ML20206E367

Proposed Tech Specs,Changing Unit 2 RAOC Envelope to Reflect Increase in Max Allowable Heat Flux Hot Channel Factor & Changing Unit 1 Tech Specs Administratively

ML20206E367
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/09/1987
From:	DUKE POWER CO.
To:
Shared Package
ML20206E351	List: ML20206E345 ML20206E367
References
TAC-65072, TAC-65073, NUDOCS 8704130592
Download: ML20206E367 (125)

Text

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INDEX LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4 1-21 Control Rod Insertion Limits..............................

FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP 0PERATION...................................

3.4 1-22

........ 3/4 1-23 FIGURE 3.1-2 (BLANK)......................................

3/4.2 POWER DISTRIBUTION LIMITS AXI AL F LUX DIF F ERENC E . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 . . . 2-1 l;

3/4.2.1 . .'. . . . . .P FIGURE 3.2-1/ AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF3/4 2-3

[ '

RATED THERMAL POWER 3Hn44::D.......................

f

= TIGU;E 3.2-lb AXIAL cLlw gyrerngycg L;u:75 AS A FU"CT40N -Of 1 PAT 5e T95E"At ace 5= (Unit 2)................... . 2/4 2-4 y 3/4 2-6 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR -qF (Z). . . . . . . . . . . . . . . . . . . . . .

FIGURE 3.2-2a K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT 3/4 2-12 (Unit 1).........g..................................

FIGURE 3.2-2b K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 3/4 2-13 (Unit 2)...........................................

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL 3/4 2-14 FACT 0R..................................................

3/4 2-16 FIGURE 3.2-3a RCS TOTAL FLOW RATE VERSUS R (Unit 1)................

FIGURE 3.2-3b RCS FLOW RATE VERSUS Rg AND R 2 FOUR LOOPS IN OPERATION (Unit 2).............................. 3/4 2-17 R00 BOW PENALTY AS A FUNCTION OF BURNUP (Unit 2). . . . . 3/4 2-18 FIGURE 3.2-4 3/4 2-19 3/4.2.4 QUADRANT POWER TILT RATI0.................................

3/4.2.5 DN8 PARAMETERS............................................

3/4 2-22 3/4 2-23 TABLE 3.2-1 DN8 PARAMETERS.......................................

3/4.3 INSTRUMENTATION ,

3/4 3-1 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.......................

8704130592 870409 PDR ADOCK 05000369 P PDR McGUIRE - UNITS 1 and 2 V Amendment No. (Unit 1)

Amendment No. (Unit 2)

I INDEX ADMINISTRATIVE CONTROLS PAGE SECTION 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) 6-9 Function..................................................

6-10 Organization..............................................

6-11 Review....................................................

6-11 Audits....................................................

6-12 Authority.................................................

6-13 Records...................................................

6-13  ?.,

6.6 REPORTABLE EVENT ACTI0N........................................

i 6.7 SAFETY LIMIT VIOLATION.........................................

6-13 dllf

............... 6-14 6.8 PROCEDURES AND PR0 GRAMS........................

6.9 REPORTING REQUIREMENTS 6-16 i

6.9.1 ROUTINE REP 0RTS.............................................. -

Startup Report............................................

6-16 ;7 6-17 Annual Reports............................................

Annual Radiological Environmental Operating Report........ 6-18 Report............ 6-18 Semiannual Radioactive Effluent Release 6-20  :

Monthly Operating Reports................................. '

P 4

6-21 l

]Ryd413'PeakingFactorLimitReport............'............

McGUIRE - UNITS 1 and 2 XXII ' Amendment No. (Unit 1)

Amendmert No. (Unit 2) .

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& wfuman=a o~sy 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION f

l 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within: 1

a. the allowed operational space defined by Figure 3.2-1 for RAOC operation, or

b. within a 15 percent target band about the target flux difference during base load operation.

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *.

• 1 .

i ACTION:

I

a. For RAOC operation with the indicated AFD outside of the Figure 3.2-1 1 1

limits, l

1. Either restore the indicated AFD to.within the Figure 3.2-1 limits within 15 minutes, or l

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux -

High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b. For Base Load operation above APL 0" with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target l

flux difference:

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

ND

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

c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER utiless the indicated AFD is within the Figure 3.2-1 limits.

• See Special Test Exception 3.10.2.

• • APLND is the minimum allowabla power level for base load operation and will be provided in the Peaking Factor Limit Report per Specifftation 6.9.1.9.

McGUIRE - UNITS 1 and 2 3/4 2-1 Amendment No.43(Unit 1)

Amendment No.24(Unit 2)

i y C H%U 94 INf*AM D *" W POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

a .~ Monitoring the indicated AFD for each OPERA 8LE excore channel:

-1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

, 2. At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFU Monitoring Alare to OPERA 8LE status.

+

b. Monitoring and logging the indicated AFD for each'0PERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed

- to exist duri'ng the interval preceding each logging. I i 4.2.1.2 The indicated AFD shall be considered outside of its limits when at

least two OPERABLE excore channels are indicating-the AFD to be outside the l

limits.

4.2.1.3 When in Base L'oad operation, the target axial flux difference of I each OPERABLE excore channel shall be determined by measurement at least once I per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are  :

not applicable. ,

5 4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effactive Full Power Days by either determining the target flux difference in conjunction with the surveillance requirements of Specification 3/4.2.2 or by linear interpolation between the most recently measured value and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable.

~

3/4 2-la Amendment No A2 (Unit 1)

McGUIRE - UNITS 1 and 2 Amendment No.23 (Unit 2) l

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Amendment No23 (Unit 2)

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AFD Limits as a Tunction of Rated Thermal Power M i ,

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3/42-3 Amendment No. (Unit 1)

McGUIRE UNITS 1 and 2 Amendment No. (Unit 2)

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! AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED TMERMAL P

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McGUIRE - UNITS 1 and 2 3/42-4 Amendment No (Unit 1)

Amendment No (Unit 2) .

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s McGUIRE - UNITS 1 and 2 3/4 2-5 Amendment No.42(Unit 1)

______ _Mndment No.23(Unit 2)_ _ _ _ _ _

POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg I LINITING CONDITION FOR OPERATION e

l j

3.2.2 F q(Z) shall be limited by.the following relationship: {c t .3 t Fq (Z) 1 [ [K(Z)] for P > 0.5 h 2.w I El Fq (Z) $ Q 3 [K(Z)] for P $ 0.5

) Where: P _ RATED THERMAL POWER THERMAL POWER '

and K(Z) is the function obtained from Figure 3.2-2 for a given d

core height location.

APPLICABILITY: MODE 1.

ACTION:

f With Fq (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% Fn(Z) exceeds the limit within 15 minutes and similarly reduce the PowVr Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER i

OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K 4) have been reduced at least 1% (in AT span) for each 1% F q (Z) exceeds the limit; and

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a. , above; THERMAL POWER may then be increased provided Fq (Z) is deponstrated through incore mapping to be within its limit.

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i 3/4 2-6 Amendment No. (Unit 1)

McGUIRE - UNITS I and 2 Amendment No. (Unit 2)

4 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 For RA0C operation,qF (z) shall be evaluated to determine if Fq (z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution '

map at any THERMAL POWER greater than 5% of RATED.TNERMAL POWER.

I

b. Increasing the measured F q (z) component of the power distribution map by 3% to account for manufacturing tolerances anu further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

c. Satisfying the following relationship: ,

t.n

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,}

F M

q (2) 1'#~I' * "I*) for P > 0.5 P x W(z)  ?

M * "I*) for P < 0.5 l F0 (z) ~< W(z) x 0.5 where F (z) is the measured F (z) increased by the allowances for i 9

manufacturing tolerances and measurement uncertainty, M is the >l '

Fg limit, K(z) is given in Figure 3.2-2, P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Speci-fication 6.9.1.9.

M

d. Measuring Fg (z) according to the following schedule: '

1. Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which Fq (z) was last determined," or

2. At least once per 31 Effective Full Power Days, whichever occurs first. l 1

i

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

McGUIRE - UNITS 1 and 2 3/4 2-7 Amendment No. Unit 1)

Amendment No. (Unit 2)

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j. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) i

e. With measurements indicating l

maximum

! [FM (z) l over z ( K(z)/ N has increased since the previous determination of Fg (z) either of l the following actions shall be taken:

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l 1) F q (z) shall be increased by 2% over that specified in Specifi-cation 4.2.2.2c. or M .

2) F g (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that is not increasing.

maximum [F (z)

I over z \ K(z)/

,' f.

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

1) Calculate the percent Fg (z) exceeds its limit by the fc11owing l expression:

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b I F M ffeaximum F0 (*)

• WI*) -

'1 r x 100 for P > 0.5 \

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s *IK(z) f r7 N3 maximum *M i

! F 0

II) * (*) > x 100 for P < 0.5 i

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(kOV'"* .m g xa-n K(z).)51 ,

2) One of the following actions shall be taken:

a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits of 3.2-1 by

! 1% AFD for each percentqF (z) exceeds its limits as deter-l mined in Specification 4.2.2.2f.1). Within.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified Ifmits, or Comply with the requirements of Specification 3.2.2 for Fg(z) b")

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exceeding its limit by the percent calculated above, or l l

! c) Verify that the requirements of Specification 4.2.2.3 for

' Base Load operation are satisfied and enter Base Load operation.

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3/4 2-8 Amendment No. Unit 1) )

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

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d POWER OISTRIB TION LIMITS l

SURVEILLANCE REQUIREMENTS (Continued)

g. The limits specified in Specifications 4.2.2.2c, 4.2.2.2e. , and 4.2.2.2f.

above are not applicable in the following core plane regions:

i 1. Lower core region from 0 to 15%, inclusive. l

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

ND 4.2.2.3 Base Load operation is permitted at powers above APL if the

! following conditions are satisfied: 1

! a. Prior to entering Base Load operation, maintain THERMAL POWER above l NO and less than or equal to that allowed by Specification 4.2.2.2  !

APL for at least the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Maintain Base Load operation l surveillance (AFD within *5% of target flux difference) during this

j

! time period. Base Load operation is then permitted providing THERMAL '

ND Il NO

! POWER is maintained between APL and APL or between APL and 100% (whichever is most limiting) and FQ surveillance is maintained

' l BL I pursuant to Specification 4.2.2.4. APL is defined as: 0 l

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APLOL = I l om"I"I"""Z [ (.2<f6'x KCZ) ] x 100%

F (Z) x W(Z)BL . p l where: F (z) is the measured Fq (z) increased by the allowances for.

manu,fgeturingtolerancesandmeasurementuncertainty. The qF limit i s ,2J6~. K(z) is given in Figure 3.2-2. W(z)BL is the cycle dependent y function that accounts for limited power distribution transients en-countered during base load operation. The function is given in.the Peaking Factor Limit Report as per Specification 6.9.1.9.

! b. During Base Load operation, if the THERMAL POWER is decreased below ND l APL then the conditions of 4.2.2.3.a shall be satisfied before i re-entering Base Load operation.

4.2.2.4 During Base Load Operation qF (Z) shall be evaluated to determine if F (Z) is within its limit by:

q

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

b. Increasing the measured F q (Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Verify the requirements of Specification 3.2.2 are satisfied.

1 4 .

l McGUIRE - UNITS 1 and 2 3/4 2-9 Amendment No. (Unit 1)

Amendment No (Unit 2) l

l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

c. Satisfying the following relationship:

s.s>

x NO [}

Z) for P > APL h F"Q(Z) < P x W(Z)BL to Ll 4

where: F"(Z) is the measured qF (Z). The F q limit is.Jeetr.

K(Z) is given in Figure 3.2-2. P is the relative THERMAL POWER.

W(Z)BL is the cycle dependent function that accounts for limited power distribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9. ,

d. Measuring F"(Z) in conjunction with target flux difference deter-l .

' mination according to the following schedule: ,

1. Prior to entering BASE LOAD operation after satisfyirig Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been ND for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to mapping, and maintained above APL "

2. At least once per 31 effective full power days.

1. ~

e. With measurements indicating maximum ( )

over Z

'has increased"since the previous determination F"(Z) either of the following actions shall be taken: ,

1. FH (Z) shall be increased by 2 percent over that specified in g

4.2.2.4.c, or

2. F (Z) shell.be measured at least once per 7 EFPD' until 2 l successive maps indicate that F (Z) i maximum [ ] is not increasing.  !

' over z f.

With the relationship specified in 4.2.2.4.c above not being satisfied, either of the following actions shall be taken:

i

1. Place the core in an equilbrium condition bre the limit in 4.2.2.2.c is satisfied, and remeasure F (Z), or 3/4 2-9a Amendment No. (Unit 1)

McGUIRE - UNITS I and 2 Amendment No (Unit 2)

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. 1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2. Comply with the requirements of Specification 3.2.2 for Fg (Z) exceeding its limit by the percent calculated with the following y; expression: 1 r M i ,

0 F (Z) x W(Z)BL ] ) -1 ] x 100 for P > APL 3

[(max. over z of [ [ S (Z) ('l

g. The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plan regions:

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

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

! 4.2.2.5 When F (Z) is measured for reasons other than meeting the requirements g

l of specification 4.2.2.2 an overall measured Fq(z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

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McGUIRE - UNITS 1 and 2 3/4 2-9b Amendment No. Unit 1)

Amendment No. (Unit 2)

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' 3/4 2-10 Amendment No.42(Unit 1) ,

McGUIRE - UNITS 1 and 2 Amendment No.23(Unit 2)  !

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McGUIRE - UNITS 1 and 2 3/4 2-11 Amendment No.42(Unit 1)

Amendment No.23(Unit 2)

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3/4 2-13 Amendment No.43(Unit 1)

McGUIRE - UNITS I and 2 Amendment No.24(Unit 2)

ADMINISTRATIVE CONTROLS J

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6.9.1.9 M PEAKING FACTOR LIMIT REPORT ,

1

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functions for RAUG and Base Load operation and the val __ j .

(as requ r be provided to the Director, N ctor Regulation, J Attention: Chief, Core ance , . . Nuclear Regulatory Commission, p Washington, D.C. 20555 at to cycle initial criticality. <  ;

g In the event th e values would be submitte ther time during  ; '

j core 1 ,

will be submitted 60 days prior to the date the _uld

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e effective unists otherwits ovemnted by the Cnemission. r

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ND Any information needed to support W(z), W(z)BL and APL will be by request f

from the NRC and need not be included in this report. '

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SPECIAL REPORTS .

.' 6.9.2 Special reports shall be submitted to the Regional Administrator of the

• NRC Regional Office wit _hin the time period specified for each report. .-

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s r.o n paa pnos n~e sau cono aptnan=~ A~r ikE VALwf FDA APL

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McGUIRE - UNITS 1 and 2 6-21 Amendment No. '(Unit 1)

Amendment No (Unit 2)

l l

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity -

during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core at or above the design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop-1 erties to within assumed design criteria. In addition, limiting the peak linear j power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F0 (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing toler-ances on fuel pellets and rods; F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of H the integral of linear power along the rod with the highest integrated power to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper [

p 2.n Q 1 bound envelope of 3-f6' times the normalized axial peaking factor is not exceeded 4l during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with ,

their respective insertion limits and should be inserted near their normal l position for steady-state operation at high power levels. The value of the i target flux difference obtained under these conditions divided by the fraction  !

of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER l 3

for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

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

Amendment No. (Unit 2)

N o CrfA w o f f 1

f*t 1~MMnw s, j,

POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

At power levels below APL , the limits on AFD are defined by Figures 3.2-1, These limits were i.e. that defined oy the RAOC operating procedure and limits.

calculated in a manner such that expected operational transients, e.g. load follow limits.

operations, would not result in the AFD deviating outside time allowed outside of the limits at reduced power levels will not result in would change sufficiently to prevent operation in the vicinity power level.

E At power levels greater than APL , two modes of operation are permissible;

1) RAOC, the AFD limit of which are defined by Figure 3.2-1, and 2) Base Load operation, which is defined as the maintenance of the AFD E within a 1 5% band is the same as about a target value. The RAOC operating procedure above APL E However, it is possible when following that defined for operation below APL .

extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with q F (z) less than its limiting value. To allow operation at the maximum permissible valu3 the Base Load operating procedure restricts the indicated AFD to relatively O smal g 1 power 5 APL ' or target band and power swings (AFD target band of 15%, APLFor Base Load operation, it i 100% Rated Thermal Power, whichever is lower). Operation outside expected that the plant will operate within the target band.of the ficant xenon redistribution such that the envelope of peaking factors would change sufficiently to prohibit continued operation in the power region defined above.

To assure there is no residual xenon redistribution impact from past operation on the Base Load operation, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> waiting period at a power leve and allowed by RAOC is necessary. During this time period load above APL changes cedure.

and rod motion are restricted to that allowed by the Ba The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD

1) outside the for at least 2 of 4 or 2 of 3 OPERABLE excore channels are:

allowed Al power operating space (for RAOC operation), or 2) outside the These alarms are active allowed AI target band (for Base Load operation).1) 50% of RATED THER when power is greater than:

, Penalty deviation minutes for Base Load or 2) APL (for Base Load operation).

operation are not accumulated based on the short period of time during which ,

operation outside of the target band is allowed.

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8'3/4 2-2 Amendment No. 43(Unit 1) i McGUIRE - UNITS 1 and 2 Amendment No. 24(Unit 2)

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• R IMP 4h AnJH Owvy POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits-on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure.that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS accep-tance criteria limit.

Each of these is measuracle but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is 1

sufficient to insure that the limits are maintained provided:

a. Control rods in a single group move together with no. individual rod insertion differing by more than + 13 steps from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6;

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

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.  ;

I FhwillbemaintainedwithinitslimitsprovidedConditionsa.through

d. above are maintained. As noted on Figure 3.2-3, RCS flow rate and power may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the power level is decreased) to ensure that the calcu- )

lated DNBR will not be below the design DNBR value. TherelaxationofFhas a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

R as calculated in Specification 3.2.3 and used in Figure 3.2.-3, accounts forFhlessthanorequalto1.49. This value is used in the various accident analyses where F influences parameters other than DNBR, e.g., peak clad tem-H perature, and thus is the maximum "as measured" value allowed.

Margin between the safety analysis limit DNBRs (1.47 and 1.49 for thimble i and typical cells, respectively) and the design limit DNBRs (1.32 and 1.34 for thimble and typical cells, respectively) is maintained. A fraction of this margin is utilized to accommodate the transition core DNBR penalty (2%) and the appropriate fuel rod bow DNBR penalty (WCAP - 8691, Rev.,.1).

When an F measurement is taken, an allowance for both experimental error q

and manufacturing tolerance must be made. An allowance of 5% is appropriate l

l McGUIRE - UNITS 1 and 2 B 3/4 2-2a Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

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McGUIRE - UNITS 1 and 2 B3/42-3 Amendment No. 42 (Unit 1)

Amendment No. 23 (Unit 2)

,p.

p c Ane's fan IrdoAMn~ o ~r POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) for a full-core ~ map taken with the Incore Detector Flux Mapping System, and a l 3% allowance is appropriate for manufacturing tolerance.

When RCS flow rate and F are measured, no additional allowances are H

necessary prior to comparison with the limits of Figure 3.2-3. Measurement errors of 1.7% for RCS total ' flow rate and 4% for Fhhave been allowed for in determination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-3. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by '

monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and ,

compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3.

The hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power factor appropriate to either RAOC or Base Load operation, W(z) or W(z)BL, to provide assurance that the limit on the hot channel factor, Fq (z), is met. W(z) accounts for the effects of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. W(z)BL accounts for l the more restrictive operating limits allowed by Base Load operation which l 1

result in less severe transient values. The W(z) function for normal operation is provided in the Peaking Factor Limit Report per Specification 6.9.1.9.

l

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McGUIRE - UNITS 1 and 2 B 3/4 2-4 Amendment No. 42 (Unit 1)

Amendment No. 23 (Unit 2)  ;

ATTACHMENT 2 JUSTIFICATION AND SAFETY ANALYSIS

I ATTACHMENT 2 JUSTIFICATION AND SAFETY ANALYSIS 4

Mr. H.B. Tucker's (DPC) November 14, 1983 letter to Mr. H.R. Denton (NRC/0NRR) described planned changes in the fuel design for McGuire Nuclear Station, Units 1 and 2. McGuire Unit I had been operating with a Westinghouse 17x17 low-parasitic (STD) fueled core. It was planned to refuel Unit I with Westinghouse 17x17 Raconstitutable Optimized Fuel Assembly (OFA) regions. As a result, future core loadings would range from an approximately 1/3 0FA - 2/3 STD transition core to i eventually an all 0FA fueled core. Major advantages for utilizing the OFA are:

(1) increased efficiency of the core by reducing the amount of parasitic material

and (2) reduced fuel cycle costs due to an optimization of the water to uranium ratio. This letter provided a Reference Safety Evaluation Report summarizing the evaluation / analysis performed on the region-by-region reload transition from the McGuire Units 1 and 2 STD fueled cores to cores with all optimized fuel. The report examined the differences between the Westinghouse OFA and STD designs and evaluated the effects of these differences for the transition to an all 0FA core.

The evaluation considered the standard reload design methods described in WCAP-9272 and 9273, " Westinghouse Reload Safety Evaluation Methodology," and the transition effects described for mixed cores in Chapter 18 of WCAP-9500-A, "Ref-erence Core Report - 17x17 Optimized Fuel Assembly." Consistent with the West-inghouse STD reload methodology for analyzing cycle specific reloads, parameters were chosen to maximize the applicability of the transition evaluations for each reload cycle and to facilitate subsequent determination of the applicability of 4 10CFR 50.59. Subsequent cycle specific reload safety evaluations were to verify that applicable safety limits are satisfied based on the reference evaluation /-

analyses established in the reference report and/or subsequent modifications to i

the McGuire licensing bases. A summary of the mechanical, nuclear, thermal and hydraulic, and accident evaluations for the McGuire Units 1 and 2 transitions to

)

an all 0FA core were given in the reference report. '

l The results of evaluation / analysis and tests described in the Reference Safety l Evaluation Report led to the following conclusions:

a. The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mech-anically compatible with the STD design, control rods, and reactor internals interfaces. Both fuel assemblies satisfy the design bases for the McGuire units,

b. Changes in the nuclear characteristics due to the transition from STD to OFA fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

i

c. The reload 0FAs are hydraulically compatible with the STD design.

d. The accident analyses for the OFA transition core were shown to provide l i acceptable results by meeting the applicable criteria, such as, minimum DNBR, l peak pressure, and peak clad temperature, as required. The previously reviewed and licensed safety limits were met. Analyses in support of this safety evaluation establish a reference design on which subsequent reload safety evaluations involving 0FA reloads can be based. (Attachment 2A of H.B. Tucker's December 12, 1983 Unit 1/ Cycle 2 0FA reload submittal presented those detailed non-LOCA and LOCA accident analyses of the McGuire Units 1 and 2 FSAR impacted by the changes as determined in Section 6.0 of the Reference Safety Evaluation Report).

I I

- . .. _ .. . . - . .. .. -. ~ . . . . -. . . .

Attachment 2 4 Page 2 i

I

e. Plant operating lLaitations given in the Technical Specifications affected by 1

use of the OFA design and positive MTC would be satisfied with the changes noted in Section 7.0 of the report.

i

McGuire Unit 2 is currently operating in Cycle 3 with Westinghouse 17x17 low parasitic (STD) fuel assemblies and optimized fuel assemblies (OFA) following previous NRC approval of an OFA reload region (reference McGuire Facility Oper-ating License amendments 42 (Unit 1)/23 (Unit 2)), with a second 0FA reload having been accomplished under the provisions of 10CFR 50.59 as indicated in a Tucker to Denton letter of February 21, 1986. The third such 0FA region is scheduled for i

the upcoming Cycle 4 refueling. (McGuire nU it 1 is currently operating in cycle 4 with its third such 0FA reload region).

Attachment 2A is the cycle-specific Reload Safety Evaluation (RSE) for McGuire i Unit 2/ Cycle 4. The RSE presents an evaluation for McGuire Unit 2, Cycle 4, which i

demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was performed utilizing the methodology described in WCAP-9273-A, " Westinghouse Reload Safety Evaluation Methodology". As indicated above, the NRC has previously approved similar OFA reloads for McGuire Unit 2 (and 1). The compatibility of the OFA design with the Standard design in a transition core as well as its adequacy in a full 0FA core was shown in the j November 14, 1983 reference safety evaluation licensing submittal and subsequent i

submittals which modified the McGuire licensing basis.

All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload have been reviewed for the Cycle 4 design. The results of new analyses and the justification for the applicability of previous.

j analyses are addressed in the cycle specific reload safety evaluation.

i Unit 2 Cycle 4 will have fuel assemblies on the core periphery with clips supplied i by Advanced Nuclear Fuels in order to reduce concerns associated with water

! jetting from the barrel / baffle region into the core (reference Tucker to Denton j letter dated August 4,1986 and McGuire License Amendment 60 (Unit 1)/41 l (Unit 2)). A 50.59 evaluation has determined that the clips do not introduce an i

unreviewed safety question.

1

' From the evaluation presented in the Cycle 4 Reload Safety Evaluation, it is concluded that the Cycle 4 design does not cause any of the safety limits to be

{ exceeded. This conclusion is based on the following:

4

1. Cycle 3 burnup is between 10500 and 11500 MWD /MTU.

i 2. Cycle 4 burnup is limited to 13500 MWD /MTU including a coastdown.

t

3. There is adherence to all plant operating limitations given in the Technical Specifications as revised by the proposed changes given in Appendix A of the Cycle 4 RSE.

4. The large break LOCA analysis for UHI removal using the 1981 EM+ BASH is approved.

4 i

i Attachment 2 Page 3 i

i

To allow the increased operating flexibility justified by the Cycle 4 RSE (At-tachment 2A) and the BASH LOCA analyses (Attachment 2B), Technical Specification changes will be needed for Cycle 4 to incorporate a wider RAOC Delta-I envelope and heat flux hot channel factor, Fq, of 2.32 (the Fq increase is also applicable )

for Unit 1 and the wider RAOC envelope was approved for Unit' I prior to Cycle 4 '

operation - reference Mr. D.S. Hood's August 22, 1986 letter to H.B. Tucker). The existing Unit 2 RAOC limits were developed based upon an Fq limit of 2.15 and thus are overly restrictive considering the current Fq limit of 2.26 or the proposed Fq limit of 2.32. The wider RAOC limits simply reflect the increased Fq limit as was j discussed in the Unit 1 Cycle 4 submittal of May 15, 1986 and the Unit 2 Cycle 4 RSE (Attachment 2A).

Attachment 2B provides the supporting documentation for the proposed increase in the heat flux hot channel factor, Fq, to 2.32 (excerpted from reference 12 of the M2/C4 RSE). The justification of the 2.32 Fq limit was performed using the NRC

approved BASH methodology and the modifications and enhancements described in revision 3 of WCAP-10266. A complete spectrum of breaks were analyzed per i

ICCFR 50.46 with the limiting break being Dn -0.6 with minimum safety injection assump n ons resulting in a predicted peak clad temperature of 1841*F. The BASH analyses satisfy the requirements of the NRC in regard to the replacement of the existing licensing basis BART analyses which included errors in modeling several core parameters (reference NRC/0IE Generic Letter 36-16). The reanalysis incor-porated recent plant changes related to the deletion of the Upper Head Injection System (UEIS) which have been previously approved by the NRC. Please note that the small break analyses which were submitted to justify deletion of the UHIS and approved by the NRC (reference Mr. D.S. Hood's May 13, 1986 letter to H.B. Tucker) are re-transmitted in Attachment 2B for completeness, although they have not changed. The McGuire FSAR will be revised accordingly in the appropriate annual l FSAR update following approval of this change.

I Attachment 1 provides copies of the McGuire Units 1 and 2 Technical Specifications (and Bases) with the appropriate Unit 2/ Cycle 4 changes indicated. Note that although the changes given in Appendix A of the Unit 2/ Cycle 4 RSE are intended to j apply only to Unit 2, the new Unit 2 AFD limits are identical to the current Unit 1 AFD limits (T.S. Fig. 3.2-la) as previously discussed, and therefore the same technical specification figure can now apply to both units since the McGuire 4

Unit 1 and 2 technical specifications are combined into one document (this has been accounted for in Attachment 1). Consequently, the McGuire Unit 1 specifi-cations are administratively affected in that a specification figure currently applying to McGuire Unit 1 only is changed to be applicable to both Units 1 and 2.

In addition, the Fq changes are applicable to (and are requested to be approved for) both McGuire units. This increased limit is justified by the analysis in 1

Attachment 2B for Unit I as well as Unit 2. Although the increase in the heat flux hot channel factor was assumed in the McGuire 2/ Cycle 4 RSF< the change is i not a consequence of the reload (Unit 2 Cycle 4 could be started with the existing l technical specifications and a 50.59 review, however revisions to surveillance software would be required as well as additional efforts by Duke Power Company j personnel) and therefore is acceptable for innnediate implementation on Unit 1.

(The constants provided in the Peaking Factor Limit Report for McGuire 1 Cycle 4 (current operating cycle) remain applicable for the higher Fq limit.) The McGuire FSAR update will accordingly reflect this change for Unit I as well as Unit 2.

)

l l Attachment 2 Page 4 l I 1 Another proposed technical specification [T.S. 6.9.1.9] change included in I

' Attachment 1 involves changing the schedule for providing the peaking factor limit report from 60 days before use (i.e. criticality) to providing it 30 days after implementation (this change is proposed for both units, and is acceptable for immediate implementation since it is not a consequence of the reload). This change is related to the future feasibility of reload design after unit shutdown (the viability of which hinges on 50.59 reviews or the core limit report concept and removal of the 60 day review requirement on the PFLR (or CLR)), as well as burdens placed on Duke (and Westinghouse) due to time constraints resulting from going to shorter refueling outages than were originally anticipated during plant licensing. A similar technical specification revision for a plant with the radial PFLR (F Spec wa license *lssuanc)e).sCorresponding recently approved provisions by theareNRC (Vogtle made to McGuire's Nuclear Station, initial T.S. 6.9.1.9 (Fq Spec) peaking factor limit report (note that this also corrects an incorrect i

addressee listed in the current McGuire T.S. 6.9.1.9 - NRC reorganization has eliminated the " Core Performance Branch"). Note also that the term " radial" is

, being deleted from the title of T.S. 6.9.1.9 as " radial" only applies for the F l

SpecandwasinadvertentlyretainedwhenMcGuireUnits1and2wenttotheFqShe W(z) functions (reference License Amendments 32 (Unit 1)/13(Unit 2) and 42(Unit 1)-

/23(Unit 2)).

Attachment 2C is the Peaking Factor Limit Report for McGuire Unit 2/ Cycle 4 which is provided in accordance with (the currently applicable) Technical Specification 6.9.1.9, transmitting the W(z) functions to be used for RAOC and Base Load Oper-ations, and the value for APL . For both RAOC and baseload operation, three specific burnup steps are provided which permit the determination of W(z) at any cycle burnup through the use of three point interpolation. The information for baseload operation has been obtained using a + 5 percent AFD about a measured target in the power range 80-100% rated thermal power. Figures 1 through 3 are i

the W(z) functions appropriate for RAOC operation and Figures 4 through 6 are for baseload operation. The appropriate W(z) function is used to confirm that the l

i Heat Flux Hot Channel Factor, Fq(z), will be limited to the values specified in j

, the Technical Specifications (note that the requested Fq limit increase to 2.32 is

reflected in the report). '

i This peaking factor report for Unit 2 Cycle 4 is based upon the proposed revision to the RAOC AFD envelope. If the revised RAOC limits are not approved in time to support the Unit 2 Cycle 4 startup, confirmation of the validity of the W(z)

functions with respect to the existing RAOC AFD envelope or generation of new W(z) functions would be required. An exemption to the (currently applicable technical specification) 60 days prior to criticality submittal schedule for the peaking factor report may be necessary (unless the proposed T.S. 6.9.1.9 revision dis-cussed above is approved in a timely manner) if revised W(z) functions are re-quired. The use of overly conservative W(z) functions increases the probability l of entering the action statement and could lead to very restrictive AFD limits I and/or reductions in reactor power. ,

. i l'

1

ATTACrDIENT 2A RELOAD SAFETY EVALUATION MCGUIRE NUCLEAR STATION UNIT 2 CYCLE 4 January, 1987 Edited by: C. R. Savage G. G. Ament Approved: a ge44 E' A. Dzenis, Mana[r Core Operations Nuclear Fuel Division ssu a .- m m

I TABLE OF CONTENTS Title h 1

1.0 INTRODUCTION

AND

SUMMARY

1 1.1 Introduction 1

1.2 General Description 2

1.3 Conclusions 3

2.0 REACTOR DESIGN 3

2.1 Mechanical Design 4

2.2 Nuclear Design 4

2.3 Thermal and Hydraulic Design 6

3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 6

3.1 Power Capability 6

3.2 Accident Evaluation 7

3.2.1 Kinetic Parameters 7

3.2.2 Control Rod Worths 7

3.2.3 Core Peaking Factors 8

4.0 TECHNICAL SPECIFICATION CHANGES 9

5.0 REFERENCES

APPENDIX A - Technical Specification Page Changes u m e-era m j

r 0

LIST OF TABLES Page Table Title Fuel Assembly Design Parameters 10 1

11 2 Kinetics Characteristics End-of-Cycle Shutdown Requirements and Margins 12 3

Control Rod Ejection Accident Parameters 13 4

l LIST OF FIGURES Page Figure Title Core Loading Pattern 14 1

e seseLLe-stosos jj

Y

1.0 INTRODUCTION

AND

1.1 INTRODUCTION

This report presents an evaluation for McGuire Unit 2, Cycle 4, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was performed utilizing the methodology described in WCAP-9273-A, " Westinghouse Reload Safety Evaluation Methodology.(1) ,

McGuire Unit 2 is operating in Cycle 3 with Westinghouse 17x17 low parasitic (STO) and optimized fuel assemblies (OFA). For Cycle 4 it is planned to refuel the McGuire Unit 2 core with Westinghouse 17x17 optimized fuel assemblies. In the OFA transition licensing submittal (2) to the NRC, approval was requested for the transition from the STD fuel design to the OFA design and the associated proposed changes to the McGuire Units 1 and 2 Technical Specifications. The licensing submittal, which has received NRC approval, justifies the compatibility of the OFA des 1gn with the STD design in a transition core as well as a full 0FA core. The OFA transition licensing submittal (2) contains mechanical, nuclear, thermal-hydraulic, and accident evaluations which are applicable to the Cycle 4 safety evaluation.

All of the accidents comprising the licensing bases which could potentially be affected by the fuel eload have been reviewed for the Cycle 4 design described herein. The results of new analyses and the justification for the applicability of previous analyses are addressed in safety evaluations for the RTD Bypass Elimination.(9) 1.2 GENERAL DESCRIPTION The McGuire Unit 2, Cycle 4' reactor core will be comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in I i

Figure 1. During the Cycle 3/4 refueling, 64 STD fuel assemblies will be i replaced with 64 Region 6 optimized fuel assemblies. A summary of the Cycle 4 l fuel inventory is given in Table 1 as shown on page 10.

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Nominal core design parameters utilized for Cycle 4 are as follows- l 3411' l Core Power (NWt) 2250 i System Pressure (psia)

Core Inlet Temperature (*F) 558.5 382,000 Thermal Design Flow (gpm)

Average Linear Power Density (kw/ft) 5.43 (based on 144" active fuel length)

1.3 CONCLUSION

S From the evaluation presented in this report, it is concluded that the Cycle 4 This design does not cause any of the safety limits to be exceeded.

conclusion is based on the following:

1. Cycle 3 burnup is between 10500 and 11500 MWD /MTU.

2. Cycle 4 burnup is limited to 13500 MWD /MTU including a coastdown.

3. There is adherence to plant operating limitations in the Technical Specifications.

4. The proposed Technical Specification change discussed in Section 4.0 of this report and provided in Append'x A is approved.

5. The large Break LOCA analysis for UHI Removal using the 1981 EM+ BASH is approved.

l I

ssme-uno. 2

V 2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The Region 6 fuel assemblies are Westinghouse OFAs. The mechanical description and justification of their compatibility with the Westinghouse STD design in a transition core is presented in the OFA transition licensing submittal.(2)

Region 6 has a smaller rod plenum spring than was used in previous fuel rugions. This new spring design satisfies a change in the non-operational 6g loading design criterion to "4g axial and 6g lateral loading with dimensional stability." Notification of Westinghouse's plans to generically incorporate this criterion change and the justification of no unreviewed safety questions were previously transmitted to the NRC via Reference 13. The reduced spring force reduces the potential for pellet chipping in the fuel rod.

Table 1 presents a comparison of pertinent design parameters of the various fuel regions. The Region 6 fuel has been designed according to the fuel performance model I4) .

The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse clad flattening model N) .

For all fuel regions, the fuel rod internal pressure design basis, which is discussed and shown' acceptable in Reference 6, is satisfied.

Westinghouse has had considerable experience with Zircaloy clad fuel. This experience is described in Reference 7. Operating experience for Zircaloy grids has also been obtained from six demonstration 17x17 0FAs(2) , four demonstration 14x14 0FAs(2) , three regions of 0FA fuel in the McGuire Unit 1 Cycle 2, 3, and 4 designs, and two regions of 0FA fuel in the McGuire Unit 2 Cycle 2 and 3 designs.

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1 2.2 NUCLEAR DESIGN The Cycle 4 core loading is designed to meet a Fg (z) x P ECCS limit of

< 2.32 x K(z). This increased limit is justified by the analysis in l Reference 12.

l Relaxed Axial Offset Control (RAOC) will be employed in Cycle 4 to enhance operational flexibility during non-steady state operation. The RAOC methodology and application is fully described in Reference 8. The analysis for Cycle 4 includes a wider full power al band than existed in Cycle 3.

The RAOC AI band was widened by utilizing peaking factor margin which had become available since the original McGuire RAOC analysis. No change to the safety parameters is required for Cycle 4 RAOC operation.

Table 2 provides a summary of Cycle 4 kinetics characteristics compared with the current limits based on previously submitted accident analyses.

Table 3 provides the control rod worths and requirements at the most limiting condition during the cycle (end-of-life) for the standard burnable absorber design. The required shutdown margin is based on previously submitted accident analysis. The available shutdown margin exceeds the minimum required.

The loading pattern contains 352 wet annular burnable absorber (WABA)(14) rods located in 52 fuel rod assemblies. Location of the WABA rods are shown in Figure 1.

2.3 THERMAL AND HYDRAULIC DESIGN The thermal hydraulic methodology, ONBR correlation and core DNB limits used )

for Cycle 4, are consistent with the current licensing basis. No significant l variations in thermal margins will result from the Cycle 4 reload. l The thermal-hydraulic methods used to analyze axial power distributions l generated by the RAOC methodology are similar to those used in the Constant is m s-en a 4

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_ m m l Y

I l

l Axial Offset Control (CAOC) methodology. Normal operation power distributions  !

l are evaluated relative to the assumed limiting normal operation power )

distribution used in the accident analysis. Limits on allowable operating l

axial flux difference as a function of power level from these considerations were found to be less restrictive than those resulting from LOCA Fg considerations.

The Condition II analyses were evaluated relative to the axial power distribution assumptions used to generate DNB core limits and resultant No Overtemperature Delta-T setpoints (including the f(aI) function).

changes in the DNB core limits are required for RAOC operation.

ss24as-arc 204 5

I 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability has been evaluated considering the consequences of those incidents examined in the FSAR(3) using the previously accepted design basis. It is concluded that the core reload will not adversely affect the ability to safely operate at the design power level (Section 1.0) during Cycle 4. For the overpower transient, th's fuel centerline temperature limit

! of 4700*F can be accommodated with margin in the Cycle 4 core. The time i dependent densification model(10) was used for fuel temperature i evaluations. The LOCA limit at rated power can be met by maintaining Fg (z) i at or below 2.32 x K(z).

l 3.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the FSAR(3) were examined. In all cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in 1) the previous applicable safety analysis, or 2) the safety evaluation performed in support of the RTD Bypass Elimination licensing submittal I9) .

A core reload can typically affect accident analysis input parameters in the l f

following areas: core kinetic characteristics, control rod worths, and core peaking factors. Cycle 4 parameters in each of these areas were examined as I discussed in the following subsections to ascertain whether new accident analyses were required.

s i

l ssme-en* 6

i I * ,

I I

3.2.1 KINETIC PARAMETERS l

j Table 2 is a summary of the kinetic parameters current limits along with the associated Cycle 4 calculated values. All of the kinetic values fall within l the bounds of the current limits.  !

l 3.2.2 CONTROL R00 WORTHS i

Changes in control ' rod worths may affect differential rod worths, shutdown margin, ejected rod worths, and trip reactivity. Table 2 shows that the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 4 meets the current limit. Table 3 shows that the Cycle 4 shutdown margin requirements have been satisfied. Table 4 is a summary of the current limit control rod ejection analysis parameters and the corresponding Cycle 4 values.

3.2.3 CORE PEAKING FACTORS 1

i Peaking factors for the dropped RCCA incidents were evaluated based on the NRC aporoved dropped rod methodology described in Reference 11. Results show that DNB design basis is met for all dropped rod events initiated from full power.

The peaking factors for steamline break and control rod ejection have been evaluated and are within the bounds of the current limits.

l l

I l

l l

l

. n a .-.,. . 7 l

l

4.0 TECHNICAL SPECIFICATION CHANGES l

l

! To ensure that plant operation is consistent with the design and safety evaluation conclusion statements made in this report and to ensure that these conclusions remain valid, a Technical Specifications change will be needed for l

Cycle 4 to incorporate a wider full power RAOC al band. This change is t

presented in Appendix A.

)

1 I

i

, l u m . -.'" 8

l

5.0 REFERENCES

l

1. Davidson, S. L. (Ed), et. al., " Westinghouse Reload Safety Evaluation Methodology", WCAP-9273-A, July 1985.

5 l

2. Duke Power Company Transmittal to NRC, " Safety Evaluation for McGuire Units 1 and 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies",

December 1983.

3. "McGuire Final Safety Analysis Report."

4. Miller, J.V., (Ed.), " Improved Analytical Model used in Westinghouse Fuel  ;

Rod Design Computations", WCAP-8785, October 1976. )

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

6. Risher, D. H., (et. al.), " Safety Analysis for the Revised Fuel Red Internal Pressure Design Basis," WCAP-8964-A, August 1978.

7. Letter from E. P. Rahe Jr. (Westinghouse) to J. Lyons (NRC),

Subject:

transmittal of " Operational Experience with Westinghouse Cores" (through December 31, 1985) (Non-Proprietary), NS-NRC-86-3184, November 26, 1986.

8. Miller, R. W., (et al.), " Relaxation of Constant Axial Offset Control-Fg Surveillance Technical Specification," WCAP-10217-A, June 1983

9. Westinghouse Transmittal to Duke Power Company, "RTD Bypass Elimination Licensing Report: Final Version," September 1985.

10. Hellman, J.M. (Ed.), " Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.

11. Marita, T., Osborne, M. P., et. al., " Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10297-P-A (Proprietary) and WCAP-10298-A (Non Proprietary), June 1983.

12. Letter DAP-87-516, L. L. Williams (Westinghouse) to K. S. Canady (Duke Power), "McGuire BASH Analysis Submittal", March 1987.

13. Letter from E. P. Rahe, Jr. (Westinghouse) to L. E. Phillips (NRC), April l l

12,1984,NS-EPR-2893,

Subject:

Fuel Handling Load Criteria (6g vs. 4g).

14. Letter from Thomas, C. 0., NRC, to Rahe, E. P., Westinghouse,

Subject:

Acceptance for Referencing of Licensing Topical Report WCAP-10021 (P),

Revision 1, and WCAP-10377 (NP), " Westinghouse Wet Annular Burnable Absorber Evaluation Report," August 9,1983.

ume-enu 9

1 TABLE 1 MCGUIRE UNIT 2 - CYCLE 4 FUEL ASSEMBLY DESIGN PARAMETERS 4* 5* 6A* 6B*

Region 1

$ 3.40 1

2.093 3.206 3.190 3.20  !

Enrichment (w/o U-235)+

Density (% Theoretical)+ 94.77 94.96 95.23 95.0 95.0 9 60 60 44 20 Number of Assemblies 15745# 18363 13606 0 0 Approximate Burnup at++

Beginning of Cycle 4 (MWD /MTU) 27581# 32580 24568 15312 15690 Approximate Burnup at++

End of Cycle 4 (MWD /MTU) l l

• Optimized Fuel - Zire grid

+ All fuel region values are as-built except Region 6 values which are nomin.al. ,,

++ Based on E0C3 = 11000 MWD /MTU, EOC4 = 13500 MWD /MTU (coastdown included)

1. The burnups noted are for the Region i fuel assemblies being used and are not an average for the whole region.

isme- "' 10 f

TABLE 2 MCGUIRE UNIT 2 - CYCLE 4 KINETICS CHARACTERISTICS

. Cycle 4 Current Limits Design f

(

+7 <70% of RTP +7 <70% of RTP Minimum Moderator +7 ramp to 0 from Temperature Coefficient +7 ramp to 0 from 70% to 100% of RTP 70% to 100% of RTP (pcm/*F)*

)

-2.9 to -0.91 -2.9 to -0.91 Doppler Temperature Coefficient (pcm/*F)*

Least Negative Doppler- -9.55 to -6.05 -9.55 to -6.05 Only Power Coefficient, Zero to Full Power, (pcm/% power)*

Most Negative Doppler -19.4 to -12.6 -19.4 to -12.6 Only Power Coefficient, Zero to Full Power (pcm/%

power)*

.44 >.44 Minimum Delayed Neutron Fraction B,ff, (%)

Minimum Delayed Neutron .50 >.50 Fraction B (%)

(Ejected R88f,t a BOL]

100 <100 l Maximum Differential Rod l Worth of Two Banks Moving l

Together (pcm/in)*

-5 l

• pcm = 10 3,

{

l l

I I

si m e-.ra2o. 11

n TABLE 3 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MCGUIRE UNIT 2 - CYCLE 4 Cycle 3 Cycle 4 Control Rod Worth (%ao) 7.27 8.03 All Rods Inserted 6.03 6.85 All Rods Inserted Less Worst Stuck Rod 5.42 6.17 (1) Less 10%

Control Rod Requirements (%ao) 3.06 3.87 Reactivity Defects (Doppler, Tavg, Void, Redistribution) 0.50 0.50 Rod Insertion Allowance 3.56 4.37 (2) Total Requirements Shutdown Margin ((1) - (2)] (%Ao) 1.86 1.80-Recuired Shutdown Margin (%ao) 1.30 1.30 6

9 is m e-ero

• 12 i

TF TABLE 4 MCGUIRE UNIT 2 - CYCLE 4 CONTROL ROD EJECTION ACCIDENT PARAMETERS Current Limit

• Cycle 4 HZP-BOC 0.75 <0.75 Maximum ejected rod worth, %ap 11.0 <11.0 Maximum Fg (ejected)

HFP-BOC 0.23 <0.23 Maximum ejected rod worth, %Ap 4.5 <4.5 Maximum Fg (ejected)

HZP-EOC 0.90 <0.90 Maximum ejected rod worth, %ap 20.0 - <20.0 Maximum Fg (ejected)

HFP-EOC Maximum ejected rod 0.23 <0.23 worth, %ap 5.9 <5.9 Maximum Fg (ejected)

• Based on the safety evaluation performed in support of the RTO Bypass Elimination licensing submittal (9) .

siua.- '" 13

,r L K J H G F E D C 8 A R P N M teo*

5 6A 5 68 5 6A 5 1

5 5 6A 4 6A 4 6A 4 6A 5 l 5 4 4 4 4 5 5 68 4 6A 4 4 4 6A 4 68 5 5 8 8 SS 8 8 5 68 5 5 4 68 4 68 4 5 5 68 5 8 8 8 8 4 5 5 5 6A 1 5 5- 5 4 6A 5 5 6A 1

' 8 4 4

6A 4 SA 4 5 4 6A 4 6A 4 5 4 6A 4 6A 8 8 8 8 4 68 6A 4 4 4 6A 1 68 4 SA 5 5 6A 1 4 8 8 8 8 4 4 4 4 6A 4 4 4 4 6A 4 4 4 68 68 1 270' 8 so* , ,

4 68 6A 4 4 4 6A 1 68 4 6A 5 5 6A 1

' 8 8 4 4 8 -

8 6A 4 6A 4 5 4 6A 4 6A 4 5 4 6A 4 6A 8 8 8 8 5 6A 4 5 5 5 1 6A 1 5 5 5 4 6A 5 4 8 4 ,

5 68 5 5 4 68 4 68 4 5 5 68 5 l 8 8 8 8 5 5 68 4 6A 4 4 4 6A 4 68 5 5

' 8 8 SS 8 8 5 5 6A 4 6A 4 6A 4 6A 5 5 4 4 4 4 5 6A 5 68 5 6A 5 15 o*

l X REGION NUMBER Y BA'S SS SECONDARY SOURCES I

I FIGURE 1 CORE LOADING PATTERN MCGUIRE UNIT 2. CYCLE 4 14

!l Il APPENDIX A TECHNICAL SPECIFICATION PAGE CHANGE Modification to Page:

3/4-2-4 S

$5ML8-870123

- - _ . _ _ . . . . . . . _ . _ _ . . . _ , . , - _ _ _ _ _ _ _ . . , _ . , _ _ _ _ , . _ ,. _. _ . . _ _ _ . _ . . - _ , . . ~ _ . .

120 110 g.30.100) - -

(10.100)

) .

= n=u - nou f

( __

( )

I C- J 80 l ACCEPTABLE j

u 70

( t

- 1 Y **

/

f a

w j o so --

(31, 0) i

(.38.50)

U 40--

30 -

20 -

10 0

20 30 40 M

-50 -40 -30 -20 -10 0 to Axial Flux Difference (% Delta-1)

FIGURE 3.2-1 AFD Limits as a Function of Rated Thermal Power McGuire Unit 2 Cycle 4 i

,s-ATTACHMENT 2B LARGE AND SMALL BREAK LOCA BASH ANALYSIS RESULTS

I' f '

15.6.4.1 Identification of Causes and Accident Descriotion Accentance Criteria and Fianncy Classification ture of the i

A loss-of-coolant accident (LOCA) is the result eofwith a pape rupFor the an total reactor coolant system (RCS) pressure boundary.

here, a major pipe break (large break) is 2 This eventas defined is a ruptur cross-sectional area equal to or greater than 1.0 ft . See Section 15.0.1 considered an ANS Condition IV event, a limiting fault.

for a discussion of Condition IV events. is defined as l

A minor pipe break (small break), as considered initing hthis a total section, a rupture of the reactor coolant pressure boundary level and (Section 5.2 cross-sectional area less than 1.0 ft' in which the normallySeeopera charging pressure.

system flow is not sufficient to sustain pressur Section 15.0.1 for a discussion of Condition 111 ibed inevents.

The Acceptance Criteria for the loss-of-coolant accident is descr 10 CFR 50.46 as follows: h a.

The calculated peak fuel element cladding temperature is bel requirement of 2200*F. the b.

The cladding temperature transient is tensinatedhing. at a time whe core geometry is still amenable to cooling. oxidation t l

i

c. The amount of hydrogen generated by fuel to interaction of 15 of the total amount of Zircaloy in the reac d.

The core remains amenable to cooling during and after the e.

The core temperature is reduced and decay heattivity is removed fo extended period of time, as required by the long lived radioac remaining in the core.

in ECCS These criteria were established to provide significant margin performance following a LOCA.

2 margin to the Acceptance Criteria limits than Descriotion of a Laroe Break LOCA Transient ssure Should a major break occur, depressurization of the RCS A safety injection decrease in the pressurizer. The the pressurizer low pressure trip setpoint is reached.d signal (SIS) is generated when the appropriat 15.6-7 N

e

a. Reactor trip and borated However, no water injection credit is taken in the LOCA complem

, fission product decay heat.

l analysis for boron content of the injection wate l neglected in the large break analysis.

b. Injection of borated water provides for heat transfer from the core l and prevents excessive clad temperatures.

i The sequence of events following a large break LOCA are presented j 15.6.4-1.

\ Before the break occurs, the unit is in an equilibrium During condition, i.e heat generated in the core is being removed of the via th At the beginning continues blowdowntophase, be transferred the entire to the RCS reactor contains coolant.

subcooled liquid which from the core by forced convection with some fully developed nucleate transition boiling and forced convection to stea boiling.

mechanisms.

The heat transfer between the Reactor Coolant System In the and the se may be in either direction depending on the relative temperatures.

l case of continued heat addition to the secondary, secondary system pr increases and the main steam safety valves may actuate to limit the pr j Makeup waterThe to the secondary side is automatically provided by SIS actuates a feedwater isolation signal which feedwater system.

isolates normal feedwater flow by closing the main feedwater isola and also initiates Theemergency feedwater secondary flow aids in flow by starting the reduction the auxiliary of Reactor Coolant

' feedwater pumps.

System pressure, h When the Reactor Coolant system depressurizes to approximately i accumulators begin to inject borated water into the reactor coolant lo I Since the loss of offsite power is assumed,The the reactor coolant pumps effects of pung assumed to trip at the beginning of the accident.

coastdown are included in the blowdown analysis.

The blowdown phase of the transient ends when the RCS pressur

! assumed at 2280 psia) falls to a value approaching that of the co atmosphere.

prior to or at the end of the blowdown, the mechanisms that are responsible for the bypassing of emergency At this time (calledcore endcooling of inject Reft 11 is complete the RCS are calculated not to be effective. t bypass) refill of the reactor vessel lower plenum begins.

when emergency core cooling water has filled the lower plenum of f core vessel, which is bounded by the bottom of the fuel rods (called bottom recovery time).

4 15.6-8 8992Q:10/073185

l I

l The reflood phase of the transient is defined as the time period lasting from l the end Frombeen the later extent that of therefill until the reactor core temperature rise has been vessel has terminated. filled with wat stage of blowdown and thee the beginning of reflood, the safety injec accumulator tanks rapidly discharge borated cooling The downconer water into the RCS contributing to the filling of the reactor vessel downconer.for the reflooding of l Water elevation The headlow provides the high head and driving head force required safety injection pumps l aid in l

I the thereactor fillingcore. of the downconer and subsequently The safety injection supply pumpedwater to mai l

k downconer and complete the reflooding process. flow as a functi cases.

j

< Continued cooling.

operation of the ECCS pumps supplies water After the water level in the associated with dissipation of residual heat.

refueling water storage tank (RWST) reaches a minimum allowable for long-term cooling of the core is obtained by switching tod the cold le recirculation phase of operation in which spilled borated water is draw the containment sump by the low head safety injection (RHR) pumps a to the RCS cold legs. The Containment Spray System continues to o further reduce containment pressure.

of the LOCA, the ECCS is realigned to supply water to the RCS hot i order to control the boric acid concentration in the reactor Descriotion of Small Break t.0CA Transient Ruptures of small cross section will cause expulsion These pumps would maintain of the coolant which can be acconunodated by the charging pumps.

l' an operational water level in the pressurizer permitting the operat l

execute an orderly shutdown. containment contains the fission products e The maximum break size for which the normal makeup system can m ,

l I

pressurizer level is obtained by comparing the calculated flow from Reactor Coolant System through the postulated break against A the char makeup flow at normal Reactor Coolant System pressure, i.e., 2250 p to makeup flow rate from one centrifugal charging pump is typically ade j

' sustain pressurizer level at 2250 psia for a break throu i diameter hole.

Should a larger break occur, depressurization of the Reactor Coolan f causes fluid to flow into the loops from the pressurizer resulting in aR

' pressure and level decrease in the pressurizer. During the earlier part of low pressurizer pressure trip setpoint is reached.

the small break transient, the effect Therefore, upward offlowthe break 3 as they are coasting down following reactor trip.The Safety Injection System is through the core is maintained. The consequences of the accident are the appropriate setpoint is reached.

limited in two ways:

t 1

l l

l 15.6-g  !

Sgg2Q:10/073105 l

l l

1. Reactor . trip and borated water injection complement void formation in the l

core and cause a rapid reduction of nuclear power to a residual level f

corresponding to the delayed fission and fission product decay.

i

2. Injection of borated water ensures sufficient flooding of the core to

' prevent excessive clad temperatures.

8efore the break occurs the plant is in an equilibrium condition, i.e., During the heat generated in the core is being removed via the secondary system.

blowdown, heat from decay, hot internals, and the vessel continues to be transferred to the Reactor Coolant System. The heat transfer between the Reactor Coolant System and the secondary system may be in either direction i depending on the relative temperatures. In the case of continued heat addition to the secondary, system pressure increases and steam dump may occur. Makeup to the secondary The side is automatically provided by the safety injection signal stops normal feedwater ,

auxiliary feedwater pumps.

flow by closing the main feedwater line isolation valves and initiates The secondary j

auxiliary feedwater flow by starting auxiliary feedwater pumps.

flow aids in the reduction of Reactor Coolant System pressures.

When the RCS depressurizes to 600 psia, the cold Dueleg toaccumulators beginpower the loss of offsite to inject water into the reactor coolant loops.

assumption, the reactor coolant pumps are assumed to be tripped at the time of '

reactor trip during the accident and the offacts of pump coastdown are l included in the blowdown analyses. i j

15.6.4.2 Analysis of Effects and Conseauences Methods of Analysis The requirements of an acceptable ECCS Evaluation Model are presented in l The requirements of Appendix K Appendix K of 10 CFR 50 (Reference 3).  !

2 regarding specific model features were met by selecting models made The assumptions which provide a significant overall conservatism in the analysis.

! pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core i

' peaking system.

factors, the containment pressure, and the perform The thermal-hydraulic calculated as required by Appendix K of 10 CFR 50.

i analyses reported in this section were performed with an upper head fluid temperature of Tc old-Larne Break Evaluation Model The analysis of a large break LOCA transient is divided into three phases:

There are three distinct (1) blowdown, transients analyzed in each phase: (2) refill, and (3) reflood.(1) the thermal-hydraulic transient in the i

i RCS, (2) the pressure and temperature transient within th core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis of the LOCA.

l The description of the various aspectsThese of thedocuments LOCA analysis methodology describe the major is ,

given in References 4,10,13 and 14.

phenomena modeled, the interfaces among the computer codes, and the features i

15.6-10 i 89920:10/092685

1 1

I The of the codes which ensure compliance with the Acceptance Criteria.

SATAN-VI (Reference 5), WREFLOOD (Reference 6) LOTIC (Referen (Reference 13), SASH (Reference 14) and LOCTA-IVi (Referenc l to assess the core heat transfer geometry and to determine if the core re i amenable to cooling throughout and subsequent to the blowdown, r The SATAN-VI computer code analyzes theThe tREFL000 reflood phases of the LOCA.

thermal-hydraulic transient in the RCS during blowdown.

computer codes are used The to calculate BART computer the thermal-hydraulic code is used to tran the reflood phase of the accident. calculate the fluid and heat trans The LOTIC computer code is used to calculate theSimilarly, the pressure containment transient during all three phases of the LOCA analysis.

LOCTA-IV computer code Fuel is used to compute parameters input to thethe thermal LOCTA-IV code transien feel rod during the three phases.

were taken from a new version of the PA0 code (Reference g).

SATAN-VI is 1 sed to calculate the RCS pressure, enthalpy, dens energy flow rates, as well as steam generator heat transfer between th l

)

primary and secondary systems, as a function of time during thel of the LOCA. SATAN-VI also calculates the accumulator water m be pressure and the pipe break mass and energy flow rates that are a i vented to the containment during blowdown. The mass and l refill phases, these data are transferred to the WREFL000 code.

energy releaseAdditional ratesSATAN-VI during blowdown and reflo output data from the end of these phases of the LOCA. blowdown and refill, including the core pressu transient, are input to the LOCTA-IV code.

BASH is an integral part of the ECCS evaluation model which provides a realistic thermal-hydraulic simulation of the reactor core and RCS durin l reflood phase of a LOCA. Instantaneous values of accumulator conditions a l

l safety injection flow at the time of completion of lower plenum l provided to BASH by WREFLOOD. substituted for WREFL000 A more in cal l enthalpy, and pressure for the detailed fuel rod model, LOCTA.The BASH l

l detailed description of the 8 ASH code A moreis available dynamic ,

i phenomena in the reactor coolantInsystem the SASH codeduring cor r' l' expected, and recent experiments have borne this out.

reflood model, SART provides the entrainment rate for a given flood l

then a system model determines loop flows and pressure dr the calculated core exitThis flow.system will produce a more dynamic flooding new entrainment transient, which rate. reflects the close coupling between core thermal-h l

and loop behavior.

l The LOTIC code is a mathematical model of the ice condenser contain LOTIC is described in detail in Reference 7.

l SATAN and WREFL000, which provide the necessar i

the containment.

provide containment boundary conditions required by BASH.

l 15.6-11 l Mg20:10/0g2805

l dding and coolant The LOCTA code is a computer program that evaluates fuel, c a temperatures during a LOCA. A more completei description ld a significant than is prese here can be found in Reference 8.uses afficients, code consisting of LOC l' the improvement in fuel rod behavior prediction. SART suploys

rod model, for the calculation of local heat transfer d coe empirical FLECNT correlation is replaced by the SART co e.fficients a rigorous mechanistic models to generate heat transfer l ingcoeby on a the LOCT to rods.

the actual flow and heat transfer regimes experienc static empirical correlation.

h=11 Break LOCA Evaluation Model - f lant accidents The NOTRUMP computer code is used in the analysis i ting ofof a loss-o -coo due to small breaks in the reactor coolant system.

is a state-of-the-art one-dimensional general d network code con number therinal of advanced non-equilibrium features. in all fluid volumes, flow it e level regine- tracking epe calculations with counter-current flooding limitations, m x ur transfer logic in multiple-stacked fluid nodes, and regine-dep to design  ;

l correlations.

(ECCS) evaluation model was developedd Small to ddetermine in the RC I basis small break LOCAs and to address the NRC Plants.'

concerns expres Break NUREG-0611, " Generic Evaluation ofl Feedwater Tl i

3

ths.

In NOTRUMP, the RCS is nodalized into volumes d intointerconnecte a

The brokenThe loop is modeled transient behavior explicitly of thewith the intact system isapplied loops lumpe determined from th t

second loop. l governing conservation equations of mass, energy and momen u throughout the system.

l References 11 and 15. the The use of NOTRUMP in the analysis involves,ith an among other thingI representation of the reactor core as heated control volumes t d w associated bubble rise model to permitInaparticular, transient m calculation.

- detailed spatial representation of various system loss-of-coolant transient. 8) code Cladding thermal analyses are performed withflow the past LOCTA-IV the (R NOTRUMP which uses the RCS pressure, f uel rod power hi hydraulic calculations, as input.

i i .

l i

15.6-12 ,

l Ogg2Q:10/0g2685

) j

,-.nn,.,.m..- .. - - - - -_,.--,w~.__ , , , we,, _,__,-. - ,, n,_-_,-,._._n--a n, _,n_, n n

l 1

f is given in Figure A schematic representation of the computer code inter aces 15.6.4-3. tinghouse ECCS The small break analysis'was performed with the approved Wes l Small treak Evaluation Model (References 8, 11 and 15). ,

i Laree truk Inout Pa. - ters and initial Ca=ditions l conditions used in l

l Table 15.6.4-1 lists important input parameters and initia l the large break analyses.

4 ditions I hall treak Innut Par -ters and Initial Con ditions used in j

Table 15.6.4-1 lists important input parameters and initial con l the small break analyses. the small break ,

The axial power distribution 15.6.4-60 and 15.6.4-61.

and core decay power assumed for l analyses are shown in Figures function of the j Safety injection flow rate to the Reactor Coolant System as aThe Sa the generation system pressure is used as part of the input. system

, of a safety injection signal. flow which is For these analyses, the This figure injection as aSI function delivery of RCSconsiders pressure. pumped ce curves depicted in Figure 15.6.4-62 f represents injection flow from the TheSI25 pumps second delay j tionbased pumpson per includes time onto theorman degraded 5 percent from the design head. require emergency buses.

since their shutoff head is lower than RCS pressure during the transient considered here. Cooling Systemi capability 102% of the and ope The hydraulic analyses are performed with the NOTRUMP c licensed NSSS core power.

. with the LOCTA-IV code using 102% of licensed NSSS core power.

Laroe Break Results 12), the Based on the results of the LOCA sensitivity studies i (Reference k

limiting large break was found to be double-ended cold leg gu (DECL6).

Therefore, only the DECL6 break is considered ECCS performance analysis. Consistent with the methodologyt case for

break discharge coefficients (CO ).

h no failures of the described in minimum safety injection was used in a calculation Reference 16 the break size whichrere in whicThe i

ECCS were assumed 49 (Maximum safeguards).sussaari l the parameters of principal Figures 15.6.4-4 through 15.6.4-F presentTransients oflthedfollowing

)

interest from the large break ECCS analyses. and where parameters are presented for each discharge coefficient ana appropriate for the worst break maximum safeguards case.

i 15.6-13 i

/0g2685 1'

_ -- __._,8_9920.:10_ , - - . - - _ _ - - - . . _ _ _ . _ . . _ _ _ ~ , - . , - _ _ . - - - . - . - - _ _ _ . - ~ . _ .

I l

Figure 15.6.44 through n The following quantities are presented at th l Figure 15.6.4-W '

1 -.

-- rod.....

(hot_ ...,

rod):... a. N l A mass velocity A-heat transfer coef ficient The heat transfer coefficient shown is calculated by the LOCTA-IV code.  !

i st The system pressure shown is the calculated pressure in  ;

Figure 15.6.4-W Core flowrates are also presented. l through 14 the core.

Figure 15.6.4-4ta.  ;

ig

' These figures show the hot spot clad temperature transient l Figure 15.6.4-426 and the clad tesperature transient at the burst location.

through at tesperature shown is also for er 'rt rt' t%ese. ,

Figure 15.6.4-296 The flui -The nodal notation of the figures is defined in locatio i Table 1 .6.5-7.

• i

! 14  ;

These figures show the core reflood transient.

Figure 15.6.4-40' '

through Jg l Figure 15.6.4+ ,

J4 These figures show the cold leg accumulator delivery l Figure 15.6.4-394 during blowdown.

through 36  ;

Figure 15.6.4-44k. J 31 l Figure 15.6.4-4k. The pumped safety injection during reflood and '

through p &n = u.e uttuumaximum and minimum safeguards ccses. I Figure 15.6.4-446 #

18 41 l

The maximum cladding temperature calculated for a large mit of 2200*F of 10 CFR 50.46.

The break is l' is less than the Acceptance Criteria ercent, which is well below the iI by 10 CFR The total core maximum embrittlement locallimitmetal of 17 water percentreaction as required 50.46. d with metal water reaction is less than 0.3 percent for all breaks, as compare the 1 percent criterion of 10 CFR 50.46, and the cladding temperature l to transient is terminated at a time when the core geometry isf still ame cooling. As a result, the core temperature will continue to drop and the a

ability to remove decay heat generated in the fuel for an extended time will be provided.

small treak Results i As noted previously, the calculated peak claddingAtemperature range of re ses11 break LOCA is less than that calculated for a large break. i

, small break analyses are presented which establishes the limiting b I l The results of these analyses arethesummarized present in Tables principal parameters of 15.6.!

Figures 15.6.4-63a through 15.6.4-71 For all cases analyzed, the interest for the small break ECCS analyses.

following transient parameters are included:

a. RCS pressure

b. core mixture height

c. hot spot clad temperature

/

15.6-14  !

920:10/0g2645 (

l For the limiting break analyzed (3 inch), the following additional transient 4

- parameters are presented (Figures 15.6.4-72 through 15.6.4-74):

a. core steam flow rate

b. core heat transfer coefficient ,

c. hot spot fluid temperature The maximum calculated peak cladding temperature for the ses11 breaks analyzed is 1488'F. These results are well below all Acceptance Criteria limits of 10

! CFR 50.46 and no case is limiting when compared to the results presented for large breaks.

Transition Core 1 m et I

The large break loss-of-coolant accidgnt (LOCA) analysis presented herein for McGuire Units 1 and 2 considered a f d1 core of optimized fuel. This is consistent with the methodology employed in the Reference Core Report 17 x 17 Optimized Fuel Assembly (0FA) for 17 x 17 0FA Transition (WCAP-9500).

When assessing the impact of transition cores on large break LOCA analysis, it must be detemined whether the 1,ransition core can have a greater calculated l i

l peak clad temperature (PCT) than either a complete core of the reference design or a couplete core of the new fuel design. For a given peaking factor,

the only mechanism available to cause a transition core to have a greater i calculated PCT than a full core of either fuel is the possibility of flow redistribution due to fuel assembly hydraulic resistence mismatch. This j hydraulic resistance mismatch may exist only for transition cores and is the  !

l j only unique difforence between a complete core of either fuel type and the i

transition core.

The difference in fuel assembly resistance (K/A2 ) for the two assembly designs [17 x 17 Standard /17 x 17 0FA) may impact two portions of the large break LOCA analysis model. One is the reactor coolant system (RCS) blowdown portion of the transient analyzed with the SATAN-VI computer code, where the l 1 higher res'istance 17 x 17 0FA assembly has less cooling flow than the 17 x 17 i standard fuel assembly. While the SATAN-VI code models the crossflows between the average core flow channel (N-1 fuel assemblies) and a hot assembly flow l

channel (one fuel assembly), experience has shown that SATAN-VI results are not significantly affected by small differences in the hydraulic resistance j between these two channels.

To better understand the transition core large break LOCA blowdown transient phenomena, conservative blowdown fuel clad heatup calculations have been performed to determine the clad temperature effect on the new fuel design for mixed core configurations. The effect was determined by reducing the axial flow in the hot assembly at the appropriate elevations to simulate the effects i

of the transition core hydraulic resistance mismatch. In addition, the M J

blowdown evaluation model was modified to account for grid heat transfer 4

enhancement during blowdown for this evaluation. The results of this analysis j have shown that no peak clad temperature penalty is observed during blowdown.

l Therefore, it is not necessary to perform a new blowdown calculation for l t

transition core configurations because the Evaluation Model blowdown

! calculation performed for the full 17 x 17 0FA core is conservative.and i bounding.

1 i

1 89920:10/092685 15.6-15

. . 1 1

1 1

The other portion of the LOCA calculation impacted by hydraulic resistance

- mismatch is the core reflood transient. Fuel assedly design specific analyses have been performed with a version of the SART computer code which accurately models mixed core cases during reflood. Westinghouse transition core designs including specific .14 x 14,15 x 15 and 17 x 17 standard to 0FA transition core cases were analyzed. For each of these cases BART modelled

! both fuel assembly types and predicted the reduction in axial flew at the

appropriate elevations. As expected, the increase in hydraulic resistance i mismatch for the 17 x 17 0FA assembly was shown to produce a reduction in j

reflood steam flow rate for the 17 x 17 0FA assembly at the mixing vane grid elevations during the transition core period. This reduction in steam flow rate is offset by the fuel grid heat transfer enhancement predicted during i - reflood. The various fuel assembly. specific transition core analyses performed resulted in peak clad temperature increases of up to 10*F for core axial elevations where pCTs can possibly occur. Therefore, the maximum PCT penalty possible for 17 x 17 0FA during transition cores is 10*. Once a full core of the 17 x 17 0FA fuel is achieved, the large break LOCA analysis with UNI removed will apply without the crossflow penalty.

1 15.6.4.3 Environmental Consecuences The postulated consequences of a LOCA are calculated for 1) offsite and 2) l control room operators. .

Offsite Dose Consecuences .

The offsite radiological consequences of a LOCA are calculated based on the j following assumptions and parameters.  ;

1. 100 percent of the core noble gases and 25 percent of the core iodines are

! released to the containment atmosphere.

i j 2. 50 percent of the core iodines are released to the containment.

3. Annulus activity which is exhausted prior to the time at which the annulus l reaches a negative pressure of -0.25 in, w.g. is unfiltered.

4. ECCS leakage begins at the earliest possible time sung recirculation can 3

begin.

i 5. ECCS leakage occurs at twice the maximum operational leakage. ,.

1

) 6. Bypass leakage is 7 percent of total containment leakage.

\ '

! 7. The offactive annulus volume is 50 percent of the actual volume.

i 8. The annulus filters become faulted at g00 seconds resulting in a 15 percent reduction in flow.

{

g. Elemental iodine removal by the ice condenser begins at 600 seconds and continues for 2540 seconds with.a removal efficiency of 30 percent.

l l

1 Sgg20:10/0g2685 15.6-16

l

10. One of the containment air return fans is assumed to fail.

11. The containment leak rate is fifty percent of the Technical Specification i

limit after 1 day.

I

12. Iodine partition factor for ECCS leakage is 0.1 for the course of the accident.

13. No credit is taken for the auxiliary building filters for ECCS leakage.

14. The redundant hydrogen recombiners and ignitors fail. Therefore, purges are required for hydrogen control.

15. The annulus reaches equilibrium after 200,000 seconds such that the only discharge is due to inleakage.

16. Water density at 160'F is used to calculate the sump water mass.

17. Other assumptions are listed in Table 15.6.4-10.

I Based on the model in Appendix 15A, the thyroid and whole body doses are The l

calculated at the exclusion area boundary and the low population zone.  :

' doses are presented in Table 15.6.5-10 and are within the limits of 10 CFR 100.

Control Room Operator Oose The maximum postulated dose to a control room operator is determined based on the releases of a Design Basis Accident. In addition to the parameters and assumptions listed above, the following apply:

1. The control room pressurization rate is 1,000 cfa; the filtered recirculation rate is 1,000 cfm.

j

2. The unfiltered inleakage into the control room is 10 cfs.

3. Other assumptions are listed in Table 15.6.4-11.

l 15.6.5 A NUMBER OF BWR TRANSIENTS Not applicable to McGuire.

l

! 15.6-11

! 89920:10/091285

i REFERENCES FOR SECTION 15.6

1. Burnett, T. W. T., et. al., 'LOFTRAN Code Description", WCAP-7907, June 1972.

Sharp, D. R., " Improved Thermal Design

2. Chelemer H., Soman, L. H.,

Procedures", WCAP-8587, July 1975.

3. " Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors",10 CFR 50.46 and Appendix K of 10 CFR

50. Federal Register, Volume 39, Number 3. January 4,1974.

" Westinghouse ECCS

4. Bordelon, F. M., Massie, H. W. and Borden, T. A.,

Evaluation Model-Susumary", WCAP-8339, (Non-Proprietary), July 1974. i Comprehensive Space Time

5. Bordelon, F. M., et. al.,

• SATAN-VI Program:

Dependent Analysis of Loss of Coolant", WCAP-8302, (Proprietary) June '

1974, and WCAP-8303, (Non-Proprietary), June 1974.

6. Kelly, R. D., et. al., " Calculated Model for Core Reflooding After a loss of Coolant Accident (WREFLOOD Code", WCAP-8170 (Proprietary) and l WCAP-8171 (Non-Proprietary), June 1974

7. Hsieh, T. and Raymond, M., "Long-Tern Ice Condenser Containment LOTIC Code Supplement l', WCAP-8355 Supplement 1 May 1975, WCAP-8354 (Proprietary), July 1974.

8. Sordelon, F. M. , et. al . , 'LOCTA-IV Program: Loss of Coolant Transient Analysis", WCAP-8301 (Proprietary) and WCAP-8305, (Non-Proprietary),

June 1974.

O., U.S.N.R.C., Letter

9. Rahe, E. P., Westinghouse letter to Thomas, C.

Number NS-EPR-2673, October 27,1982,

Subject:

' Westinghouse Revised PAD Code Thermal Safety Model", WCAP-8720, Addendum 2 (Proprietary).

10. Eiche1dinger, C., " Westinghouse ECCS Evaluation Model, February, 1978 Version", WCAP-9220 (Proprietary) February, 1979, and WCAP-9221  ;

(Non-Proprietary) February,1978. l

11. Lee, H., Tauche, W. D., Schwarz, W. R., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10081-A, August 1985.

l 12. Salvatori,.R., " Westinghouse Emergency Core Cooling System - Plant Sensitivity Studies", WCAP-8340, (Proprietary) July 1974. l 1

I

13. Young, M., et. al., '8 ART-1A: A Computer Code for the Best Estimate Analyzed Reflood Transients", WCAP-9561-P-A,1984 (Westinghouse i Proprietary).

i I

t 15.6-18 89920:10/092685

l l

An Integrated Core and RCS Reflood Code A 3t g g Ie

14. Kabadi, J. N., et. al.

• BASH:for Analysis of PWR Loss-of-Co 3 (Westinghouse Proprietary). ll Break

'NOTRUNP, A Nodal Transient sma J..

15. Neyer. P. E. and Kornfilt, and General Network Code". WCAP-10000-A, August 1985.

RC).

E. P. (Westinghouse). letter to Tedesco. R. L. (USN l

16. Rahe No. NS-EPR-2538, December 1981.

l i

l l

l I

I l

15.6-19 10/092685

.._ 09920:------- - ___.___. _ _ _. "*W'= mwn,_-,p_ _

TA8LE 15.6.4-1 Innut Parameters Used in the ECCS Analyses Large Break Small Break Parameter 12.88 12.21 Peak Linear Power (kw/ft)

(includes 1025 factor) 2.32 2.32 Total Peaking Factor, Fq Chopped See Figure Power Shape Cosine 15.6.4-60 Fuel Assembly Array 17 X 17 17 X 17 Optimized Optimized Nominal Cold leg Accumulator 950 950 Water Volume (f t3/ accumulator)

Nominal Cold Leg Accumulator 1350 1350 Tank Volume (ft 3/ accumulator)

Minimum Cold Leg Accumulator 600 600 Gas Pressure (psia)

Pumped Safety Injection Flow See Table See Figure 15.6.4-6 15.6.4-62 Steam Generator Initial Pressure (psia) 987.0 987.0 5 5 Steam Generator Tube Plugging Level (5) l f

L 89920 10/091285 15.6-20

l TABLE 15.6.4-2 l Laree Break LOCA Time Seauence of Events C 0.8 C 0.6 CD - 0.4 DD =ECLG DD =ECLG DECLG (sec) _

(sec) __( sec) 0.0 0.0 0.0 Start 0.46 0.46 0.47 Reactor Trip Signal 2.6 2.7 2.9 Safety Injection Signal 12.7 15.3 21.0 Cold Leg Accumulator Injection 27.6 27.7 27.9 Pump Injection 33.1 44.5 End of Bypass 28.42 34.8 44.6 26.6 End of 8 lowdown 4s.5 St. I G 4.4 44h9- -44

• Bottom of Core Recovery 4'r:$-

6 (,. 9 -70.6 U.0  ;

.34 4 ,IS,4-~

Cold leg Accumulator Empty 43.re-l l

l l

l l

l 15.6-21 89920:10/092685

l i

TA8LE 15.6.4-3 Laree Break LOCA Time Seauence of Events Maximum Safeguards C0 = 0.6 DECLE fsec) 0.0 Start 0.46 Reactor Trip Signal 2.7 Safety Injection Signal 15.3 Cold Leg Accumulator Injection 27.7 Pump Injection 33.1 End of Bypass 34.8 End of 810wdown 50.3 49 Bottom of Core Recovery 7Y 7

.26.6 Cold Leg Accumulator Empty I

i f

1 1

l 15.6-22 09920:10/092645

TABLE 15.6.4-4 Laree Break LOCA Results

_ Fuel Claddine Data 0.8 C 0.6 C 0.4 C

DD =ECLG DD =ECLE DO =ECLG 1941 1614 1795

+94* M62 Mit Peak Clad Temperature (*F) 9.ao 6.25 6.25 4rM-Peak Clad Temperature Location (ft) 2.'1 I l. l(o 2 46 +:44-l 4,48 6r4-l Local Ir/H 20 Reaction (max). (5) 7.oo 7.oo 7.o o W +ree 4rM-Local Ir/H2O Location (ft)

<0.3 <0.3

<0.3 Total Ir/H 2O Reaction, (5) 59,1 ff7 -M ri-Hot Rod Surst Time, (sec) 4riht-6.00 6.00 i Hot Rod Burst Location. (ft) 1 i

I l

l i

6 _

15.6-23

TA8LE 15.6.4-5 Larce Break LOCA Results Fuel Claddina Data Maximum Safeguards i

C 0.6 DOE

= ctg  ;

RESULTS /803 N

Peak Clad Temperature (*F)

Too M

Peak Clad Temperature Location (ft) 7. 18 M

Local Ir/H 2O Reaction (max), (5) 7 ao ,

4:00-Local Ir/H2 0 Location (ft)

<0.3 Total Ir/H 2O Reaction. (5) 57 . 5 Hr+

Hot Rod Burst Time (sec) 6.00 Hot Rod Burst Location (ft) l l

! l I

1 l

l l '

15.6-24 89920:10/092645

. . - _ , . _ . _ _ . -- - _ . - . _ _ - - - - . _ - - - . _ . . _ . - - - . - . .~.. --_--__ - . .

l I .

TABLE 15.6.4-6 Safety Iniection PW Flow Assumed for treaks Greater Than or Equal to 10 inches MINIMUM SAFEGUAROS: 51 Flow (1b/sec)

Pressure insia) 493.2 437 3 14.7 318.6 34.7 315.5 54.7~ 198 4 14.1 1M.2 114.7 81.2 214.7 58.5 614.7 0.0 1014.7 3014.7 MAXIMUP. SAFEGUARDS: SI Flow

{1b/sec)

Pressure fosiaL_ 1078 997 14.7 914 34.7 826 1

54.7 731 74.7 621 94.7 171 114.7 166 214.7 l i

314.7 l

l l

l 15.6-25 1

- - - - - ~ - . .

._ .__ 3992g:10/092685. - - . . - . - - - _ _ . _ _ _ . . , . _

! i l

TABLE 15.6.4-7 11 areak i e 4 T h t :::::e of twents 2 in 3 in 4 in 6 in 11851 13R11. 138$1 18811 0.0 0.0 0.0 'O.o start ,

I 23.58 19.24 9.09 6.35 1 Reactor Trip N/A 1026.5 681.9 289.2 Top of Core uncovered .

3 Cold Les Accumulator Injection N/A N/A 991.9 393.2 Peak Clad Temperature Occurs N/A 1737.6 982.0 453.5 i

N/A >2400 1642.1 503.2

' Top of Core Covered l

)

1 1

,l I

l l l 1

I I

l i

1 1

i i .

l

( .

\

) .

i I

89920:19/092885 15.6-26

t TABLE 15.6.4-6

~

l * - 11 Breat tara Results fpelCladdianBata Lit 1 11 1.2L iit RESULTS Peak Clad Temperature (*F) N/A 1488 1348 1189 N/A 12 12 12 Peak Clad Location (ft)

N/A 0.81 0.10 0.078 Local Ir/H 2O Reaction (max). (5) ,

N/A 12 12 12 Local Ir/H 2O Reaction Location Ft.

N/A <0.3 4.3 <0.3 Total Ir/N 20 Reaction. (5)

N/A N/A N/A N/A Not Rod Surst Time. (sec)

N/A N/A N/A N/A Not Rod Burst Location. (ft) ,

l l

l n

I senannaminass_ is.s-n

i No Chage Fros *

. 1984 FSAR Update

( .

Table 15.6.4-11 (Page 1 of 2)

Parameters for LOCA Offsite Dose Analysis

. 1. Data and assumptions used to estimate '

radioactive source from postulated i

accidents

a. Power Level (ltft) 3565

b. Failed fuel 100% of fuel rods in core

c. Activity released to containment stao-sphere from failed fuel and available for release (percent of core activity)

Noble gases 100 Iodines 25

d. Iodine activity released to containment 50 sump and available for release via ECCS leakage outside contaiment t (percent of core activity)

e. Iodine fractions (organic, elemental, Regulatory Guids 1.4 and particulate) *

! 2. . Data and assumptions used to estimate act-ivity released

a. Containment free volume i Upper containment volume (ft2) 6.70E+05 Lower containment volume (ft3 ) 3.68E+05 Total containment free volume (ft2) ,. 1.038E+06

b. Containment leak rate (percent of contaidment volume per day) ,

0< t < 24 hrs 0.2 t > 24 hrs >

0.1

c. Bypass leakage fraction 0.07

d. Annulus ventilation iodine filter efficiency (percent) 2, 95 15.6-28 198t. Update

++-+*---e- y e +p-w,- , e .- ,- ++

l

! No Change Fros

~

1984 FSAR Update Table 15.6.4-11 (Page 2 of 2) .

. Parameters for LOCA Offsite Oose Analysis

e. Total annulus ventilation flow rate (cfe) 0-900 sec . 8000 900+ sec 6800

f. Time at which annulus ventilation system is operable (sec) 23

g. Annulus volume (ft3)  ; 422,361

3. Dispersion data .

I

a. Distance to exclusion area boundary (m) 762

b. Distance to low population zone (a) 8850

c. X/Q at exclusion area boundary (sec/m3) .

0-2 hrs 9.0E-04

d. X/Q at low population zone (sec/m3) 0-8 hrs 8.0E-05 8-24 hrs 5.2E-06 1-4' days 1.7E-06 l 4+ days ,

3.7E-07

4. Dose data

a. Method of dose c'alculations Appendix 15A I b .* Dose conversion assumptions ' Appendix 15A

c. Doses (Rem)

Case 1 (With ECCS leakage)

Exclusion Area Boundary Whole Body 3.0 Thyroid 2.0E+02 Low Population Zone whole Body 6.2E-01 Thyroid . 6.5E+01 Case 2 (Without ECCS leakage)

Exclusion Area Boundary Whole Body 2.9 Thyroid 1.8E+02 Low Population Zone Whole Body .

6.1E-01

')

'/

Thyroid 5.7E+01 15.6-29 MHuroTo

l No Change From

.: 1984 FS M Update Table 15.6.4-12 (Page 1 of 2)

Parameters for LOCA Control Room Dose Analysis l

1. Data and assumptions used to estimate l radioactive source from postulated l accidents j

a. Power Level (mit) 3565 l

b. Failed fuel 100% of fuel rods in core

c. Activity released to containment atmos-phare from failed fuel and available for l

release (percent of core activity)

Noble gases 100 Iodines 50 ,

d. Iodine activity released to contaisent e- 50 sump and available for release via ECCS leakage outside containment (percent of core activity)

e. Iodine fractions (organic, elemental, Regulatory Guide 1.4 C - and particulate)

2. Data and assumptions used to estimate act-ivity. released ,

a. Containment free volume (ft3) 1.038E+06

b. Containment leak rate (percent of containment volume per day)

Octc24 hrs 0.2 tI21 hrs 0.1

. c. 8ypass l'eakage fraction 0.07

d. Control room pressurization rate (cfm) 1000

e. Control room filtered recirculation rata (cfm) 1000

f. Control room unfiltered:in-leakage (cfs) 10

g. Control room volume (ft3) 116,000

h. Control room pressurization and re- ". 99 f circulation removal efficiencies for

(, . iodine (percent) ,

15.6-30 nervurnwn

a

.' No Change From 1984 FSAR Update Table 15.6.4-12 (Page 2 of 2) -

Parameters for LOCA Control Room Dose Analysis

3. Dispersion data - Control room int'a'ke X/Q (sec/m3) 0-8 hrs 1.0E-03 8-24 hrs 7.0E ~1-4 days 4.5E-04 4+ days 2.4E-04

4. Dose data  :

a. Method of dose calculations Appendix 15A i 1

b. Dose conversion assumptions Appendix 15A l 1

c. Ooses (Rem)

Whole body -

2.2E-01 ,

Thyroid 2.6E+01 1 Skin 4.1 l

l l

l I

. 1

< s 15.6-31 i e

.- 1

. l

.. _ _ _ . .. 14715.64 e

, jL M MlCL#tS NACTOR TRIP (COhPENSATED PMSSURIZER PNSSUE) SIGNAL, l l

g PLAFED SAFETY INKCTION SIGNAL (HI-l CONT. PRESS OR LO PNSSURIZER P)

L PLAFED SAFETY INKCTION BEGINS ( ASSUMING OFFSITE POWER AVAILABLE)

O w

COLD LES arviu a ATOR INKCTION w

N CONT. BEAT PEMOVAL SYSTEM INITIATION ( ASSLAdING OFF5ITE POWER AVAILABLE) l 3r PtArED SarETY INKCTION BEGINS ( ASSLAdING LOSS OF OFF5ITE POWER) oc OF eLOWOOWN REFILL 3r BOTTOM OF CORE NCOVERY a

l 7 COLD LES ACr1u a aTORS EMPTY L

0

• O D

CORE OUDOED JL O SWITCH TO COLD LEG RECIRCULATION ON RWST LOW LEVEL ALARM g (MANUAL ACTION)

'O T

SWITCH TO LONG-TERM RECIRCULATION (MANUAL ACTION)

E R

M C

O O

L -

I N

G \

l 1

y l Figure 15.6.4-1: Sequence of Events for Large Break Lossof Coolant Analysis l

l I

3- - -

h e*

14715.27 G

. 1 2

f

! .I N

_ E m

)

tK

> _ 1m 8_ g

~

_ g a

=

n ,

i

, 1 1,

, Hi i ,

i i a_ .

a_ a (Visci) sanssswa saw

2 .. .

50 40 -

6- '

k.

d in so w

E g 20 -

' w E

o O

10 -

o 1 I I I I I I 0 250 500 750 1000 1250 1500 1750 2000 TIhE (SEC)

Figure 15.6.4-67

5 4" Cold Leg Break Core Mixture Height vs. Time I I I

. e 14715.3 l

l i

3000 d

' 2500 -

V1 w

• 2000 -

F 1500 -

S .

W

>- 1000 -

O l

C

< 500 - .

d -

I O

500 750 1000 1250 1500 TIME (SEC)

Figure 15.6.4 68: 4" Cold Log Brook Hot Spot Clad Temperature vs. Time e, . , , ,.,, . . -. , .,-.,.-.-.-g- , . - - g - . . ._ _ , . , . , - - - - ,

I.

14715.30 a

a j 8 8 n e ,

l l

8- O M

g a:

g l

1 i

b .E Y a l mo 8, F 5

S g- ~

! 1

- 8 I

I l I O

l l _

(yIsci) sunss3Wd sou

. , , eoe e CORE MIXTURE LEVEL (PT) o 5 5 8 $ E o i a i i

1

= 8 -

E a

G -

in t

w

?

O no I

E -

e H 8 E

E ml c ,

e

• n 1 E

i

- 8 1 5 l 3 -

l 1 a -

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= l l

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• l

_ , . I

i i

. 14715.32 i

l 1

i 3000 j I

l

- l k.

sn 2500- -

h 2000 - i 1

1 H 1500 -

I

k. '

i S '

w M, 1000 - f o

0 4 500 -

d -

O I I I I l l

! 200 300 400 500 600 700 300 300 TIME ISEC) 1 l

l l

l Figure 15.6.471: 6" Cold Leg Break Hot Spot Clad Temperature vs. Time

l .

1 CORE STEAM FLOW (L5/SEC) l i

o .

O

, i i . I I I m

~

,W

+

0 O

~

r- -4 3 *

?.

E k "

n $

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s a .

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~

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< 5 -

% g-h, -

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y 'a =

5  :

F

< 2 -

Y I I I I I l l l l l 8000 l100 3200 1300 1400 1500 1600 1700 1900 1900 2000 .

TIME (SEC) l i

2 l

Figure 15.6.4-73: 3" Cold Leg Break Core Heat Transfer Coefficient vs. Time E j l

, I

)

, 14718.00 2000 1750 -

1500 -

k.

m w-1250 -

s 7 1000 -

b o 750 -

o d

500 -

250 -

I I I I 0

o 500 1000 1500 2000 2500 TIME (SEC) 1 ,

a Figure 15.6.4 74: 3" Cold Log Break Hot Spot Fluid Temperature vs. Time i

l

_a g 4

4 O

i #

I 9

14715.9 l

l D

R O

v cJ Q

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w R 6, m

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l 1944 Update 1

ATTACHMENT 2C PEAKING FACTOR LIMIT REPORT FOR McGUIRE UNIT 2 CYCLE 4

PEAKING FACTOR LIMIT REPORT FOR MCGUIRE UNIT 2 CYCLE 4 RAOC AND BASE IDAD OPERATION This Peaking Factor Limit Report is provided in accordance with Paragraph 6.9.1.9 of the McCuire Unit 2 Technical Specifications.

The McGuire Unit 2 Cycle 4 elevation dependent W(z) values for RAOC operation at'beginning, middle, and near end-of-life are shown in Figures 1 through 3, respectively. This information is sufficient to determine W(z) versus core height for Cycle 4 burnups in the range of 0 MWD /MTU to 13500 MWD /MTU through the use of three point interpolation.

The McGuire Unit 2 Cycle 4 elevation dependent W(z) values for base load operation between 80% and 100% of rated thermal power with a 15 percent AFD about a measured target value at 150. 6000, and 12500 .

1 MWD /MTU Cycle 4 burnups are shown in Figures 4 through 6, respectively.

This information is sufficient to determine W(z) versus core height for

Cycle 4 burnups in the range of 0 MWD /MTU to 13500 MWD /MTU through the use of three point interpolation.

W(z) values for RAOC and base load operation vere calculated using the method described in Part B of Reference 1.

" The minimum allowable power level for base load ops. ration, APLF , for McCuire 2 Cycle 4 is 80 percent of rated thermal pc er.

The appropriate W(z) function is used to confirm that the heat flux hot channel factor, Fq(z), will be limited to the Technical Specification values of:

A IA (K(z)] for P > 0.50 and pq ,) p Fq(z) $ 4.64 (K(z)} for P $ 0.50 i

i The appropriate elevation dependent W(z) values, when applied to a ,

l power distribution measured under equilibrium conditions, demonstrates that the initial conditions assumed in the LOCA are met, along with the j f

ECCS acceptance criteria of 10CFR50.46.

(1) WCAP 10216 P A, Relaxation of Constant Axial Control Fq ,

l Surveillance Technical Specification l

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' e'~ 'm

• 0.8 1.1805

'

• 7.0 1.1808

7.2 1.1885 7.4 1.1834 t.ts i 7.5 1.1752

• 7.5 1.1970 I i w 5 m 8.0 1.1983 4  !' 8.2 1.1473

" 8.4 1.1340 8.0 1.1192 8.8 1.1078 S.0 1.1121 I O.2 1.1277 tes ' 9.4 1.1417

! 8.8 1.1812 l

9.8 1.1500

' 10.0 1.1718

• e a se :: 10.2 1.1833 e

l iM a

BOTTOM e

CORE E IGHT (FEET) Top e 10.8 22 1.0000 e 11.0 1.0000 i

e 11.2 1.0000 e 11.4 1.0000

• 11.8 1.0000 e 11.s 1.0000 FIGURE 1 e 13.0 1.0000 ,

4 MCGulRE UNIT 2 CYCLE 4 RAOC W(Z) AT 150 MWD /MTU f

(OP ANO SOTTOM 15% EXCLUDED AS PER TECH SPEC 4.2.2.2.0

ISIS 4T ME (FT.) 1NZ)

• 0.0 1.0000 e 0.2 1.0000 W e 0.4 1.0000 e 0.8 1.0000 e 0.8 1.0000 a 1.0 1.0000

• 1.2 1.0000 La ,

e 1.4 1.0000 e 1.8 1.0000

• "

  e e 1.8 1.0000 e ,

2.0 1.2501 2.2 1.2484

• 2.4 1.2315

o 2.8 1.2180 2.8 1.2072

3.0 1.2034 3.2 1.2044

• e 9 3.4 1.2007 1.2* o 3.8 1.2118

• 3.8 1.2170

-- - e 4.0 1.2109 "o*" o..o 4.2 1.2210

" 4.4 1.2103 e ,e 4.0 1.2153 t.m 4.8 1.2104 S.O 1.2194

^

S.2 1.2204 w 1.2294 S.4 St S.S 1.2438

" 8.8 1.2300 0.0 1.2707 S.2 1.2858 S.4 1.2971 S.S 1.3083 f.t2 S.8 1.3108 7.0 1.3121 7.2 1.3101 7.4 1.3038 7.8 1.2932 7.8 1.2830 8.0 1.2727 S.2 1.2297 ,

8.4 1.2483 1 S.S 1.2229 8.8 1.2206 W S.O 1.2144 j I I I I- 3.2 1.2120 l 1 3.4 1.2003 l t 8.8 1.2148 l S.S 1.2283 I ' I I 10.0 1.2379 2 4 s e is is 10.2 1.2403 e

BOTTOM CORE E IGHT (FEET) Top l,,, l," .

e 10.8 1.0000 e 11.0 1.0000 e 11.2 1.0000 e 11.4 1.0000 e 11.0 1.0000 e 11.8 1.0000 FIGURE 2 e it.o 1. coco i MCGUIRE UNIT 2 CYCLE 4 i

RAOC W(Z) AT 6000 MWD /MTU I TOP AND SOTTOM 15% EXCLUDED AS PER TECH SPEC 4.2.2.2.0 1 1

1 M I eff EOL (F7.) W(2) e 0.0 1.0000 e 0.2 1.0000

" e 0.4 1.0000 e 0.0 1.0000 e 0.8 1.0000 e 1.0 1.0000 a 1.2 1.0000 1.a - e e 1.4 1.0000 e 1.8 1.0000 o e e 1.8 1.0000 2.0 1.2744 2.2 1.2001

• 2.4 1.2458 s.s 2.0 1.2321 o a

• 2.8 1.2183

- e 3.0 1.2002 3.2 1.1811

• 3.4 1.2017

• 3.8 1.2130

, ,,e . , ' ' ,

3.8 1.2204

- 4.0 1.2270 4.2 1.2318 s 1.2347 4.4

• 4.8 1.2385

• m 4.8 1.2329 m S.0 1.2312 N S.2 1.2381 4

v 5.4 1.2901 B S.0 1.2001 S.8 1.2890 S.0 1.3031 S.2 1.3182 S.4 1.3300 S.S 1.3382 i" 8.5 1.3425 7.0 1.3427 1.3389 7.2 7.5 1.3171 -

7.8 1.3038 l

.. 8.0 1.2883 )

8.2 1.2718 I 8.4 1.2575 8.5 1.2448 8.8 1.2288

'** 8.0 1.2198 9.2 1.2187 9.4 1.2207 s.S 1.23ss

--' O.8 1.2332 10.0 1.2410 t.m e u 10.2 1.2488 a 2 4 s e CORE HEIGHT (FEET) TOP e 10.8 1.0000 l BOTTOM e 10.8 1.0000 e 11.0 1.0000 e 11.2 1.0000 I e 11.4 1.0000 l

• 11.8 1.0000 l

• 11.8 1.0000 i FIGURE 3

• 12.0 1.0000 l MCGUIRE UNIT 2 CYCLE 4 l RAOC W(Z) AT 12500 MWD /MTU 1

TOP AND 80T70M 15% EXCLUDED AS PER TECH SPEC 4.2.2.2.G l

I 1

DEIGif SOL i (FT.) W(Z) 1 e 0.0 1.0000 e 0.2 1.0000 s.u a 0.4 1.0000

• 0.8 1.0000 e 0.8 1.0000 e 1.0 1.0000 ,

e 1.2 1.0000 I e 1.4 1.0000 e 1.8 1.0000 e 1.8 1.0000 2.0 1.0008 2.2 1.08S3 a 2.4 1.0848 o 2.8 1.0838

'# ,o,

• 2.8 1.0828

. ee 3.0 1.0012

• 3.2 1.0804

'8 = 3.4 1.0873 1.0854

'e 3.0 o 3.8 1.0839 e ,

4.0 1.0829 4.2 1.0817

• 4.4 1.0003

• 4.8 1.0787

'o 4.8 1.0788

^ ' ' S.0 1.0748 5.2 1.0727 U *a 5.4 1.0702 3: ,

• 5.0 1.0879 o S.8 1.0847 e o S.O 1.0812 S.2 1.0544

• 8.4 1.0884

^ S.S 1.0823 S.8 1.0881

.e

' 7.0 1.0729 7.2 1.0774 7.4 1.0818 1 as 7.8 1.0854 m a' 7.8 1.0848 8.0 1.0818 8.2 1.0843 8.4 1.0882 8.8 1.0878 8.8 1.0884 S.0 1.0888 l

S.2 1.0081

}

8.4 1.0072 0.5 1.0879 I 9.8 1.0892

' I 10.0 1.1010

,, e u 10.2 1.1024 e a e e a e 10.4 1.0000 C0RE EIGHT (FE1) 70p e 20.0 1.0000 BOTTOM e 10.8 1.0000 e 11.0 1.0000 e 11.2 1.0000

• 11.4 1.0000 e 11.8 1.0000 e 11.8 1.0000 FIGURE 4 e 12,0 1,0000 MCGUIRE UNIT 2 CYCLE 4 BASELOAD W(Z) FOR POWERS BETWEEN 80% AND 100% OF RATED THERMAL POWER WITHIN +5% AFD OF THE MEASURED TARGET 150 MWD /MTU TOP AND 80TTOM 1S*4 EXCLUDED AS PER TECH SPEC 4.2.2.4.0

i

. o OEIGff ISM.

(FT.) W(Z) e 0.0 1.0000 e 0.2 1.0000

• e 0.4 1.0000 e 0.8 1.0000 e 0.8 1.0000 e 1.0 1.0000 e 1.2 1.0000 e,s e 1.4 1.0000 e 1.0 1.0000 e 1.8 1.0000 2.0 1.1125 2.2 1.1008 f.s 2.4 1.1043 2.0 1.0000 2.8 1.0043 3.0 1.0888 fN 3.2 1.0838 3.4 1.0785 3.8 1.0774 3.8 1.W794 4.0 1.0754 f.et 4.2 1.0742

~

4.4 1.0728

.,no 4.5 1.0712 1.0084

,e-- -- 4.8 l n ,

e- S.0 1.0873 N o S.2 1.0000 v "8 -

5.4 1.0028 w

m e 5.0 1.0508 e

n

.. S.8 1.0000

' o- S.0 1.0588

• 'n. -t S.2 1.0MO

" e .'

S.4 1.0715

'e ' S.S 1.0700

".

• 8.8 1.0798

' 7.0 1.0828

• '.' 7.2 1.0840 7.4 1.0000 7.5 1.0008 7.8 1.0840 3.0 1.0084 S.2 1.1015 8.4 1.1041 8.0 1.1000 8.8 1.1072 0.0 1.1078 S.2 1.1078 0.4 1.1078 9.9 1.1085 9.5 1.1007 10.0 1.1088 s.ge e 4 e e e a 10.2 1.1101 e

00RE E C M (FEE 1) 7ap e ,, ,4 1.0,00,,0 1,0. ,,,

BOTTOM e 10.8 1.0000 e 11.0 1.0000 e 11.2 1.0000 e 11.4 1.0000

• 11.8 1.0000 e 11.8 FIGURE 5 e 33,o 1,.0000

, coco MCGUIRE UNIT 2 CYCLE 4 I

BASELOAD W(Z) FOR POWERS BETWEEN 80% AND 100% OF RATED THERMAL POWER WITHIN +5% AFD OF THE MEASURED TARGET 6000 MWD /MTU TOP ANO BOTTOM 15% EXCLUDED AS PER TECH SPEC 4.2.240

.__ _ _ _a _ _ - -

i l

M 1 9ff EDL l (PT.) W(Z) l

• 0.0 1.0000 e 0.2 1.0000 t.m e 0.4 1.0000

• 0.8 1.0000

• 9.8 1.0000 e 1.0 1.0000

• 1.2 1.0000 I s.s e 1.4 1.0000 >

• 1.8 1.0000 l e 1.8 1.0000 2.0 1.1308 2.2 1.1273 1.88 2.4 1.1179 2.0 1.1000 2.8 1.0079 i 3.0 1.0870 3.2 1.0011 i

t.e4 3.4 1.0828 7". 3.8 1.0848 a a 3.8 1.0051 T "

4.0 1.0057 4.2 1.0080 i t.it

• e a 4.4 1.0000 l neocenai>* 4.0 1.0837 '

'e' 4.8 1.0838 i 1.0843

^ 5.0 N

w ,, ,

. 5.2 1.0013 i e 5.4 1.1008 1 Et

- 5.0 1.1008 ]

5.8 1.1172

.a4'ooe _

' O.0 1.1237 i t.as

^' B.2 1.1?l80 l 8.4 1.1327 ,

S.S 1.1300 )

0.8 1.1357 7.0 1.1350 s.m 7.2 1.1327 )

7.4 1.1292 7.0 1.1243

' 7.8 1.1100 8.0 1.1108 .

t.es 8.2 1.1008 8.4 1.1114

! 8.0 1.1134 .

8.8 1.1143 I

' O.0 1.1145 1.s ' O.2 1.1142 t 9.4 1.1135 S.S 1.1130 i

. 0.8 1.1130

' ' 10.0 1.1141 (

t.m w 10.2 1.1147 )

9 8 4 e a it BOTTOM CORE E IGHT (FEET) ' yop l M 12 e 10.8 1.0000 e 11.0 1.0000 e 11.2 1.0000 l e 11.4 1.0000 l e 11.0 1.0000 e 11.8 1.0000 FIGURE 6 e 12.0 1.0000 MCGUIRE UNIT 2 CYCLE 4 l BASELOAD W(Z) FOR POWERS BETWEEN 80% AND 100% OF RATED THERMAL POWER WITHIN +5% AFD OF THE MEASURED TARGET 12500 MWD /MTU I

TOP AND BOTTOM 18% EXCLUDED AS PER TECH SPEC 4.2.2.4.G 1

1

ATTACHMENT 3 ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION

l l

! ATTACEMENT 3 i ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION l

I As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendments involve ~significant hazards considerations, as defined by i 10 CFR 50.92. Standards for determination that a proposed amendment involves no 4

significant hasards considerations are if operation of the facility in accordance l j with the proposed amendment would not: 1) involve a significant increase in the 1 i

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.

l I. McGuire 2/ Cycle 4 Reload Related Technical Specification Changes:

i The proposed amendments to incorporate a wider full power RAOC delta I band and j BASH Code LOCA analysis results are consistent with the design and safety eval-

] ustion conclusion statements made in the McGuire Unit 2 Cycle 4 reload safety i evaluation. The reference safety evaluation report submitted by Mr. H.B. Tucker's i November 14, 1983 letter to Mr. H.R. Denton summarised the evaluation performed on

{ the region-by-region reload transition from the McGuire Units 1 and 2 standard

{ (STD) fueled cores to cores with all optimised fuel (OFA). The report examined i

the differences between the Westinghouse STD design and 0FA design and evaluated i the effects of these differences for the transition to an all 0FA core. The

! report (approved by the NRC) justifies the compatibility of the OFA design with 1 j the STD design in a transition core as well as a full 0FA core, and contains i summaries of the mechanical, nuclear, thermal-hydraulic, and accident evaluations I

which along with subsequent modifications to the McGuire licensing bases, are i applicable to the Cycle 4 safety evaluation. Cycle specific reload safety eval-l untions verify that applicable safety limits are satisfied based on the reference j 4

evaluation / analyses established in the reference report.

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j The McGuire Unit 2/ Cycle 4 reload safety evaluation (Attachment 2A) presents an {

evaluation which demonstrates that the core reload will not adversely affect the safety of the plant. All of the accidents comprising the licensing bases which could potentially be affected by the fuel reload were reviewed for the Unit 2

Cycle 4 design. The results of new enalyses and the justification for the ap-plicability of previous analyses is presented in the cycle specific reload safety j evaluation. The results of these evaluation / analysis and tests lead to the j following conclusions:

! a. The Westinghouse OFA reload fuel assemblies for McGuire 1 and 2 are mechanically compatible with the STD design, control rods, and reactor internals interfaces. Both fuel assemblies satisfy the design bases for the McGuire units.

j b. Changes in the nuclear characteristics due to the transition from STD to i

0FA fuel will be within the range normally seen from cycle to cycle due to fuel management effects.

j c. The reload 0FAs are hydraulically compatible with the STD design.

i

! d. The accident analyses for the OFA transition core were shown to provide l acceptable results by meeting the applicable criteria, such as, minimum

! DNBR, peak pressure, and peak clad temperature, as required. The j previously reviewed and licensed safety limits are met.

J~

Attachment 3 Page 2 b

e. Plant operating limitations given in the Technical Specifications affected by the reload will be satisfied with the proposed changes.

From these evaluations, it is concluded that the Unit 2 Cycle 4 design does not cause the previously acceptable safety limits to be exceeded. Further, the reload ,

and associated changes to Unit 2's operating limitations have no effects not  !

previously evaluated and approved on accident causal mechanissa or probabilities.

i The commission has provided examples of amendments likely to involve no signi-ficant hazards considerations (48 FR 14870). One example of this type is (vi), "A change which either may result in some increase to the probability or consequences of a previously analysed accident or may reduce in some way a safety margin, but

where results of the change are clearly within all acceptable criteria with i

respect to the system or component specified in the standard review plan: for

, example, a change resulting from the application of a small refinement of a

previously used calculational model or design method". Because the evaluations l previously discussed show that all of the accidents comprising the licensing bases which could potentially be affected by the fuel reload were reviewed for the ,

i Unit 2 Cycle 4 design and conclude that the reload design does not cause the previously acceptable safety limits to be exceeded, the above example can be

) applied to this situation. In addition, the NRC has previously issued no sig-l l nificant hazards consideration determinations for similar McGuire Unit 2 (and 1) i reload amendments. Consequently, example (iii) which states "For a nuclear power i

1 reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with i

the technical specifications and regulations are not significantly changed, and l that NRC has previously found such methods acceptable.", also applies.

l Another example of actions not likely to involve a significant hazards consi-

! deration is (1), "A purely administrative change to technical specifications: For example, a change to achieve consistently throughout the technical specifications, i correction of an error, or a change in nomenclature". Accordingly the changes to i Unit I specifications which do not change the content for Unit 1 but which eli-

! minate certain specific distinctions between units within the common document are l administrative in nature and involve no significant hazards considerations.

4 l

i II. Unit 1 Increased Heat Flux Hot Channel Factor j

)

The proposed amendments would allow an increase in the heat flux hot channel factor (Fq(s)) from 2.26 to 2.32 on McGuire Unit 1 (as well as Unit 2). Analysis 4

with the NRC approved Westinghouse 1981 evaluation model in conjunction with the 1

BASH Reflood Code has been completed for McGuire Unit 1 (and Unit 2) to verify l cceeptable performance of the ECCS system. The supporting documentation is l provided in Attachment 2B.

I 4

I l

I

Attachment 3 Page 3 i

Since the results of the analysis show that the proposed change can be accom-p11shed with conformance to the requirements of 10CFR 50.46 (Acceptance Criteria i for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors), and the change is'to an operating limitation involving no changes to hardware or other i

accident causal mechanisms and can have no effect on accident probabilities, the three standards for determination that no significant hazards considerations are involved are met. In addition, example (vi) cited in Part I above (which covered, among other changes, an identical change on Unit 2) is also applicable to this Unit 1 change.

{ III. Peaking Factor Limit Report Schedule and Title Change (Units I and 2) i The proposed amendments would change the schedule for providing the peaking factor l limit report (for both units) from 60 days before use (i.e. criticality) to l 30 days after implementation. This is a change to an administrative requirement and has no impact on operating limitations or accident causal mechanisms or J probabilities. Further, a similar change has been recently approved for a plant j (Vogtle Nuclear Station) with a radial PFLR (F Spec . The deletion of the term

" radial" from the title of T.S. 6.9.1.9 is purely an a)dministrative change cor-recting the title to reflect the fact that McGuire Units 1 and 2 now use the Fq Spec W(z) functions (for which the term " radial" is not appropriate) instead of the previously used F Sp PartIabovealsoapplies.ec. For this aspect of the change example (i) cited in Based upon the preceding analyses, Duke Power Company concludes that the proposed l

amendments do not involve a significant hazards consideration.

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- - . - - - . - - - - - _ . ._ . _ . - - . _ - . - - - .