ML20085L112

Rev 3 to Millstone Unit 2,Cycle 11 Core Operating Limits Rept

ML20085L112
Person / Time
Site:	Millstone
Issue date:	10/31/1991
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20085L105	List: ML20085L102 ML20085L112
References
NUDOCS 9111010220
Download: ML20085L112 (10)

Text

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Docket No. 50-336 i

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i hillistone Unit No. 2  !

Cycle 11 .

Core Operating Limits Report ,

Revision 3 6

e 5

l October,1991 9111010220 911023 PDR ,

P ADOCK 05000336-PDR

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1. CollE OPI'ItATING 1.1511TS REPOl(T This Core Operating Limits lleport for hlillstone 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:

Sellien Specificatkn 2.1 3/4.1.1.1 SilUTDOWN hlARGIN - T , > 200*F 2.2 .3/4.1.1.2 SilUTDOWN h1ARGIN - T , s 200'F 2.3 3/4.1.1.4 h1oderator Temperature Coeffielent 2.4 3/4.1.3.6 Regulating CEA Insertion Limits 2.5 3/4.2.1 Linear lleat Rate 2.6 3/4.2.3 Total Integrated Radial Peaking Factor - Ff 2.7 3/4.2.6 DNU hiargin Terms appearing in capitalized type are DEFINED TERhtS as defined in Section 1.0 of the Technical Specifications.

2. OPERATING LIhllTS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been d'veloped using the NRC-approved methodologies specified in Section 3.

2.1 S11UTDOWN h1ARGIN - T.,, > 200*F (Specification 3/4.1.1.1)

The SilUTDOWN h1ARGIN shell be 2 3.6G AK/K 2.2 SilUTDOWN hlARGIN - T.,, s 200*F (Specitication 3/4.1.1.2)

The SilUTDOWN h1ARGIN shall be 2 2.0% AK/K 2.3 hloderator Temperature Coefficient (Specification 3/4.1.1.4)

The moderator temperature coefficient shall be:

a. Less positive than 0.7 x 10-4 AK/K!*F whenever TilERh1AL 10WER is s 70% of RATED TilERh1AL POWER,

b. Less positive than 0.4 x 10-4 AK/K/'F whenever TilERh1AL POWER-is > 70% of RATED TIlERh1AL POWER,

c. Less negative than -2.8 x 10" AK/K/*F at RATED TilERhlAL POWER.

October,1991

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. Page 2 24 Regulating CIM insertion 1.imits (Specification 3!.l.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 insestion 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 1leat 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-i 340,000 ppm.-

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 gpm.

During operation with the linear heat rate being monitored by the Excore Detector hionitoring System, the AXlAL 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 h1onitor 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 due to axial fuel densification and thermal expansion, and

5. A TilERh1AL POWER measurement uncertainty factor of 1.02.

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

October,1991 '

. P.ge 3 2.6 Total Integrated Itadial l'eaking Factor - li (Specification 3/4.2.3)

The calculated value of li, defined as li - li,(1+T ),y shall be limited to:

a. 0.90 c PP s 1.00 li s 1.790 - (0.15 x l'F)

b. 0.80 < PP s 0.90 li s 1.925 - (0.30 x PF)

e. 0.70 < l'F s 0.80 li s 2.205 - (0.65 x PF)

d. PF s 0.70 li s 1.750 -

where.

PP = TilERh1AL POWER divided by RATED TilERhiAL POWi!R 2.7 DNil h1argin (Specification 3/4.2.6)

The DNil_ margin shall be preserved by maintaining the cold leg temperature, pressurizer pressure, reactor coolant flow rate, and AXIAL.

SilAPE INDEX within the following limits:

Parameter Limits Four Reactor Coolant

. Pumps _D.priations

a. Cold Leg Temperature s 549'F

b. Pressurizer Pressure 2 2225 psia *

c. Reactor Coolant Flow Rate 2 325,000 ppm

d. AXlAL SilAPE INDEX FIGURE 2.7-1

• Limit not applicable during either a TilERh1AL l'OWER ramp increase in excess of 5% of RATED TilliRh1AL POWER per minute or a TilERh1AL POWER step increase of greater than 10% of RATED TilERh1AL POWER.

October,1991

l' age 4

3. ANAI ATICAl h1121'llODS The analytical methods used to dete mine the cote operating limits shall be those previously resiewed and approved by the NI(C, specifically those described in the following documents:

3.1 XN-75-27(A), lley. O and Supplements 1 though 5. "liuon Nuclear Neutronics Design hlethods for l'ressurized Water Reactors," Euon Nuclear Company. Rev. O dated June 1975, Supplement I dated September 1976, Supplement 2 dated December 1980, Supplement 3 dated Septembe 1981, Supplement 4 dated December 1986, Supplement 5 dated I;ebruary 1987.

3.2 ANF-84-73(P), Rev. 3. " Advanced Nuclear Fuels hiethodology f or l'ressurited Water Reactons: Analysis of Chapter 15 livents," Advanced Nuclear Fuels Corporation, dated hlay 1988.

3.3 NN-NF-82-21( A), lley. 1. " Application of liuon Nuclear Company PWR The: mal hlargin hiethodology to hiixed Core Configurations." lixxon Nuclear Company, dated September 1983.

3.4 ANF-84-93(A), Rev. O and Supplement 1, "Steamline lireak blethodology for PWR's," Advanced Nuclear Fuels Corporation. Rev. O dated h1 arch 1989, Supplement 1 dated h1 arch 1989.

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

3.6 NN-NF-82-49(A), llev.1 and Supplement 1, "Enon Nuclear Company Evaluation h1odel liNEh1 PWR Small lireak htodel," Advanced Nuclear Fuels Corporation, both reports dated April 1989.

3.7 EXiiN1 PWR Large lireak LOCA Evaluation hiodel as defined by:

a. XN-NF-82-20(A), Rev. I and Supplements 1 through 4, " linen Nuclear Company Evaluation h1odel liNEht/I'WR ECCS hiodel Updates," Euon Nuclear Company. All reports dated January 1990.

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

c. NN-NF-81-58(A), Rev. 2 and Supplements 1 through 4. "RODEN2 Fuel Rod Thermal-htechanical Response Evaluation h1odel," Euon Nuclear Company. Rev. 2 and Supplements 1 and 2 dated h1 arch 1984, Supplements 3 and 4 dated June 1990.

October,1991

. l' age 5 i i

d. XN-NF-85-16(4), Volume 1 through Supplement 3; Volume 2, ller.1 and Supplement 1, "I'WR 17x17 Fuel Cooling Tests l'rogram," lixxon Nuclear Company. All reports dated February 1990.

e. XN-NF-85-105(A), Rev. O and Supplement 1, " Scaling of FCTF Based Reflood lleat Transfer Correlat.'on for Other Hundle Designs," ,

lixxon Nuclear Company, Doth reports dated January 1990.  !

3.8 XN-NF-78-44(Ah "A Generic Analysis of the Control Rod IIjection l Transient for Pressurized Water Reactors," Iluon Nuclear Company, dated  :

October 1983. '

3.9 XN-NF-621(A), Rev.1. "tixxon Nuclear DND Correlation of I'WR Fuel Design," Exxon Nuclear Company, dated September 1983.

An acceptable hillistone 2 speelfie application of these analytical methodologies is e described in ANF-88-126 "blillstone Unit 2 Cycle 10 Safety Analysis Report," ,

dated October 1988. ,

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l October,1991 l

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