS pressure is limited from exceeding the_ design pressure by more than 10%, in accordance with Section lli of the ASME Code (Ref. 2).' To ensure system integrity, all RCS components are

-hydrostatically tested at 125% of design pressure, according to the ASME Code requirements prior to initial operation when there is no fuel in the core. Following inception of unit operation, RCS components shall be

. pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB. If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100,

" Reactor Site Criteria" (Ref. 4).

APPLICABLE - The RCS pressurizer safety valves, the main steam safety valves SAFETY ANALYSES (MSSVs), and the reactor high pressure trip have settings established to ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, as specified in -

Section ill of the ASME Code for Nuclear Power Plant Components Catawba Units 1'and 2 B 2.1.2-1 Revision No. O I

i

% l

f-RCS Pr:ssura SL B 2.1.2 I

BASES b APPLICABLE SAFETY ANALYSES (continued)

(Ref. 2), for anticipated operational occurrences. During the transient, no control actions are assumed, except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.

The Reactor Trip System setpoints (Ref. 5), together with the settings of the MSSVs, provide pressure protection for normal operation and AOOs.

The reactor high pressure trip setpoint is specifically set to provide protection against overpressurization (Ref. 5). The safety analyses for both the high pressure trip and the RCS pressurizer safety valves are performed using conservative assumptions relative to pressure control devices.

More specifically, no credit is taken for operation of the following:

a. Pressurizer power operated relief valves (PORVs);

b. Steam Generator (SG) PORVs;

c. Steam Dump System;

d. Rod Control System;

e. Pressurizer Level Control System; or

f. Pressurizer spray valves.

SAFETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code, Section lil, is 110% of design pressure. The maximum transient pressure allowed in the RCS piping, valves, and fittings under ASME Code Section ill (Ref. 2) is 120% of design pressure.

The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1,2,3,4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized. .

Catawba Units 1 and 2 B 2.1.2-2 Revision No. O

I RCS Pressure SL B 2.1.2 BASES Mo SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in VIOLATIONS MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100,

" Reactor Site Criteria," limits (Ref. 4).

The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

If the RCS pressure SL is exceeded in MODE 3,4, or 5, RCS pressure rnust be restored to within the SL value within 5 mir.utes. Exceeding the RCS pressure SL in MODE 3,4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. The action does not require reducing MODES, since this would require reducir.g temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28. ;

2. ASME, Boiler and Pressure Vessel Code, Section Ill, j 1971 Edition, Winter 1971 Addenda. l

3. ASME, Boiler and Pressure Vessel Code,Section XI, j Article IWB-5000,

4. 10 CFR 100.

5. UFSAR, Section 7.2.

Catawba Units 1 and 2 8 2.1.2-3 Revision No. O

RTS instrumentation B 3.3.1 BASES' /./*

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) overpressure protection systems provide overpressure protection when below MODE 4.

9. Pressurizer Water Level-Hioh The Pressurizer Water Level-High trip Function provides a b' ackup i signal for the Pressurizer Pressure-High trip and also provides protection against water relief through the pressurizer safety valves._ These valves are designed to pass steam in order to achieve their design energy removal rate. A reactor trip is actuated '

prior to the pressurizer becoming water solid. The setpoints are based on percent of instrument span. The LCO requires three channsis of Pressurizer Water Level-High to be OPERABLE. The 1' pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channelis not required to address control / protection interaction concerns. The level channels do not actuate the safety valves, and the high pressure reactor trip is set ;

c below the safety valve setting. Therefore, with the slow rate of

{

charging available,' pressure overshoot due to level channel failure cannot cause the valve to lift before reactor high pressure trip.

In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 Interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be, slow and the operator will have sufficient time to evaluate unit conditions and take coirective actions.

10. : Reactor Coolant Flow-Low

a. Reactor Coolant Flow-Low (Sinale Looo)

The Reactor Coolant Flow-Low (Single Loop) trip Function ;

ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while i avoiding reactor trips due to normal variations in loop flow. l Above the P-8 setpoint, which is approximately 48% RTP, a

, loss of flow in any RCS loop will actuate a reactor trip. The -

Ng

' Catawba Units 1 and 2 B 3.3.1 Revision No. O

L ms

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) g setpoints are based on a minimum measured flow of 95,5T1 gpm(Und 1) OS,250 ypm (unn z). Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 10.b) because of the lower power level and the greater margin to the design limit DNBR.

b. Reactor Coolant Flow-Low (Two Loops)

The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violsting the DNBR limit due to low flow ir, two or more RCS loops while avoiding reactor trips due to normal variations in loop flow, j Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The g fbo setpoints are based on a minimum measured flow of.00,000 gprrHUr" 1) 99 pan gnm (Ur4 2k Each loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE.

l in MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.

Catawba Units 1 and 2 B 3.3.1-18 Revision No.k

RCS Pr:ssura, Temper:tura, and Flow DNB Limits B 3.4.1 1 i

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within :imits assumed in the safety analyses.

The safety analyses (Ref.1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the !

normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed.

The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will cause the reactor core to approach DNB limits.

The RCS coolant average temperature limit is consistent with full power i operation within the nominal operational envelope. Indications of temperature are averaged to determine a value for. comparison to the limit.

A higher average temperature will cause the core to approach DNB limits.

The RCS volumetric flow rate normally rernains constant during an operational fuel cycle with all pumps running. Flow rate Indications are wag Le fl%Lil averaged within a loop and then summed among the four loops to come MW prev /de/ up with a value for comparison to the limit.%CS flow rate end mER"^ L: 3 7gpg pg ayer "'?y be "tredM nfF Tataet ene encthe'; ne ehnwn in Cgre M $. I g 4, / 4fto ensure that the calculated DNBR will not be below the desian DNBR 4

t

) value gwer RCS flow will cause the core to approach DNB limits.

Operation outside these DNB limits increases the likelihood of a fuel cladding failure in a DNB limited event.

- l l

APPLICABLE The requirements of this LCO represent the initial condition for transients '

SAFETY ANALYSES analyzed in the plant safety analyses (Ref.1). The safety analyses have shown that transients initiated from the limits of this LCO will result in I meeting the acceptance criteria, including the DNBR criterion. This is the acceptance limit for the RCS DNB parameters. Changes to the unit that could impact these parameters must be assessed for their impact on the acceptance criteria. A key assumption for the analysis of these events is ;

that the core power distribution is within the limits of LCO 3.1.6, " Control i Bank Insertion Limits"; LCO 3.2.3, " AXIAL FLUX DIFFERENCE (AFD);"

and LCO 3.2.4, " QUADRANT POWER TILT RATIO (OPTR)."

Catawba Units 1 and 2 B 3.4.1-1 Revision No.

RCS Prcssure, Temp:trature, and Flow DNB Limits I B 3.4.1 BASES

~

APPLICABLE SAFETY ANALYSES (continued) 4 ,.

The pressurizer pressure limits and the RCS averagefemperature limits correspond to analytical limits of-2d $5.3 pc6 and 24.u^F used in the safety analyses, with allowance for measurement uncertainty.

The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses.

Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.

RCS total flow rate contains a measurement error of 2.1% based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicatort snd becomes the set of elbow tap coefficients used for

. RCS flow measurereent. Potential fouling of the feedwater venturi. whicn might not have beca detected, could have biased the result from these past precision Iwat balances in a nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi raises the l nominal flow measurement allowance to 2.2% for no fouling f (avarmee 4 The LCO numerical values in Table 3.4.1-1 for pressure and4e# rote are j given for the measurement location with adjustments for the indication instruments.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned Icss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

Catawba Units 1 and 2 B 3.4.1-2 Revision No. %

RCS Pr:ssura, Temp;ratura, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)

Another set of limits on DNB related parameters is provided in SL 2.1.1,

" Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

ACTIONS .A_d Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these pe.rameters not within LCO limits, action must be taken to restore parameter (s).

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience. 7HeAM4 c. Powest. wif. Le u neL win t%ww. %e c. -

' o f 2. k e mer A cour/s%# pie.h wS m.e.

B.1jal B.g rpm Ac.t.h A.i.

eoo RCS total flow rate is not a controllable parameter and is not expected to

( MO,bd W* v 7/)f t, lo o $t" ,h;ary =during steady

+he roninn state operation.

ni rer+rirtos nnarenc e qIfr;gur; the indicated 0.4.14, thenRCS total flow THERMAL j rate i k POWER may not exceed" " limit che"' 2- '50 !!gurfin addition, the e

gp Power Range Neutron Flux - High Trip Setpoint must be reduced from the ominal setpoint by?n ?meu-t 0;u2! t0 '50 NEP"^.L OWER Hmi' ' rem N eh";er '- Figurc 3.4.14 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . "Or examp!:,if T!;ERMAL 2.#/o R. N - PO?!EG ;5 linided42 M*i. +"ea 'he +r!p retpo!r' "'uct be reduccd by de di"erence hetwgon the limit and RTP which je-4b The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

C.1. C.2.1. C.2.2. and C.2.3 g foo if the indicated RCS total flow rate isF"" '"e re;!:n of prehlb;ted

.cpc ation ' , I:;urc 3.4.1 1, thend!50 '"e c^mhinaHed RCS total flow and NA"^L POWEftmust be restored to*: !"t chov.a in the ? wwe f^r rer+ricted operation within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months cr power must be reduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1. If THERMAL POWER is reduced to less than 50%

/ W '00 D i

Catawba Units 1 and 2 B 3.4.1-3 Revision No.K I

RCS Pr:ssura, Temp:ratura, tnd Flow DNB Limits B 3.4.1 BASES ACTIONS (continued)

If 8'4ON M RTP, the Power Range Neutron Flux - High Trip Setpoint must also be W M *a I'*- reduced tp 5 55% RTP. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpointecrogn!:0; that, . "5 per/Or reducod, the 0;fety enalysis mu,T,pt.e..: - ent!rFed anri thnra le ne urger' n00d to reduc 0 the trlp 0tpe:nio. This is a sensitive operation that may inadvertently trip the Reactor Protection System. Operation is permitted to continue provided t'eder*ed b the RCS total flow is ver'd 'n ha wi+hia +he reg!cn of per~!00!b!e er

), ) ft, t 0 o 7 rae'ric+ed ep0:0FeWithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.

D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of

'12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to tha operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are avalfable to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS. The j 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to i I

regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

Catawba Units 1 and 2 8 3.4.1-4 Revision No.

j

RCS Pr:ssura, TemperAtura, end Flow DNB Limits B 3.4.1 I

BASES No [

b

, SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.1.3 This surveillance demonstrates that the RCS total flow rate remains within the required limits. Aiarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS flow rate and related equipment status (e.g. RCP voltage and frequency, etc.). The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess potential degradation and to verify operation within safety analysis assumptions.

SR 3.4.1.4 Calibration of the installed RCS flow instrumentation permits verification that the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.

The Frequency of 18 months is consistent with operating experience.

REFERENCES 1. UFSAR, Section 15.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4 Catawba Units 1 and 2 B 3.4.1-5 Revision No. O

Attachment 2a !

Proposed New McGuire Technical Specifications i

l l

SLs 2.0 670 DO NOT OPERATE IN THIS AREA 660 650 640 g- 2400 psia

'v 630 m 2280 psia I-

$ 620 2100 psia 610 3

600 1945 psia 590 ACCEPTABLE OPERATION ;

580 i 0.0 . 0.2 0.4 0.6 0.8 1.0 1.2 Fraction Of Rated Thermal Power i

Figure 2.1.1-1 j Reactor Core Safety Limits - l Four Loops in Operation ;

McGuire Units 1 and 2 2.0-2 Amendment Nos.

h

L RTS Instrum3ntation 3.3.1 f

Table 3.3.1-1 (page 2 of 7) f Reactor Trip System Instrumentation l

APPLICABLE MODES OR OTHER~

SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

6. Overtemperature AT 1,2 4 E SR 3.3.1.1 Refer to Refar to SR 3.3.1.3 Note 1 (Page Note 1 SR 3.3.1.6 3.3.1 18) (Page SR 3.3.1.7 3.3.1-18)

SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.17

7. Overpower AT 1.2 4 E SR 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 2 (Page Note 2 SR 3.3.1.6 3.3.1-19) (Page SR 3.3.1.7 3.3.1-19)

SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.17

8. Pressurtzer Pressure

a. Low 1(f) 4 M SR 3.3.1.1 31935 psig 31945 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. High 1,2 4 E SR 3.3.1.1 5 2395 psig 5 2385 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

9. Pressurizer Water $(f) 3 M SR 3.3.1.1 $ 93% $ 92%

Level- High SR 3.3.1.7 SR 3.3.1.10

10. Reactor Coolant Flow-Low

a. Single Loop 1(9) 3 per loop N SR 3.3.1.1 3 87 % 3%

SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. Two Loops 1(h) 3 per loop M SR 3.3.1.1 2 87 % 188%

SR 3.3.1.7 I SR 3.3.1.10 SR 3.3.1.16

11. Undervoltage RCPs 3(f) i per bus M SR 3.3.1.9 3 5016 V 2 5082 V SR 3.3.1.10 SR 3.3.1.16 (continued)

(f) Above the P-7 (Low Power Reactor Trips Block) interlock.

(g) Above the P-8 (Power Range Neutron Flux) Interlock.

l (h) Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) intertock.

McGuire Units 1 and 2 3.3.1-15 Amendment NoS.

RTS Instrum::nt: tion 3.3.1 Table 3.3.1-1 (page 5 of 7) l Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following Trip Setpoint by more than 4.4% of RTP.

1 AT (1 * 5 ' SAT- * #

T - T' + K, (P - P') * (AI)

(1 +r28)si + r3s, a[K, - K, (1 + r, s) _ (1 + r s)

Where: AT is measured RCS AT by loop narrow range RTDs, *F.

l ATo is the indicated AT at RTP, *F.

l s is the Laplace transform operator, sec-1.

T is the measured RCS average temperature, *F.

T' is the nominal Tavg at RTP,5 585.1 *F.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, = 2235 psig Ki = Overtemperature AT reactor trip setpoint, as presented in the COLR, K2 = Overtemperature AT reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K3 = Overtemperature AT reactor trip depressurization setpoint penalty ,

coefficient, as presented in the COLR, t i ,t2 = Time constants utilized in the lead-lag controller for AT, as presented in the COLR,

= Time constants utilized in the lag compensator for AT, as presented in the t3 l COLR, t 4, is

= Time constants utili: ad in the lead-lag controller for T.,,, as presented in the COLR, t.

= Time constants utilized in the measured T.., lag compensator, as presented in the COLR, and,

= a function of the indicated difference between top and bottom detectors of f i(AI) the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for qi - q, between the " positive" and " negative" f (Al) breakpoints as 3

presented in the COLR; f (AI) = 0, where qi and q, are percent 3

RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + q is total THERMAL POWER in percent of RATED THERMAL POWER; Continued l

McGuire Units 1 and 2 3.3.1-18 Amendment Nos.

F RCS Prtssura, Temper:tura, end Flow DNB Limits 3.4.1 l

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.

APPLICABILITY: MODE 1.

NOTE Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer pressure or A.1 Restore DNB parameter (s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average to within limit. j temperature DNB i parameters not within limits.

I B. RCS total flow rate B.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months '

< 390,000 gpm but POWER to < 98% RTP.

> 386,100 gpm.

AND B.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux- High Trip Setpoint below the nominal setpoint by 2% RTP.

(continued)

McGuire Units 1 and 2 3.4.1-1 Amendment Nos. l I

1 l

RCS Pressura, Temperatura, end Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rate C.1 Restore RCS total flow rate 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

< 386,100 gpm. to 2 386,100 gpm.

O.B C.2.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,

POWER to < 50% RTP. )

i AND C.2.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux - High Trip Setpoint to 5 55% RTP.

AND C.2.3 Restore RCS total flow rate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 2 386,100 gpm.

D. Required Action and D.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

l McGuire Units 1 and 2 34.1-2 Amendment Nos.

F:

RCS Pr;ssure, Tcmper:tura, end Flow DNB Limits 3.4.1 .

I l

)

l Table 3.4.1-1 (page 1 of 1) ;

. RCS DNB Parameters

]

PARAMETER INDICATION No. OPERABLE LIMITS CHANNELS-

'1. ' Indicated RCS Average meter 4 5 587.2 *F Temperature meter 3 5 586.9 'F computer '4 5 587.7 'F computer 3 5 587.5 *F

2. Indicated Pressurizer meter 4 2 2219.8 psig Pressure meter 3 2 2222.1 psig I

computer 4 2 2215.8 psig computer 3 2 2217.5 psig

3. RCS Total Flow Rate 2 390,000 gpm.

l McGuire Units 1 and 2 3.4.1-4 Amendment Nos.

1 1

o RTS Instrumsntition B 3.3.1 BASES-APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure protection when below MODE 4.

9. Pressurizer Water Level-Hiah The Pres'surizer Water Level-High trip Function provides a backup

~s ignal for the Pressurizer Pressure-High trip and also provides

- protection against water relief through the pressurizer safety valves. These valves are designed to pass steam in order to achieve their design energy removal rate. A reactor trip is actuated prior to the pressurizer becoming water solid. The setpoints are based on percent of instrument span. The LCO requires three channels of Pressurizer Water Level-High to be OPERABLE. The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is not required to address control / protection interaction concems. The

- level channels do not actuate the safety valves, and tb high pressure reactor trip is set below the safety valve setting.

Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the _ safety valve to lift before reactor high pressure trip, in MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and take corrective actions.

10. Reactor Coolant Flow-Low

a. Reactor Coolant Flow-Low (Sinole Looo)

The Reactor Coolant Flow-Low (Single Loop) trip Function -

ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow.

Above the P-8 setpoint, which is approximately 48% RTP, a loss of flow in any RCS loop will actuate a reactor trip. The setpoints are based on a minimum measured flow of 97,500 1

McGuire Units 1 and 2 B 3.3.1-16 Revision No.

RTS Instrumsntation B 3.3.1 1 BASES-APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) gpm. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 i below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 10.b) because of the lower power level and the greater margin to the design limit DNBR.

b. Reactor Coolant Flow-Low (Two Loops)

The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the

- DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The setpoints are based on a minimum measured flow of 97,500 gpm.. Each loop has three flow detectors to monitor flow.

The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE. l In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip ,

must be OPERABLE. Below the P-7 setpoint, all reactor j trips on low flow are automatically blocked since power !

distributions that would cause a DNB concern at this low I power level are unlikely. Above the P-7 setpoint, the reactor ,

trip on low flow in two or more RCS loops is automatically I enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR. ,

l l

McGuire Units 1 and 2 8 3.3.1-17 Revision No.

l i

m_..

RCS Prtssura, Temperatura, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

. B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND: These Bases address requirements for maintaining RCS pressure,

- temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref.1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed.

The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will cause the reactor core to approach DNB limits.

The RCS coolant average temperature limit is consistent with full power operation within the nominal operational envelope. Indications of temperature are averaged to determine a value for comparison to the limit. A higher average temperature will cause the core to approach DNB limits.

The RCS volumetric flow rate normally remains constant during an operational fuel cycle with all pumps running. Flow rate indications are averaged within a loop and then summed among the four loops to come up with a value for comparison to the limit. A lower RCS flow will cause the core to approach DNB limits. RCS flow rate may be slightly reduced provided THERMAL POWER is also reduced to ensure that the calculated DNBR will not be below the design DNBR value.

Operation outside these DNB limits increases the likelihood of a fuel cladding failure in a DNB I;mited event.

APPLICABLE - The requirements of this LCO represent the initial conditions for SAFETY ANALYSES transients analyzed in the plant safety analyses (Ref.1). The safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the. acceptance criteria, including the DNBR ;

criterion. This is the acceptance limit for the RCS DNB parameters. )

~ Changes to the unit that could impact these parameters must be j McGuire Units 1 and 2 B 3.4.1-1 Revision No.

RCS Prrssura, Tamperaturo, end Flow DNB Limits B 3.4.1 BASES 1

APPLICABLE SAFETY ANALYSES (continued)'

assesshd for their impact on the acceptance criteria. A key assumption

' for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, " Control Bank Insertion Limits"; LCO 3.2.3,

" AXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, " QUADRANT POWER TILT RATIO (QPTR)."

The pressurizer pressure limits and the RCS average temperature limits -

correspond to analytical limits of 2205 psig and 589.1'F used in the l

safety analyses, with allowance for measurement uncertainty.

The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2);

)

LCO This LCO specif.es limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses. Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.

RCS total flow rate contains a measurement error of 1.7% based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as ,

determined during these heat balances, were averaged for each elbow l tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCC flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner.

Therefore, a penalty of 0.1% for undetected fouling of the feedwater l venturi raises the nominal flow measurement allowance to 1.8% for no i fouling.

The LCO numerical values in Table 3.4.1-1 for pressure and average temperature are given for the measurement location with adjustments for the indication instruments.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.

McGuire Units 1 and 2 B 3.4.1-2 Revision No.

t: m:

p RCS Pressura, Temperatura, and Flow DNB Limits B3A1 BASES _

APPLICABILITY (continued)

A Note has been added to indicate the limit on pressurizer pressure it not applicable during short term operational transients such as a THERMA L l-

, POWER ramp increase > 5% RTP per minute or a THERMAL POWER l step increase > 10% RTP. - These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

Another set of limits on DNB related parameters is provided in SL 2.1.1,

" Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

ACTIONS: &1 Pressurizer pressure and RCS average temperature are controllable and

measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter (s).

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

B.1 and B.2 l

RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is

< 390,000 gpm but > 386,100 gpm, then THERMAL POWER may not exceed 98% RTP. '.HERMAL POWER must be reduced within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The Complet.'on Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1.

In addition, the Power Range Neutron Flux - High Trip Setpoint must be ,

reduced from the nominal setpoint by 2% RTP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The l !

Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and l- there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

L l

I McGuire Units'1 and 2 B 3.4.1-3 Revision No.

e l

RCS Pr;;ssura, Temperatura, and Flow DNB Limits B 3.4.1 BASES ACTIONS -(continued)

C.1. C.2.1. C.2.2. and C.2.3 If the indicated RCS total flow rate is < 386,100 gpm, then RCS total flow must be restored to 2 386,100 gpm within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or power must be reduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1. If THERMAL POWER is reduced to less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint must also be reduced to s; 55% RTP. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection System. Operation is permitted to continue provided the RCS total flow is restored to 2 386,100 gpm within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint. ,

_Dm.i If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

McGuire Units 1 and 2 B 3.4.1-4 Revision No.

l Attachment 2b ;

Proposed New Catawba Technical i

Specifications l

i l

c SLs 2.0 670 DO NOT OPERATE IN THIS AREA 660 650 640 g 2400 psia L 630

~? 2280 psia t-F--

620 0"

2100 psia 610 600 1945 psia i

590 ACCEPTABLE OPERATION 580 0.0 - 0.2 0.4 0.6 0.8 1.0 1.2 e Fraction of Rated Thermal Power Figure 2.1.1 1 Reactor Core Safety Limits Four Loops in Operation Catawba Units 1 and 2 2.0-2 Amendment No.

l l

RCS Pr:ssura, Temp:ratura, and Flow DNB Limits 3.4.1

' 3.4 l REACTOR COOLANT SYSTEM (RCS) 1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits .

LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in

- Table 3.4.1-1.

APPLICABILITY: MODE 1.

NOTE - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ------

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

'A. Pressurizer pressure or A.1 Restore DNB parameter (s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average . to within limit.

temperature DNB parameters not within limits.

B. RCS total flow rate - B.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

< 390,000 gpm but POWER to $; 98% RTP.

2 386,100 gpm.

AND B.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux- High Trip Setpoint below the nominal setpoint by 2% RTP.

(continued)

Catawba Units'I and 2 3.4.1-1 Amendment Nos.

RCS Pr:s:ura, Temp:ratura, and Flow DNB Limits 3.4.1

' ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rate . C.1 ' Restore RCS total flow rate 2 isours

< 386,100 gpm.- to 2 386,100 gpm.

9.8 - q C.2.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to < 50% RTP.

AND C.2.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Ne'r.ron Flux - High Trip St< point to s 55% RTP.

AND C.2.3 Restore RCS total flow rate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 2 386,100 gpm.

D. Required Action and D.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

l I

Catawba Units 1 and 2 3.4.1-2 Amendment Nos. l l

RCS Pr:ssura, Tempt f, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)

RCS DNB Parameters No. OPERABLE PARAMETER INDICATION CHANNELS LIMITS l

1. Indicated RCS Average meter 4 -5 587.2 *F Temperature - Unit 1 meter 3 s 586.9 *F computer - 4 s 587.7 *F computer 3 s 587.5 *F Indicated RCS Average meter 4 5 592.9 *F Temperature - Unit 2 meter 3 s 592.6 *F computer 4 s 593.4 *F computer 3 s 593.2 *F l

2. Indicated Pressurizer meter 4 2 2219.8 psig Pressure meter 3 22222.1 psig computer 4 2 2215.8 psig computer 3 2 2217.5 psig

3. RCS Total Flow Rate 2 390,000 gpm l Catawba Units 1 and 2 3.4.1 -4 Amendment Nos.

l J

RTS Instrument: tion i

B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY '(continued)

, setpoints are based on a minimum measured flow of 97,500 gpm. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 10.b) because of the lower power level and the greater margin to the design limit DNBR.

b. Reactor Coolant Flow-Low (Two Looos)

The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The setpoints are based on a minimum measured flow of 97,500 gpm. Each loop has three flow detectors to monitor flow.

The flow signals are not used for any control system input.

i The LCO requires three Reactor Coolant Flow-Low channels l per loop to be OPERABLE. I In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power. 3 distributions that would cause a DNB concern at this low i power level are unlikely. Above the P-7 setpoint, the reactor i trip on low flow in two or more RCS loops is automatically j enabled. Above the P-8 setpoint, a loss of flow in any one !

loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.

)

- Catawba Units 1 and 2 B 3.3.1-18 Revision No.1

)

i

F RCS Pressura, Temperatura, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

' The safety analyses (Ref.1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure,-

temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed.

The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will '

cause the reactor core to approach DNB limits.

The RCS coolant average temperature limit is consistent with full power operation within the nominal operational envelope. Indications of temperature are averaged to determine a value for comparison to the limit.

A higher average temperature will cause the core to approach DNB limits.

The RCS volumetric flow rate normally remains constant during an operational fuel cycle with all pumps running. Flow rate indications are averaged within a loop and then summed among the four loops to come up with a value for comparison to the limit. A lower RCS flow will cause the core to approach DNB limits. RCS flow rate may be slightly reduced provided THERMAL POWER is also reduced to ensure that the calculated DNBR will not be below the design DNBR value.

- Operation outside these DNB limits increases the likelihood of a fuel ,

cladding failure in a DNB limited event.

APPLICABLE The requirements of this LCO represent the initial condition for transients

' SAFETY ANALYSES analyzed in the plant safety analyses (Ref.1). The safety analyses have i shown that transients initiated from the limits of this LCO will result in l meeting the acceptance criteria, including the DNBR criterion. This is the I acceptance limit for the RCS DNB parameters. Changes to the unit that ;

could impact these parameters must be assessed for their impact on the j acceptance criteria. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, " Control !

Bank Insertion Limits"; LCO 3.2.3, " AXIAL FLUX DIFFERENCE (AFD);" l and LCO 3.2.4, "OUADRANT POWER TILT RATIO (OPTR)."

Catawba Units 1 and 2 B 3.4.1-1 Revision No.1

i; RCS Prcssura, Temp:ratura, and Flow DNB Limits i B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued).

The pressurizer pressure limits and the RCS average temperature limits correspond to analytical limits of 2205 psig and 589.1*F (Unit 1) and 594.8 F (Unit 2) used in the safety analyses, with allowance for measurement uncertainty.

The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO . This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses.

Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.

RCS total flow rate contains a measurement error of 2.1% based on the performance of past precision heat balances and using the result to .

calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating -

RCS flow. This set of coefficients establishes the calibration of the RCS -

flow rate Indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance to 2.2% for no fouling.

The LCO numerical values in Table 3.4.1-1_ for pressure and average temperature are given for the measurement location with adjustments for the indication instruments.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature,'and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern. ;

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term l perturbations where actions to control pressure variations might be !

counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

Catawba Units 1 and 2 B 3.4.1-2 Revision No.1

RCS Pr:ssura, Tcmparatura, and Flow DNB Limits B 3.4.1 BASES _

APPLICABILITY (continued)

Another set of limits on DNB related parameters is provided in SL 2.1.1,

" Reactor Core SLs " Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must

. check whether or not an SL may have been exceeded.

ACTIONS . L1.,

Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter (s).

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based i on plant operating experience.

B.1 and B.2 l RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is

< 390,000 gpm but 2 386,100 gpm, then THERMAL POWER may not exceed 98% RTP. THERMAL POWER must be reduced within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1.

In addition, the Power Range Neutron Flux - High Trip Setpoint must be reduced from the nominal setpoint by 2% RTP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The l l Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

l l

C.1. C.2.1. C.2.2, and C.2.3 If the indicated RCS total flow rate is < 386,100 gpm, then RCS total flow must be restored to 2 386,100 gpm within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or power must be reduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1 if THERMAL POWER is reduced to less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint l Catawba Units 1 and 2 B 3.4.1-3 Revision No.1

V i

RCS Pr:ssura, Temperatura, and Flow DNB Limits B 3.4.1

, BASES ACTIONS (continued) l l

must also be reduced to 5; 55% RTP. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to

! reset the trip setpoints is consistent with Required Action B.2. This is a l l sensitive operation that may inadvertently trip the Reactor Protection System. Operation is permitted to continue provided the RCS total flow is restored to 2 386,100 gpm within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 l hours is reasonable considering the increased margin to DNB at power

! levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.

_D_:.1, if the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

l l

l Catawba Units 1 and 2 B 3.4.1-4 Revision No.1

U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 1 of 19 Description of the Proposed Changes and Technical Justification Proposed Revision to Technical Specification Figure 2.1.1-1 Replace the current Technical Specification (TS) Reactor Core Safety Limits, Figure 2.1.1-1, with the attached figures. As discussed in the Bases of the Technical Specifications, the Reactor Core Safety Limit curves show the loci of points of thermal power, reactor coolant system pressure and average temperature below which the calculated DNBR is not less than the design DNBR value, or the average enthalpy at the vessel exit is less than or equal to the enthalpy of saturated liquid. The Bases also stipulate that these curves are based upon the TS minimum Reactor Coolant System flow rate. Changes to these figures for all four McGuire and Catawba units are necessary due to the increase in the minimum RCS total flow rate limit to 390,000 gpm.

The current Reactor Core Safety Limits figure for McGuire Units 1 and 2 was revised in March 1994 by License Amendment Nos. 141 and 123, respectively. The corresponding figure for Catawba Unit 1 was revised in December 1993 by License Amendment No. 113. These amendments were requested as a result of the continued plugging and sleeving of steam generator tubes and the resulting reduction in primary system flow. In addition, a hot leg temperature streaming phenomenon had affected the ability to accurately measure flow. The Reactor Core Safety Limits curves for each of the three units were revised to coincide with a reduction in the minimum RCS total flow rate limit from 385,000 gpm to 382,000 gpm. Catawba Unit 2, which has not been burdened by significant steam generator tube plugging, currently has a minimum RCS total flow rate limit of 385,000 gpm. The current Reactor Core Safety Limits figure for Catawba Unit 2 was revised in March 1993 by License Amendment No. 101.

The Reactor Core Safety Limits figure does not affect the normal operation of the facility. These curve.e are only used to determine the need for further safety evaluations following postulated overheating or power excurs..on transients. Violation of these limits would result in a unir shutdown, with unit restart prohibited until authcrized by the Commission. The Reactor Protection System trip sctpoints have been selected to ensure that the Reactor Core Safety Limits are not exceeded.

E 1 U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 2 of 19 Proposed Revision to Technical Specification Table 3.3.1-1 (MNS only)

The low RCS flow reactor trip setpoint is changed from 91% to 88%

of the loop minimum flow rate limit and the corresponding allowable value is changed from 90 to 87%. The setpoint reductica is intended to preclude spurious reactor trips that might cccur following the increase in the minimum RCS total flow rate limit due to normal flow noise and the lower flow indication historically observed in Loop A at McGuire Unit 1. This Reactor i Protection System trip functions in the mitigation of the partial i loss of forced reactor coolant flow and the reactor coolant pump shaft seizure (locked rotor) accidents.

T', the nominal T-average at Rated Thermal Power (RTP), is changed.from < 585.1 F to 5 585.1 F. This is a correction of a typographical error that was incorporated in the Improved l Technical Specifications (ITS) in November, 1998. Pre-ITS Amendments 175/157 had T' listed as $ 585.1 F.

Proposed Revision to Technical Specification Figure 3.4.1-1, Table 3.4.1-1, 3.4.1 CONDITIONS B and C, and 3.3.1 and 3.4.1 Bases !

l It is proposed to increase the minimum RCS total flow rate limit l to 390,000 gpm from the current values of 382,000 gpm for both the'McGuire units and Catawba Unit 1 and 385,000 gpm for Catawba Unit 2. The proposed 390,000 gpm is then relocated from Figure 3.4.1-1 to Table 3.4.1-1. The current power / flow tradeoff stairstep is revised to allow operation with a 2% core power reduction in the event that the measured RCS total flow rate falls between the 390,000 gpm limit and 386,100 gpm (a 1% flow deficit). If the measured total flow rate is found to be less than 386,100 gpm, existing TS Condition C would be entered. The proposed power / flow tradeoff is relocated from Figure 3.4.1-1 to CONDITIONS B and C of 3.4.1. The note on Figure 3.4.1-1 regarding penalty for undetected feedwater venturi fouling and measurement uncertainty for flow has been included in Bases 3.4.1. Figure 3.4.1-1 is deleted since its content has been relocated as described above.

It is proposed to revise the RCS average temperature and pressurizer pressure limits in Table 3.4.1-1 consistent with the temperature and pressure assumptions made in the reanalyzed UFSAR Chapter 15 DNB transients. This change is being made in order to guarantee that the plant is operated within the limits assumed in the safety analyses, and, thereby, ensure that the analyses remain valid. Bases 3.3.1 is revised to reflect the total

U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 3 of 19 390,000 gpm limit (97,500 gpm per loop). Bases 3.4.1 is revised to reflect the proposed changes to Figure 3.4.1-1, Table 3.4.1-1, 3.4.1 CONDITIONS B and C, and pressurizer pressure and RCS average temperature analytical limits used in the reanalyzed DNB safety analyses. An editorial change is made in Bases 3.4.1, page B 3.4.1-2, to reflect that the numerical values in Table 3.4.1-l'for pressure and average temperature, not pressure and flow rate, are given for the measurement location with adjustments for the indication instruments.

The RCS total flow rate limit figure for McGuire Units 1 and 2 was last revised in March 1994 by License Amendment Nos. 141 and 123, respectively. The corresponding figures for Catawba Units 1 and 2 were revised by License Amendment Nos. 113 and 101 in December 1993 and March 1993, respectively. The current criteria for minimum RCS total flow rate limit have been arrived at through a series of successive reductions in the original TS values. These changes were required to accommodate the hot leg i streaming phenomenon and a significant increase in steam generator tube plugging at the McGuire units and Catawba Unit 1.

The steam generator tube plugging at Catawba Unit 2 has been very j small (approximately 1.1% as of October 1998); therefore, the j measured flow decreases since Catawba 2 startup were primarily l the result of the hot leg streaming phenomenon and its impact on the calorimetric calculation used to determine the RCS flow.

Miniuum Measured RCS Flow Background l The Technical Specification minimum RCi total flow rate limit provides core protection and ensures that the flow is maintained 1 within the normal steady-state envelope of operation assumed in the accident analyses. The revised flow limit is consistent with the Updated Final Safety Analysis Report (UFSAR) Chapter 6 & 15 transient analyses. Duke Energy maintains a plan for each future reload design which phases in any design and technology improvements that are needed to support desired cycle lengths, minimize cost, and balance the impact on plant operating margins.

In core reload design, longer cycle lengths, smaller feed batch sizes, and the use of features such as axial blankets help to reduce costs. However, most of these innovations impact the plant in terms of higher peaking factors and reduced operating margin. Operating margins will get even tighter if some of the available flow margin is not used. The excess Reactor Coolant System flow margin can be used to improve the UFSAR Chapter 15 analysis results and allow more operating margin to accommodate future reload designs.

Past changes in the minimum RCS total flow rate limit Technical Specification have been reductions due to SG tube plugging and

r U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 4 of 19 the effects of hot leg streaming on the calorimetric calculation useduto determine the minimum measured flow. These reductions alleviated the decreasing flow margin but resulted in tighter core design and operating margin.

To eliminate the impact of hot leg streaming on the.RCS flow measurement, the elbow tap method of measuring flow was adopted.

In the elbow tap method, the flow rate is measured based on the pressure drop in the elbow tap, elbow tap flow coefficient, and cold leg density. The TS flow rate limit is applicable in Mode 1 only. Approval of the elbow tap method in Amendment Nos. 133 and 135 for McGuire Units 1 and 2 and Amendment Nos. 128 and 122 for q Catawba Units 1 and 2 allowed some immediate flow margin gain and provided assurance that future flow decreases would be predictable with regard to future plant changes such as steam generator tube plugging. Also, as a result of the steam generator replacement at McGuire Units 1 and 2 and Catawba Unit 1, significant increases in RCS flow were realized. This is i primarily due to eliminating the flow penalty due to tube plugging in the old steam generators. Currently, the lowest total RCS flow rate measured among the four units is in excess of 394,000 gpm. Following the increase in the Technical Specification minimum RCS total flow rate limit to 390,000 gpm, there will still be greater than 4,000 gpm of flow margin available to accommodate future steam generator tube plugging, etc. The level of steam generator tube plugging necessary to l reduce the RCS flow by this amount has been conservatively j estimated to be approximately 6%, which should provide many i cycles of plant operation before the increased TS minimum RCS ,

total flow rate limit will be approached. l I

All four McGuire and Catawba units are scheduled to undergo a transition from Framatome Mark-BW fuel assemblies to Westinghouse j RFA fuel beginning with Catawba Unit 2 Cycle 11. The new j Westinghouse fuel has a slightly higher design pressure drop than the current Framatome fuel. Once a full core loading of Westinghouse fuel has been achieved, a resultant decrease in the RCS flow rate of approximately 0.6% is expected. The proposed revision to increase the minimum RCS total flow rate limit will be credited to offset the transition core effects.

Power / Flow Trade-Off Background During initial operation of McGuire Unit 2, the measured RCS flow had inadequate margin to the flow rate assumed in the accident safety analyses. On November 18, 1983, a request to revise the McGuire Technical Specifications was submitted to the NRC to lower the RCS flow requirement by 2% for Unit 2, and to implement a 1% power / flow tradeoff stairstep in TS Figure 3.2-3b (which

e i

U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 5 of 19 corresponds to the current Figure 3.4.1-1). The existing TS figure.for both units at that time required a reduction to 90%

power if the' measured flow rate was lese than the TS flow limit, as long'as the measured flow rate was grester than 95% of the TS flow limit. The intent of the revision request was to allow operation at 99% power, rather than 90% power, if a small deficit in.RCS flow was measured. Westinghouse developed the technical basis for the power / flow tradeoff stairstep and performed the safety evaluation for the TS submittal. The relationships between power, flow, and DNBR at full power were used to conservatively establish a 1% reduction in power for each 1%

reduction in flow below the full power TS flow limit. No transient analyses were performed at the reduced flow / reduced power initial condition. For additional conservatism, this tradeoff was increased to 2% power per 1% flow in the final TS revision package.

On February 4, 1984, Amendment Nos. 28 and 9 were issued for McGuire Units 1 & 2 by the NRC, which was the first implementation of the 2% power per 1% flow tradeoff stairstep.

Subsequently, in November 1987, Amendment Nos. 34 and 25 extended this same approach to Catawba Units 1 & 2. The Catawba safety evaluation, which was performed by Duke Power, generally followed the Westinghouse evaluation. The Duke safety evaluation for Catawba also required that the high flux reactor trip setpoint be reduced by an amount equivalent to the power reduction required i by the stairstep. This revision was intended to retain the J margin between the transient analysis initial conditions and the high flux trip setpoint, thereby ensuring the validity of the related safety analyses. This additional requirement was )

included in the McGuire Technical Specifications beginning with

{

the McGuire 1 Cycle 8 reload package. j Recent engineering evaluations raised concerns regarding the I validity of the power / flow stairstep. In particular, for events not relying upon the high flux or delta-T reactor trips, and those initiating at reduced power, there are no analyses in place to show that a flow reduction of up to 5% was acceptable for all events. For stairstep safety evaluations similar to but more l recent than the McGuire evaluation in 1983, Westinghouse had addressed the flow reduction of up to 5% allowed by the stairstep. The 5% flow reduction was addressed by either assigning DNB margin or explaining why the existing analyses remained valid. It was determined that the original safety analyses had not considered all possible initiating conditions for design basis transients. Administrative directions were put in place prohibiting use of the power / flow tradeoff until the TS could be changed to eliminate the resulting potential non-conservatisms. The proposed TS changes remedies this situation,

f U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 6 of 19 which was first reported to the NRC in LERs 369/97-10 and 413/97-007.

In January of 1994, the indicated RCS flow rate at Catawba Unit 1 decreased to less than the TS limit, and power was reduced to 98%

per the TS power / flow tradeoff. In March of 1994, the elbow tap method of RCS flow measurement was adopted, which solved the problem of hot leg streaming affecting the flow indication. The de-rating of Catawba Unit 1 was the only period of operation in which either McGuire or Catawba Units used the power / flow tradeoff. It is noted that the actual flow rate values were found to have remained above the TS required flow during this de-rated period.

Impact of Technical Specification Changes on UFSAR Analyses The current TS minimum RCS total flow rate limits are 382,000 gpm at both McGuire units and Catawba Unit 1 and 385,000 gpm at Catawba Unit 2. The current UFSAR Chapter 6 and 15 analysis assumptions are bounded by a minimum RCS total flow rate limit of 382,000 gpm and a maximum flow rate limit of 420,000 gpm.

Therefore, the revised TS minimum RCS total flow rate limit of 390,000 gpm will continue to be within the range of bounding flow assumptions used in these accident analyses. Those transients for which the maximum RCS total flow rate limit is conservative will not be impacted by a change in the TS minimum RCS total flow rate limit. The effect of increasing the minimum RCS total flow rate limit will primarily be to narrow the range between the minimum and maximum flows assumed in the accident analyses.

Since the current measured RCS flow rate is below the maximum flow rate assumed in the accident analyses and future changes in the measured RCS flow will likely be in the downward direction, the narrowed range is not a significant concern. Since the intent of increasing the TS minimum RCS total flow rate limit is to provide more margin in the core design limits, a survey of the '

pertinent UFSAR accident analyses was performed to determine which transients provide limiting core design limits for each cycle.

Below is the justification for why an increase in the Technical Specification minimum RCS total flow rate limit and a revised power / flow tradeoff will not have any adverse impact on any of the accident analyses presented in Chapter 15 of the UFSAR. In )

addition, none of the analyses presented in Chapters 3, 4, or 6 is significantly affected by this change. Therefore, the proposed change in the TS minimum RCS total flow rate limit will i not result in a reduction of safety margin for these analyses. j

F L U.S. Nuclear Regulatory Comunission Attachment-3 f June 24, 1999 Page 7 of 19 l

LOCA Blowdown Reactor Vessel and Loop Forces (UFSAR Sections

! 3.6.4.1 and 3.9.1.5)

.The primary factors which affect the blowdown forces resulting from a LOCA are RCS' pressure, vessel inlet and outlet fluid i

temperatures, and to a lesser degree,.the loop and vessel flowrates. The proposed flow limit is within the upper and lower bounds supported by the existing LOCA blowdown forces analyses.

Therefore, the increase in minimum RCS total flow rate limit would have no adverse effect on vessel internals, core components, and coolant loop piping structural adequacy.

Thermal Hydraulic Design (UFSAR Section 4.4)

The thermal hydraulic design for the McGuire and Catawba units was evaluated for the increase in minimum RCS total flow rate

, limit to 390,000 gpm. The increased flow rate limit resulted in a slight increase in the DNB Safety Limit lines, therefore the Reactor Core Safety Limit figure (Technical Specification Figure 2.1.1-1) will be revised. The axial flux difference limits (Technical Specification 3.2.3) are unchanged, and all of the current thermal hydraulic design criteria continue to be satisfied at the increased flow limit conditions.

As noted above, the current core thermal limits are conservative with respect to the increased minimum RCS total flow rate limit of 390,000 gpm. Based upon these protection limits, it was determined that the current overtemperature and overpower AT (OTAT/OPAT) setpoint equation constants (see Notes 1 and 2 of

-Table 3.3.1-1) for the McGuire and Catawba units are conservative and provide the necessary reactor protection.

Containment Functional Design (UFSAR Section 6.2.1)

The pertinent effect of an increase in the minimum RCS total flow rate limit on a reanalysis of the containment functional design analyses is the resultant impact on RCS temperatures. A review was performed relative to the assumpticns used for the following UFSAR containment analyses:

6.2.1.1 Peak Reverse Differential Pressure Analysis

6.2.1.2 Containment Subcompartment Analysis e 6.2.1.3 Mass and Energy Release Analysis for Postulated

! Loss of Coolant Accidents

6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures inside Containment

U.S. Nuclear Regulatory Comunission Attachment 3 June 24, 1999 Page 8 of 19

6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies of Emergency Core Cooling System Analysis 6.2.1.1 involves the peak reverse differential pressure across the operating deck separating' upper and lower containment.

Analysis 6.2.1.2 involves the short-term or blowdown peak containment pressure analysis following a LOCA or SLB, including' subcompartment pressurization analyses. This analysis simulates the compression of the initial air mass in containment immediately following the pipe rupture and lasts only seconds.

The results and conclusions from these UFSAR analyses are not affected by an increase in minimum RCS total flow rate limit, and no reanalysis is necessary.

The remaining three analyses involve sustained mass and energy releases into containment. RCS average temperature limit will remain unchanged with the change in minimum RCS flow rate limit.

Therefore, the initial stored energy in the RCS fluid and piping will remain approximately the same. Further, a constant RCS average' temperature implies that the driving temperature difference for primary to secondary heat transfer will remain unchanged. These two parameters, initial energy content and rate of energy transfer, are the means by which mass and energy releases influence containment response for the transients analyzed in Chapter 6 of the UFSAR. Since the increase in minimum RCS total flow rate limit is being made with a negligible change in RCS temperature limit, there will be negligible effect on the mass and energy releases calculated in UFSAR Chapter 6 events.

Accident Analysis (UFSAR Chapter 15)

The following section specifically addresses the impact of the j revised Technical Specifications on transients presented in l Chapter 15 of the McGuire and Catawba UFSARs.

The analyses that provide limiting core design limits for each cycle, typically DNB analyses where low RCS flow is conservative, were reanalyzed in order to provide the necessary margin for future core designs. All reanalyses are performed consistent with the NRC-approved methodology described in References 1 '

through 4. In general, the previous analyses used the BWCMV CHF correlation with a peak pin radial power (Fa) of 1.50 and an RCS flow of 382,000 gpm to determine the minimum DNBR. The transients providing limiting core design limits for each cycle were reanalyzed using the BWUZ CHF correlation with an Faa of 1.60 and the higher RCS flow of 390,000 gpm. The evaluation and

j l

U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 9 of 19 treatment of the power / flow tradeoff impact (if any) is performed on a-transient-specific basis. Specifically, this evaluation considers two separate cases for full power transients: 100% RTP

/ 390,000 gpm and 98% RTP / 386,100 gpm. {

The following description of the evaluation and reanalysis of limiting events applies to both McGuire units and to Catawba Unit 1. The existing Catawba Unit 2 DNB analyses were performed assuming an RCS flow rate of either 382,000 gpm or 385,000 gpm.

Since_both flow values are lower than the above 386,100 gpm limit,.the current analyses remain bounding. Upon completion of the unit-specific reanalyses for Catawba Unit 2, additional core peaking margin will be made available for future core designs.

A. Feedwater system malfunction causing an increase in feedwater flow-(15.1.2)

This ANS Condition II event is analyzed to show that DNB does not occur. The impact of a reduced power - reduced flow initial condition is explicitly analyzed in order to ensure that the most limiting case is found. The full power / full flow case is determined to be limiting.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The minimum DNBR is calculated to be 1.94, which is well above the 1.50 BWUZ SCD design limit.

B. Excessive increase in secondary steam flow (15.1.3)

This ANS Condition II event is analyzed to show that DNB does not occur. The impact of a reduced power - reduced flow initial condition is determined by a qualitative evaluation. This evaluation concludes that the reduction in initial power would more than offset the decrease in initial RCS flow.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. .This transient analysis showed that the reactor reached an' equilibrium condition that did not challenge the OPAT or OTAT reactor trip functions, which are designed to protect the core against DNB. Therefore, an explicit DNBR calculation was not performed.

I

U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 10 of 19 C. Inadvertent opening of a steam generator relief or safety valve (15.1.4)

This ANS Condition II event is analyzed to show that DNB does not occur. Since the transient is initiated from a zero power initial condition, a 1% reduction in the RCS total flow initial condition is assumed.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000, DPC-NE-3001,_and DPC-NE-3002. The peak thermal power level reached was less than half of that resulting from the steam system piping failure event. Therefore, an explicit DNBR calculation was not performed.

D. Steam system piping failure (15.1.5)

This ANS Condition III & IV event is analyzed to the more stringent Condition II criterion of ensuring that DNB does not occur. Since the transient is initiated from a zero power initial condition, a 1% reduction in the RCS total flow initial condition is assumed.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3001. The minimum DNBR is calculated to be 1.46, which is well above the 1.31 BWUZ non-SCD design limit.

E. Turbine trip (15.2.3)

This ANS Condition II event is analyzed to show that peak primary and secondary system pressures do not exceed the applicable limits. For the peak primary pressure case, the impact of a reduced power - reduced flow initial condition is determined by a qualitative evaluation. This evaluation concludes that the !

reduction in initial power and RCS flow would not change the j large amount of margin to the peak primary pressure acceptance criterion. For the peak secondary pressure case, a maximum RCS 5 flow has been shown to be conservative for this event; therefore, no reanalysis is necessary.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC- )

NE-3002. The calculated peak primary pressure is 2650.1 psig, which is significantly below the acceptance criterion of 2733.5 psig (110% of 2485 psig).

I I

U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 11 of 19 F. Loss of normal feedwater flow (15.2.7)

This ANS Condition II event is analyzed to demonstrate the capability of the secondary system to effectively cool the reactor core. 'For the short-term core-cooling case, the impact of a reduced power - reduced flow initial condition is explicitly analyzed in. order to ensure that the most limiting case is found.

The full power / full flow case is determined to be limiting. For the long-term core-cooling case, the impact of a reduced power -

reduced flow initial condition is determined by a qualitative evaluation. This evaluation concludes that the reduction in initial power would more than offset the decrease in initial RCS flow.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The minimum DNBR is calculated to be 2.31, which is well above the 1.50 BWUZ SCD design limit. The adequacy of the long-term core cooling capability is verified by the prevention 1 of hot leg boiling. A minimum subcooling of 40 F was reached during-the post-trip overheating phase of the transient. )

l G. Feedwater system pipe break (15.2.8)

This ANS Condition IV event is analyzed to demonstrate the capability of the secondary system to effectively cool the j reactor core. The impact of a reduced power - reduced flow q initial condition is determined by a qualitative evaluation. .

This evaluation concludes that the reduction in initial power would more than offset the decrease in initial RCS flow.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The adequacy of the long-term core cooling capability is verified by the prevention of hot leg boiling. A minimum subcooling of 16.9 F was reached during the overheating phase of '

the feedline break transient.

H. Partial loss of forced reactor coolant flow (15.3.1)

This ANS Condition II event is analyzed to show that DNB does not occur. The reduced low RCS flow reactor trip setpoints are incorporated into the transient reanalysis. The impact of a reduced power - reduced flow initial condition is determined by a qualitative evaluation. This evaluation concludes that the reduction in initial power would more than offset the decrease in initial RCS flow. !

l L

U.S. Nuclear Regulatory Comunission Attachment.3 June 24, 1999 Page 12 of 19 The analysis is performed in accordance with.the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002.- The minimum DNBR is calculated to be 2.16, which is well above the 1.50 BWUZ SCD design limit.

I. Complete loss of forced reactor coolant flow (15.3.2)

This ANS Condition II event is analyzed to show that DNB does not occur. The impact of a reduced power - reduced flow initial condition is explicitly analyzed in order to ensure that the most limiting case is found. The full power / full flow case is determined to be limiting.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The minimum DNBR is calculated to be 1.64, which is well above the 1.50 BWUZ SCD design limit.

J. Reactor coolant pump shaft seizure - locked rotor (15.3.3)

This ANS Condition IV event is analyzed to show that the peak primary system pressure does not exceed the applicable limit and to determine the percentage of fuel rods that experience DNB.

This fuel failure percentage is then applied in an offsite dose analysis in order to ensure that the radiological consequences do q not exceed a small fraction of the 10CFR100 limits. The reduced ;

low RCS flow reactor trip setpoints are incorporated into the transient reanalysis. The impact of a reduced power - reduced flow initial condition is determined by a qualitative evaluation.

This evaluation concludes that the reduction in initial power 3 would more than offset the decrease in initial RCS flow, i i

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The calculated peak primary pressure is 2534.6 psig, which is significantly below the acceptance criterion of 2733.5 psig (110% of 2485 psig). Since the minimum DNBR calculated with a standard axial power shape is found to fall below the 1.50 i design limit, maximum allowable radial peaking (MARP) curves are l generated in order to determine the number of fuel rods, if any, that experience DNB. The revised MARP curves allow greater radial peaking for all axial peaks and locations. Therefore, the fuel. failure assumption in the current offsite dose calculation remains valid and no reanalysis is required.

U.S. Nuclear Rwgulatory Cnemission Attachment 3 June 24, 1999 Page 13 of 19 K. Uncontrolled RCCA bank withdrawal from a subcritical or low power startup condition (15.4.1)

This ANS Condition II event is analyzed to show that peak primary system pressure does not exceed the applicable limit and that DNB does not occur. Since the transient is initiated from a zero power initial condition, a 1% reduction in the RCS total flow initial condition is assumed.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The calculated peak primary pressure is 2709.5 psig, which is below the acceptance criterion of 2733.5 psig (110% of 2485 psig). The minimum DNBR is calculated to be 2.70, which is well above the 1.50 BWUZ SCD design limit.

L. Uncontrolled RCCA bank withdrawal at power (15.4.2)

This ANS Condition II event is analyzed to show that peak primary and secondary system pressures do not exceed the applicable limits and that DNB does not occur. The impact of a reduced power - reduced flow initial condition is explicitly analyzed in order to ensure that the most limiting case is found. Also, for all cases initiated from an initial power level at or below 98%

RTP, a l% reduction in the RCS total flow initial condition is assumed.

The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The calculated peak primary and secondary pressures are 2679.6 psig and 1278.5 psig, respectively. The corresponding acceptance criteria are 2733.5 psig (110% of 2485 psig) and 1303.5 psig (110% of 1185 psig), Since the minimum DNBR i calculated with a standard axial power shape is found to fall l below the 1.50 design limit, MARP curves are generated in order to determine the number of fuel rods, if any, that experience DNB. The revised MARP curves allow greater radial peaking for all axial peaks and locations. Therefore, the conclusion that no fuel failures occur remains valid for the revised analysis.

M. Dropped RCCA rod (15.4.3a)

This ANS Condition II event is analyzed to show that DNB does not occur. The impact of a reduced power - reduced flow initial condition is explicitly analyzed in order to ensure that the most limiting case is found. The full power / full flow case is determined to be limiting.

l.

U.S.. Nuclear Regulatory Commaission Attachment 3 June 24, 1999 Page 14 of 19 LThe analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3001. The minimum DNBR is calculated to be at the 1.50 BWUZ SCD design limit.

N. Single uncontrolled rod withdrawal (15.4.3d)

This ANS Condition III event is analyzed to determine the percentage of fuel rods that experience DNB. This fuel failure i percentage is then applied in an offsite dose analysis in order i to ensure that the radiological consequences do not exceed a small fraction of the 10CFR100 limits. The impact of a reduced power - reduced flow initial condition is explicitly analyzed in order to ensure that the most limiting case is found. The full I power / full flow case is determined to be limiting. I The analysis is performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3002. The minimum DNBR is calculated to be 1.90, which is well above the 1.50 BWUZ SCD design limit. Despite the significant margin to the DNBR limit, MARP curves are generated in order to determine the number of fuel rods, if any, that experience DNB. The revired MARP curves allow greater radial peaking for all axial peaks and locations. Therefore, the fuel failure assumption in the current offsite dose calculation remains valid and no reana'.ysis is required.

O. Spectrum of RCCA ejection accidents (15.4.8)

This ANS Condition IV event is analyzed to show that the peak fuel pellet enthalpy and the peak primary side pressure do not exceed the acceptance criteria, and to determine the percentage of fuel rods that experience DNB. This fuel failure percentage is then applied in an offsite dose analysis in order to ensure that the radiological consequences are well within the 10CFR100 limits. The impact of a reduced power - reduced flow initial condition is conservatively analyzed in order to ensure that the most limiting case is found.

The analysis ic performed in accordance with the analytical model and methodology described in topical reports DPC-NE-3000 and DPC-NE-3001. The peak fuel pellet enthalpy is determined to be 101 cal /gm, which is significantly below the acceptance criterion of 280 cal /gm. The calculated peak primary pressure is 2693.7 psig, which is well below the acceptance criterion of 3000.0 psia.

Since the minimum DNBR calculated with a standard axial power shape is found to fall below the 1.31 design limit, MARP curves are generated in order to determine the number of fuel rods, if any, that experience DNB. The revised MARP curves allow greater

r U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 15 of 19 radial. peaking ~for all axial peaks and locations. Therefore, the fuel failure assumption in the current offsite dose calculation remains valid and no reanalysis is required.

The design limits determined by the above transient reanalyses are provided in the form of MARP curves and accident analysis input assumptions. The MARP curves and accident analysis input assumptions are utilized during the cycle-specific core design process as limits which the cycle design must meet to satisfy the licensing basis accident analyses. Reanalysis of these transients at a higher flow rate of 390,000 gpm provides

-additional core peaking margin for future core designs and {

results in significant DNB margin gains. For the analyses that produce the MARP curves used as limits in the cycle-specific core design process, this increased margin is reflected in the newly generated curves.

For'the following transients reanalysis is not required, since either a) the analysis is unaffected by the Technical J Specification changec b) the transient is non-limiting and any changes will have a favorable positive impact on the analysis results, or c) the transient is bounded by a more limiting transient of the same ANS Condition which is being reanalyzed.

A. Feedwater system malfunction causing a reduction in feedwater temperature (15.1.1)

This ANS Condition II event is bounded by the increase in

-feedwater flow and the increase in steam flow events. Therefore, a quantitative analysis of this transient is not performed.

B. Loss of external load (15.2.2)

This ANS Condition II event is bounded by the turbine trip event.

Therefore, a quantitative analysis of this transient is not performed.

C. Inadvertent closure of main steam isolation valves (15.2.4)

This ANS Condition II event is bounded by the turbine trip event.

Therefore, a quantitative analysis of this transient is not performed.

D. Loss of non-emergency AC power to the station auxiliaries (15.2.6)

This ANS Condition II event is analyzed to demonstrate the adequacy of the natural circulation cooling in removing core decay heat. Since the transient response is not particularly

r:

l-U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 16 of 19

. sensitive to the initial RCS flow rate, a reanalysis is not performed.

E. Reactor coolant pump shaft break'(15.3.4)

This ANS Condition IV event is bounded by the locked rotor event.

Therefore, a quantitative analysis of this transient is not performed.

F. Statically misaligned rod (15.4.3c)

This ANS Condition II event is analyzed to show that DNB does not occur. The core power distribution resulting from the misaligned control rod is compared to maximum allowable peaking limitt (MAPS). These MAPS, which preclude DNB, are conservatively based on an assumed RCS total flow rate of 382,000 gpm. Any negative impact of the higher RCS total flow rate limit on the calculated core peaking would be more than offset by improvementa in the M AP limits. Therefore, no reanalysis of this non-limiting ew'nt iu necessary.

G. Dropped RCCA bank (15.4.3b)

This ANS Condition II event is bounded by the dropped rod event.

Therefore, a quantitative analysis of this transient is not '

performed.

1 H. Startup of an inactive reactor coolant pump at an incorrect temperature (15.4.4)

This ANS Condition II event is analyzed to show that DNB does not occur. The current analysis, which assumes an RCS flow of 382,000 gpm, is shown by a qualitative evaluation to support the desired increase in core peaking factors. Therefore, no reanalysis of this~ nun-limiting event is necessary.

I. Chemical and volume control system malfunction that results in a decrease in boron concentration in the reactor coolant (15.4.6)'

.This ANS Condition II event is analyzed to ensure that the dilution is terminated, by manual or automatic means, prior to a i loss of shutdown margin. No explicit assumption is made for the l RCS flow rate, since the analysis results are completely insensitive to this parameter. Therefore, a reanalysis of this event is not required.

w U.S. Nuclear Regulatory Comunission Attachment 3 June 24, 1999 Page 17 of 19 l

1 J. Inadvertent loading and operation of a fuel assembly in an improper position (15.4.7)

This ANS Condition II event is analyzed to show that DNB does not occur. No explicit assumption is made for the RCS flow rate, since the analysis results are completely insensitive to this parameter. Therefore, a reanalysis of this event is not required.

K. Inadvertent operation of ECCS during power operation (15.5.1)

This ANS Condition II event is analyzed to show either that pressurizer overfill does not occur, or that the consequences of the overfill are not unacceptable. No explicit assumption is made for tDe ECS flow rate, since the analysis results are completely in2ansitive to this parameter. Therefore, a rcanalysis of this event is not required.

L. Chemical and volume control system malfunction that increases reactor coolant inventory (15.5.2)

The results of this ANS Condition II event are bounded by the i boron dilution and inadvertent safety injection transients.

Therefore, a quantitative analysis of this transient is not performed.

M. Inadvertent opening of a pressurizer safety or relief valve (15.6.1) i This ANS Condition II event is analyzed to show that DNB does not occur. The current analysis, which assumes an RCS flow rate of 382,000 gpm, already supports the desired increase in core peaking factors. Therefore, no reanalysis of this non-limiting event is necessary.

N. Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment (15.6.2) l This ANS Condition II event is analyzed to ensure that the l radiological consequences do not exceed a small fraction of the l 10CFR100 limits. No explicit assumption is made for the RCS flow I rate, since the analysis results are completely 2nsensitive to this parameter. Therefore, a reanalysis of this event is not required. .

l

U.S. Nuclear Regulatory Commission Attachment 3 Maus 24, 1999 Page 18 of.19 O. Steam generator tube rupture (15.6.3)

This ANS Condition IV event is analyzed to show that a) DNB does not occur, b) the calculated offsite doses do not exceed a small fraction of the 10CFR100 limits, and c) SG overfill is avoided (Catawba only). The current analysis, which assumes an RCS flow of-382,000.gpm, is shown by a qualitative evaluation to support the desired increase.in core peaking factors. Therefore, no reanalysis of this non-limiting event is necessary. Since the results of the analysis cases focusing on the two other

. acceptance criteria are not particularly sensitive to the initial ;

RCS flow rate, no reanalyses of these cases are performed either.

P. Loss of coolant accidents (15.6.5)

These ANS Condition III & IV events are analyzed in accordance with 10CFR50.46 and 10CFR50 Appendix K. The current LOCA analyses are performed assuming an RCS total flow rate of 382,000 gpm, which is the current minimum TS flow rate limit. Although the results of the LOCA analyses are not particularly sensitive to the initial RCS flow rate, an increase in this minimum RCS total flow rate limit could only be an analysis benefit.

Therefore, a reanalysis of this event is not required.

Q. Postulated secondary system pipe rupture outside containment This event is analyzed to ensure that the Doghouse equipment qualification temperature limit is not exceeded. A maximum RCS ,

flow rate has been shown to be conservative for this event; I therefore, no reanalysis is necessary.

Conclusion An increase in the Technical Specification minimum RCS total flow rate limit and the revised power / flow tradeoff will not adversely affect the steady-state or transient analyses documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs. The limiting UFSAR Chapter 15 transients were reanalyzed to provide new cycle-specific core design limits in the form of new MARP curves and accident analysis input assumptions.to be utilized during the cycle-specific core design process.

Summary It is proposed to increase the current minimum RCS total flow rate limit to 390,000 gpm and to permit operation with the measured RCS total flow rate between 386,100 and 390,000 gpm with a 2% core power reduction. The reactor core safety limits j l

l u i

r~ q U.S. Nuclear Regulatory Commission Attachment 3 June 24, 1999 Page 19 of 19 figure, which is based upon the minimum RCS total flow rate limit, is revised accordingly.

In order to provide more margin in the core design limits and allow more flexibility for future cycle-specific core design, the analyses that establish these limits were reanalyzed at the 1 proposed TS minimum RCS total flow rate limit. The impact of the power / flow tradeoff is determined for each reanalyzed event either by qualitative evaluation or by explicit reanalysis. A j reduced low RCS flow reactor trip setpoint is also incorporated into the affected transient reanalyses.(McGuire only). A typographical error associated with T' in Note 1 of Table 3.3.1-1 is corrected (McGuire only). The RCS DNB parameter limits are revised to be consistent with these DNBR transient reanalyses.

Conditions B and C of TS 3.4.1 and the Bases associated with TS 3.3.1 and 3.4.1 are also revised.

References

1. DPC-NE-3 00 0-PA, " Duke Power Company, Oconee Nuclear Station, McGuire Nuclear Station, Catawba' Nuclear Station, Thermal-Hydraulic Transient Analysis Methodology," Duke Power Company, Revision 2 (SER dated October 14, 1998).

2. DPC-NE-3 0 01-PA, " Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology," Duke Power Company, November 1991.

3. DPC-NE-3 0 02-A, " Duke Power Company, McGuire Nuclear Station, Catawba Nuclear Station,' FSAR Chapter 15 System Transient Analysis Methodology," Duke Power Company, Revision 3 (May 19, 1999). i

4. DPC-NE-2 0 05P-A, " Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology," Duke Power Company, Revision 1, November 1996.

5 .' L i c e n s e e E v e n t R e p o r t 369/97-10, Revision 0, McGuire Nuclear Station, Unit 1 and 2, Docket No. 50-369.

6. Licensee Event Report 413/97-007, Revision 0, Catawba Nuclear Station, Unit 1 and 2, Docket No. 50-413.

7 J.S. Nuclear Regulatory Commission Attachment 4 June 24, 1999 Page 1 of 2 No Significant Hazards Considerations As required by 10 CFR 50.91, this analysis is provided to determine whether the requested amendments involve significant hazards considerations, as defined by 10 CFR 50.92. An amendment request involves no significant hazards considerations if operation of the facility in accordance with the requested 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.

Criterion 1 - Would operation of the facility in accordance with the requested amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No component modification, system realignment, or change in operating procedure will occur which could affect the probability of any accident or transient. The increase in RCS total flow rate limit will not change the probability of actuation of any Engineered Safety Feature or other device. In order to provide more margin in the core design limits and allow more flexibility !

for future cycle-specific core design, the analyses that )

establish these limits were reanalyzed at the proposed TS minimum RCS total flow ~r ate limit. The impact of the power / flow tradeoff is determined for each reanalyzed event either by qualitative ;

evaluation or by explicit reanalysis. l l

An increase in the Technical Specification minimum RCS total flow I rate limit and the revised power / flow tradeoff will not adversely l affect the steady-state or transient analyses documented in l Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear ;

Station UFSARs. The reduced RCS low flow reactor trip setpoint i and allowable value will not increase the consequences of the partial loss of forced reactor coolant flow and reactor coolant pump shaft seizure accidents. In these transient reanalyses, the minimum DNBR and peak primary system pressure acceptance criteria are not adversely affected. Therefore, the proposed changes will not involve an increase in the probability or consequences of an accident previously evaluated.

r U.S. Nuclear Regulatory Commission Attachment 4 June 24, 1999 Page 2 of 2 Criterion 2 - Would operation of the facility in accordance with the requested amendment create the possibility of a new or different kind of accident from any previously evaluated? I No component modification, system realignment, or change in operating procedure will occur which could create the possibility of a new or different kind of accident. As described in Attachment 3, the proposed increase in Technical Specification minimum RCS total flow rate limit and revised power / flow tradeoff will not adversely affect the steady-state or transient analyses documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3 - Would operation of the facility in accordance with the requested amendment involve a significant reduction in a margin of safety?

These amendments will not involve a significant reduction in a margin of safety. As described in Attachment 3, the increase in minimum RCS total flow rate limit and revised power / flow tradeoff will not adversely affect the steady-state or transient analyses documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs. DNBR, fuel clad integrity, reactor vessel integrity and containment integrity will not be adversely affected by the proposed changes. Therefore, the proposed !

changes will not involve any reduction in a margin of safety. l

Conclusion:

Based on the above analysis, McGuire and Catawba Nuclear Stations I conclude that the requested amendments involve no significant i hazards considerations.

^

r- 1 U.S. Nuclear Regulatory Commission Attachment 5 j June 24, 1999 Page 1 of 1 Environmental Assessment / Impact Statement l

Pursuant to 10 CFR 51.22(b), an evaluation of this license amendment request has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9) of the regulations.

Theses proposed amendments have no direct impact on any effluent generation or control systems. The changes do not impact the type or quantity of effluents from the station. The proposed 1 changes do not impact individual or cumulative occupational radiation exposure.

It has been determined that there is:

1) No significant hazards consideration (see Attachment 4);

2) No significant change in the types, or significant increase in the amounts, of any effluents that may be released

- offsite; and

3) No significant increase in individual or cumulative occupational radiation exposures involved. l Therefore, these amendments to the McGuire and Catawba Technical Specifications meet the criteria of 10 CFR 51.22 (c) (9) for categorical exclusion from an environmental impact statement.

1 b

ML20196H755

Text

Conclusion:

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