ML20077H042

Rev 2 to Millstone Unit 2 Cycle 11 Core Operating Limits Rept

ML20077H042
Person / Time
Site:	Millstone
Issue date:	06/30/1991
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20077H041	List: ML20077H039 ML20077H042
References
NUDOCS 9107050032
Download: ML20077H042 (10)

Text

.-

e Docket No. 50-336 hiillstone Unit No. 2 Cycle 11 Core Operating Limits Report Revision 2 l

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June.1991

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1. CORE OPERATING LIMITS REPORT This Core Operating Limits Report for h1illstone 2 has been prepared in accordance with the requirements of Technical Specification 6.9.1.7. The Technical Specifications affected by this report are listed below:

Section Specification 2.1 3/4.1.1.1 SilUTDOWN h1ARGIN - T ,, > 200*F 2.2 3/4.1.1.2 SHUTDOWN h1ARGIN - T ,, s 200'F 2.3 3/4.1.1.4 hloderator Temperature Coefficient 2.4 3/4.1.3.6 Regulating CEA Insertion Limits 2.5 3/4.2.1 Linear ileat Rate 2.6 3/4.2.3 Total Integrated Radial Peaking Factor - Ff 2.7 3/4.2.6 DNB hlargin Terms appearing in capitalized type are DEFINED TERN 1S as defined in Section 1.0 of the Technical Specifications.

2. OPERATING LIMITS l The cycle-specific parameter limits for the specifications listed in Section 1.0 are I

presented in the following subsections. These limits have been developed using the NRC-approved methodologies specified in Section 3.

2.1 SHUTDOWN N1ARGIN - T ,g > 200*F (Specification 3/4.1.1.1)

The SHUTDOWN NIARGIN shall be 2 3.6% AK/K 2.2 SHUTDOWN NIARGIN - T.,3 s 200'F (Specification 3/4.1.1.2) l The SHUTDOWN hlARGIN shall be 2 2.0% AK/K 2.3 hloderator Temperature Coefficient (Specification 3/4.1.1.4) l The moderator temperature coefficient shall be:

I a. Less positive than 0.7 x 10" aK/K/'F whenever THERMAL POWER is s 70% of RATED THERhlAL POWER,

b. Less positive than 0.4 x 10" AK/K/*F whenever TilERMAL POWER l

is > 709 of RATED TilERNIAL POWER, l

c. Less negr.tive than -2.8 x 10 4 aK/K/ F at RATED THERNIAl.

POWE R.

June.1991

, Page 2 2A Regulating CEA insertion I imits (Specification 3/4.1.3.6)

The regulating CEA groups shall be limited to the withdrawal sequence and to the insertion limits shown in Figure 2.4-1. CEA insertion between the Long Term Steady State Insertion Limits and the Transient insertion Limits is restricted to:

a. s 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval, b, s 5 Effective Full Power Days per 30 Effective Full Power Day interval, and

c. s 14 Effective Full Power Days per calendar year.

2.5 Linear Heat Rate (Specification 3/4.2.1)

The Linear heat rate, including heat generated in the fuel, clad and moderator, shall not exceed:

a. 15.1 kw/ft when the reactor coolant flow rate measured per Specification 4.2.6.1 2 340,000 gpm.

b. 14.5 kw/ft when the reactor coolant flow rate measured per Specification 4.2.6.1 2 325,000 gpm and < 340,000 ppm.

During operation with the linear heat rate being monitored by the Excore Detector hlonitoring System, the ANIAL SilAPE INDEX shall remain within the limits of Figure 2.5-1.

During operation with the linear heat rate being monitored by the incore Detector Monitor System, the alarm setpoints shall be adjusted to less than or equal to the limit when the following factors are appropriately included in the setting of the alarms:

1.* Flux peaking augmentation factors as shown in Figure 2.5-2,

2. A measurement-calculational uncertainty factor of 1.07,

3. An engineering uncertainty factor of 1.03, 4.* A linear heat rate uncertainty factor of 1.01 dix to axial fuel densification and thermal expansion, and

5. A TIIERMAL POWER measurement uncertainty factor of 1.02.

• These factors are only appropriate to fuel batches "A" through "L"

.lu ne. 1991

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i Page 3 l .

2.6 Total Integrated Radial Peaking Factor Ff (Specification 3/4.2.3)

The calculated value of Ff, defined as Ff = Fr(1+T q), shall be limited to:

a. 0.90 < PF s 1.00 Ff s 1.790 - (0.15 x PF)

b. 0.80 < PF s 030 Ff s 1.925 - (0.30 x PF)

c. 0.70 < PF s 0.80 Ff s 2.205 - (0.65 x PF)

d. PF s 0.70 FI s 1.750 where, PF = TiiERhtAL POWER divided by RATED TilERh1AL POWER 2.7 DNB hlargin (Specification 3/4.2.6)

The DNB margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, and AXIAL SilAPE INDEX within the following limits:

ParametcI Limili Four Reactor Coolant Pumps Operations

a. Cold Leg Temperature s 549'E

b. Pressurizer Pressure 2 2225 psia *

c. Reactor Coolant Flow Rate 2 325.000 gpm

d. AXIAL SilAPE INDEX FIGURE 2.7-1

• Limit not applicable during either a TilERhlAL POWER ramp increase in excess of 59 of RATED TliERhlAL POWER per minute or a TilERhlAL POWER step increase of greater than 10% of RATED TilERhlAL POWER.

June.1991

Page 4

3. ANAIX1'ICAL hils!'llODS The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

3.1 XN-75-27(A), Rev. O and Supplements 1 though 5, " Exxon Nuclear Neutronics Design hiethods for Pressurized Water Reactors," Exxon Nuclear Company. Rev. O dated June 1975, Supplement 1 dated September 1976, Supplement 2 dated December 1980, Supplement 3 dated September 1981, Supplement 4 dated December 1986, Supplement 5 dated February 1987.

3.2 ANF-84-73(P), Rev. 3, " Advanced Nuclear Fuels hiethodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, dated hlay 1988.

3.3 XN-NF-82-21(A), Rev.1, " Application of Exxon Nuclear Company PWR Thermal hlargin hiethodology to hlixed Core Configurations." Exxon Nuclear Company, dated September 1983.

3.4 ANF-84-93(A), Rev. O and Supplement 1 "Steamline Break hiethodology for PWR's," Advanced Nuclear Fuels Corporation. Rev. O dated hlarch 1989, Supplement i dated hlarch 1989.

3.5 XN-75-32(A), Supplements 1, 2, 3, and 4, " Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, dated October 1983.

3.6 XN-NF-82-49(A), Rev.1 and Supplement 1, " Exxon Nuclear Company Evaluation Model EXEh! PWR Small Break Model," Advanced Nuclear Fuels Corporation, both reports dated April 1989.

3.7 EXEM PWR Large Break LOCA Evaluation N1odel as defined by:

a. XN-NF-82-20(A), Rev.1 and Supplements 1 through 4, " Exxon Nuclear Company Evaluation N1odel EXEhl/PWR ECCS hiodel Updates," Exxon Nuclear Company. All reports dated January 1990.

b. XN-NF-82-07(A), Rev.1 " Exxon Nuclear Company ECCS Cladding Snelling and Rupture hlodel.' Exxon Nuclear Company. dated November 1982.

c. XN-NF-81-58(A), Rev. 2 and Supplements 1 through 4, "RODEX2 Fuel Rod Thermal-hlechanical Response Evaluation Alodel," Exxon Nuclear Company. Rev. 2 and Supplements 1 and 2 dated N1 arch 1984, Supplements 3 and 4 dated June 1990.

June. loul

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d. NN-NF-85-16(A). Volume 1 through Supplement 3; Volume 2, Rev.1 and Supplement 1, "PWR 17x17 Fuel Cooling Tests Program," Exxon -

Nuclear Company. All reports dated February 1990.

e .' -NN-NF-85-105(A), Rev, 0 and Supplement 1, " Scaling of FCTF Based Reflood lleat Transfer Correlation for Other Bundle Designs,"

Exxon Nuclear Company. Both reports dated January 1990.

3.8 XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection '

Transient for Pressurized Water Reactors," Exxon Nuclear Company,. dated October 1983.

3.9 XN-NF-621(A), Rev,1 " Exxon Nuclear DNB Correlation of PWR Fuel Design," Exxon Nuclear Company, dated September 1983.

An acceptable h1illstone 2 specific application of these analytical methodologies is described in ANF-88-126, "htillstone Unit 2 Cycle 10 Safety Analysis Report,"

dated October 1988.

June.1<i91

(1.00, BANK 7 La ' 160 STEPS) I g 1.00 h

m H 0.90 A O

Z .. .. ... . ... ... ... .. .. . .. .. . . .. .. . .. .. . _. . . . .. .. . .. .. .. ... ... . ..

(0.80. BANK 7 d4 160 Steps) -

6 8

0.80 - (0.80, BANK 7 & 108 Steps)

C g j i i i i e Z -- --

'g-1 (0.73, B ANK 7 @ 97 Steps) - -- -- -- - -- -* * - - --- - - - - -- --- -- --- --- -- --

r fo 0.70 l - I o- . . -l. .. .. .. ..

N. . g. (0.65 BANK 6 G 135 Steps) ... . .. .. . . .. .. . .. ... ... .. ... ... . ..

4 E l E 0.60 I I t

N  !

g E

.y.

l .. \.. 0.56. B ANk 6 J4 90 Steps)

I

.[. . . . .. _. .. .. .. .. .._ _. .. ..

! 0.50 3 l l 5 . .. .I . . . .. . .. .. .. .. .. .. .. . .. _. .. .. ..L.. . .. ... .. .. ..

' O J J

c 0.40 1Er g wez t ww t- mz - TRANSIENT INSERTION LIMIT l i .9 o I S >. O o . .I . .. ..

g.. . . . __ _ . ..

...f . .. . . .. .. .. . .. .. .. . ..

0,30 I or w r m \ i

? I- m *~w E w

.. l+.*m- . .. . .

I

. . , . . .\_ . .. . . . .. .. . . ... .. ... .. . .

8 ' '

0.20 I N(0.20, iBANK 5 @ 72 Steps)

N.I..

l 0.10 \  ;

% ..l. .. .. ... ... .. .. ..

0 (.. (0.0. BANK 4 @ 72 Steps) f 7 l 5 3 I

BANKS 180 W $8 72  % 0 m W W 72  % 0 40 W 108 72  % 0

]

k- 'l 5 6 4  ;

e h 180 144 108 72 36 0 180 144 108 72 36 0 -

8 CEA Position (Steps Withdrawn) [

t FIGURE 2.4-1 CEA Insertion Um; vs. THERMAL POWER with Four Reactor Coolant Pumps Operating i

-g 130 - '-

i ' i=

r-CD

-1 120 -

O i Z rT1 UNACCEPTABLE OPERATING REGION 8

110 -

C 5

m 100 -

B D 7

l 0 90 -

5 ACCEPTABLE OPERATING

, $ REGION

' 80 -

l E

$ Break Point =

R 70 -

A. (-0.30, 65)

A B. (-0.04, 100)

C, ( O.00, 102) E 60 - D. t 0.08, 200)

• E. ( 0.30, 65) 50 -

Note: Point "C" is provided for setting plant instrumentation nojy. Operation is not

,. 40 - a!! owed above 100% of rated power.

4 30 -  :: , , , , ,  ::

-0.4 -0.3 -0.2 -0.1 -0.0 0.1 0.2 0.3 0.4 Axial Shape Index -

c_

C

, ~3

.O FIGURE 2.5-1 AXIAL SHAPE'INDEX vs. PERCENT OF ALLOWABLE POWER LEVEL o

Q J

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

..)

i 1.0000 .

5 r

r in

-4 O

z -

m 1.0500 l

C Z

R to en a 1.0400 O

u.

z O

G y 1.0300 0

a 1.0200 l

1.0100 l

I 80 100 120 140 0 20 40 60 c_

C

=

-CD DISTANCE FROM BOTTOM OF CORE. INCHES

=

$

• FIGURE 2.5-2 Augmentation Factor vs. Distance from Bottom of Core &* 3 I

(only appi: cable to fuel batches - A~ through "L") s

.:. =

Page o a

1.2 1.1 1.0 (-0.10,1.0) - - (0.15,1.0) cc w

3 UNACCEPTABLE UNACCEPTABLE a

OPERATION REGION

- k-\ -

OPERATION REGION ~

0.9 2

a .. . .. . .. .. .. .. . . . .. .. . . . . . . ..

w x

s O

g .. .. . .. . .. . .. .. . . . .. . .. . . .. .. .. .. ..

• 0.8 ( -0.30, 0.80) ACCEPTABLE -- (0.30. 0.80) u- OPERATION O .. . ... . .. .. .. .. . . - -- -- -- -- -- -- -- -

REGION O

P O

E u

0.7 0,6 - -~ - - ~ -

0.5 -- -- --- - --- - - - -- '- -

0.4

-0.6 - 0.4 -0.2 0 0.2 0.4 0.6 Axial Shape Index FIGURE 2.7-1 AXIAL SHAPF INDEX Operating Limits with Four Reactor Coolant rumps t )pmating MILLSTONE - UNIT 2 June.1991 J