Text

Rev 4 to Millstone Unit 2,Cycle 11 Core Operating Limits Rept
	ML20091A548
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Site:	Millstone
Issue date:	03/23/1992
From:	; NORTHEAST NUCLEAR ENERGY CO.
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	NUDOCS 9203270273
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Docket No. 50-136 Millstone Unit No. 2 Cycle 11 Core Operating Limits Report Revision 4 9203270273 920323 PDR ADOCK 05000336 Y

PDR

Page 1 1.

ColtE Ol'EHATING 1.ISilTS 1(EPoltT This Core Operating Limits Report for hfillstone 2 has been prepared in accordance with the requirements of Technical Speelfication 6.9.1.7. The Technical Specifications af fected by this report are listed below:

&cilen Speelficatien 2.1 3/4.1.1.1 SilVTDOWN h1ARGIN - T.,, > 200*P 2.2 3/4.1.1.1 SilVTDOWN h1ARGIN - T.,, s 200'F 2.3 3/4.1.1.4 hioderator Temperature Coefficient 2.4 3/4.1.3.6 Regulating CEA inserticri Limits 2J 3/4.2.1 Linear lleat Rate 2.6 3/4.'.3 Total integrated Radial Peaking Factor - P/T 2.7 3/4.2.6 DND hlargin Terms appearing in capitalized type are DEFINED TERhlS as defined in Section 1.0 of the Technical Specifications.

2.

OPERATING 1.lMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are pre 3ented in the following subsections. These limits have been developed using the NRC-approved methodologies speelfied in Section 3.

2.1 SliUTDOWN h1ARGIN - T.,, > 200*F (Specification 3/4.1.1.1)

The hilUTDOWN hiARGIN shall be 2 3.6% AK/K 2.2 SilUTDOWN h1ARGIN - T.,, s 200'F (Speelfication 3/4.1.1.2)

The SilUTDOWN h1ARGIN shall be 2 2.0% AK/K 2.3 N1oderator Temperature Coeffielent (Speelfication 3/4.1.1.4)

The moderator temperature coefficient shall be:

Less positive than 0.7 x 10~' AK/K/'F whenever TilERh1AL POWER a.

is s 70% of RATED TilERh1AL POWER, b.

Less positive than 0.4 x 10-4 3K/K/'F whenever TilERh1AL POWER

' is > 70% of RATED TilERhIAL POWER, Less negative than -2.8 x-10-* SR/K/'F at RATED TilERh1AL c.

POWER, m

Page 2 j

2.4 Regulating CEA insertion Limits (Specification 3/4.1.3.6)

The regulating CEA groups shall be limited to the withdrawal sequence and 1

to the insertion limits shown in Fl ure 2.4-1. CEA insertion between the f

Long Term Steady State Insertion Lin,s and the Transient insertion Limla I

is restricted to:

a.

s 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, b.

s 5 Effective Full Power Days per 30 Effective Full Power Day 4

interval, and s 14 Effective Full Power Days per calendar year.

c.

2.5 Linear llent Rate (Specification 3/4.2.1)

The Linear heat rate, including heat generated in the fuel, clad and f

moderator, shall not exceed:-

15.1 kw/ft when the reactor coolant flow rate measured per a.

Specification 4.2.6.12 340,000 gpm.

b.

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

I!uring operation with the linear heat rate being monitored by the Excore l

Detector Monitoring System, the AXIAL SHAPE 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 'he following factors are appropriately included in the set"ng 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 enginec-ing uncertainty factor of 1.03, 4.*

A linear heat rate uncertainty factor of 1.01 due to axial fuel dersification and thermal expansion, and 5.

A TilERhlAL POWER measurement uncertainty factor of 1.02.

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

t i

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4 - -.,,., -.. - -. - -, -._. m.,,~

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4 l'agt 3 2.6 Total Integrated Radial Peal.ing Factor - F T (Speelfication 3'l.2.3)

The calculated value of F,T, defined as 12,T - F,(1+T ), shall be s 1.64, y

2.6.1 11 P,T > 1.64, then REFER to the ACTION statement for Speelfication ?.2.3. Refer to Figure 2.6-i for the power dependent F,' limits.

2.7 DNU hialgin (Speelfication 3/4.2.6)

The DNil margin shall be preserved by maintaining the cold leg temperature, pressuriter pressun, reactor coolant flow rate, and ANIAL SilAl'li INDEN within the following limits, l'aimuttet th11113 Four Reactor Coolant Pumns Opciallons_

a.

Cold Leg Temperature s$49'F b.

Pressurizer Pressure 2 2225 psia

e.

Reactor Coolant Flow Rate 2 325,000 ppm d.

ANIAL SilAPl! INDEX FIGURE 2,7-1 Limit not applicable during either a TilERh1AL POWiiR ramp increase in e.seess of SG of RATED TilERh1AL POWER per sninute or a TilERhlAL POWER step increase of creater than 10% of RATED TiiERh1AL POWER.

4 e

s Page 4 3.

ANAIXIICAL METilODS 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 I 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 Pr. : urized Water Peactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, dated May 1988.

3.3 XN-NF-82-21(A), Rev.1. " Application of Exxon Nuclear Company PWR Thermal Margin hiethodology to Mixed Core Configurations." Exxon Nuclear Company, dated September 1983, 3.4 ANF-84-93(A), Rev. O and Supplement 1 "Steamline Ilreak hiethodology for PWR's," Advanced Nuclear Fuels Corporation. Rev. O dated March 1989, Supplement I dated March 1989.

3.5 XN-75-32(A), Supplements 1, 2, 3 and 4, Computational Procedure for y

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 hlodel," Advanced Nuclear Fuels Corporation, both reports dated April 1989.

3.7 EXEh! PWR Large Break LOCA Evaluation h1odel as defined by:

XN-NF-82-20(A), Rev. I and Supplements 1 through 4, " Exxon a.

Nuclear Company Evaluation Model EXEht/PWR ECCS Model Updates," Exxon Nuclear Company. All reports dated January 1990.

b.

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

XN-NP-81-58(A), Rev. 2 and Supplements l through 4, "RODEX2 c.

Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company. Rev. 2 and Supplements 1 and 2 dated March 1984, Supplements e and 4 dated June 1990.

l Page 5 i

d.

XN-NF-85-16(A), Volume 1 through Supplement 3; Volume 2, llev.1 l

and Supplement 1, "PWit 17x17 Puel Cooling Tests Progsam," !!xxon l

Nuclear Company. All reports dated February 1990.

NN-NP-85-105(A), Rev. O and Supplement 1 " Scaling of FCTF e.

liased Reflood lleat Transfer Correlation for Other Bundle Designs,"

Exxon Nuclear Company._ Both reports dated January 1990.

3.8 XN-NP-78-44(A), "A Generic Analysis of the Control itod Ejection l

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

r 3.9 NN-NF-621(A), Rev.1 " Exxon Nuclear DND Correlation of PWR Puel Desig1," Exxon Nuclear Company, dated September 1983.

An accer'able hlllistone 2 specific application of these analytical methodologies is i

described in ANP-88-126. "h1111 stone Unit 2 Cycle 10 Safety Analysis Report,"

dated October 1988, f

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0.2 0.4 0.6 Axlal Shapo Index FIGURE 2.7-1 AXIAL SHAPE INDEX Operating Limits with Four Reactor Coolant Pumps Operating MILLSTONE - UNIT 2 March,1992

. _.

ML20091A548

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