ML20106E331
| ML20106E331 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/30/1992 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20106E327 | List: |
| References | |
| NUDOCS 9211060339 | |
| Download: ML20106E331 (10) | |
Text
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' Docket Noi- 50-336 7
Millstone (Jnit No. 2 Cycle 12 Core Operating Limits Report Revision - s
- i 9211060339 921030
-PDR -ADOCK 05000336.
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- 1. col?E OPEllATING LIMITS REPORT This Core Operating Limits Report for hiillstone 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:
SGlion SEGification 2.1 3/4.1.1.1 SilUTDOWN h1ARGIN - Tm, > ?00*F 2.2 3/4.1.1.2 SIRTfDOWN h1ARGIN - Tm, s 200*F 2.3 3/4.1.1.4 hloderator Temperature Coefficient 2.4 3/4.1.3.6 Regulating CEA Insertion Limits
.15 3/4.2.1 Linear lleat Rate 2.6 3/4.2.3 TOTAL UNRODDED INTEGRATED RADIAL PHAKING FACTOR - F/
il 3/4.2.6 DNB hlargin Terms appearing in capitalized type are DEFINED TERNIS as defined in Section 1.0 of the Technical Specilications. e
- 2. OPEllATING l.lMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits l' ave been developed using ,
the N - C-approved methadologies specified in Section 3. { 2.1 SHUTDOWN hlARGIN - Tmg > 200*F (Specification 3/4.1.1.1) The SilUTDOWN h1ARGIN shall be 2 3.6% AK/K 2.2 SHUTDOWN hlARGIN - Tog s 200*F (Specification 3/4.1.1.2) The SilUTDOWN hlARGIN shall he 2 2.0% .AK/K 2.3 N1oderator Teniperature 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 wheneser THERNIAL POWER is s 70% of RATED THERh1As. POWER,
- b. Less positive than 0.4 x 10-4 AK/K/*F whenever THERh1AL POWER is > 70% of RATED THERh1AL POWER.
- c. Less negative than -2.8 x 10-4 AK/K/*F at RATED THERhlAL POWER.
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, L Page 2 2.4 Regulating CEA Insertion Limits'(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 A-L CEA insertion between the Long Term Steady State Inseition Limits and the Transier411. 'ertion Limits is restricted to:
- a. s 4 hours per 24 hour interval,
- b. -s 5 Effective Full Power Days per 30 Effective Full Power Day interval, and
- c. s 14 Elfective Full Power Days per calendar year. ,
2.5 Linear llect Rate _(Specification 3/4.2.1) , The Linear heat rate, including heat generated in the fuel, clad and moderator, shall not exceed 15.1 kw/ft. During operation with the' linear heat rate being monitored by the Excorel Detector Menitoring System, the AXlAL SHAPE INDEX 'shall remain . within the liinits of Figure 2.5-1. During operation with the linear heat rate being monitored by the incore: . Deuctor Monitor System, the alarm setpoints shall be adjusted to less than- ) or equal to the limit when the fc' lowing factors are appropriately: include <l iwthe setting of the alamts: 1.' A fliix peaking augmentation factor of 1.055,
- 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 THERMAL POWER measurement uncertainty: factor.of 1.02.
*These factors are only appropriate to fuel batches "A" through "L".
P 2.6 TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR - F/ (Specification 3/4.2.3) The calculated value of F,T shall t e s 1.69.
- 2. 6.1 - The Power Dependent 1/ limits'are shown in Figure 2.6-1. 3 F
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2.7 - DND hlargin'(Specification 3/4.2.6) The DND margin shall be preserved by maintaining the cold leg __ temperature, pressurizer pressure. reactor' coolant flow rate, and AX1AL SHAPE INDEX within the following limits: l Pntaneter 1,imits Four Reactor Coolant Pumps Operations
- u. Cold Leg Temperature .s 549'FL
- b. Pressurizer Pressure 2 2225 psia *
. c. Reactor Coolant Flow Rate 2 360.000 gpm -
- d. AX1AL SifAPE INDEX . FIGURE 2.7 .1 Limit not applicable during either a TilERh1AL POWER ramp-increase in excess of SG of RATED TI-lERMAL POWER per minute or a TIlERh1AL POWER step .
increase of greater than 10% of RATED THERhlAL POWER. L e P t d
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- 3. ' ANAIXi'ICAL METilODS-4 The' analytical methods used to determine the core operating limits shall be those F 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 Methods for Pressurized Water Reactors," ISxon , Nuclear Company. Rev, O dated June 1975, Supplenient 1 dited September 1976, Supplement 2 dated December 1980. Supplemem 3 dated September - 1981. Supplement 4 dated December 1986, Supplemein 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 hla) 1988. 3.3 XN-NF-82-21(A), Rev.1, ' Application of Exxon Nuclear Company' PWR Thermal Margin Metodology to Mixed Core Configurations." Exxon Nuclear Company, omed September 1983. 3.4 ANF-J4-93(A), Revi 0 and Supplement 1, "Steamline Break hiethodology for PWR's," Advanced Nuclear Fuels Corporation. Rev. O dated March 1989, Supplement 1 dated March 1989. 3.5 XN-75-32(A), Supplements'1, 2, 3, and 4, " Computational Procedure for Evaluating Fuel Rod Bowing," Exy.on Nuclear Company, dated ' October
~
1983. 3.6 XN-NF-82-49(A), Rev.1 and Supplement 1,~ " Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model,". Advanced . Nuclear Fuels Corporation, both reports dated April 1989, 3.7 EXEM PWR Large Break LOCA Evaluation Modelas defined by:
- a. XN-NF-82-20(A), Rev. I and Supplements 1 through 4, " Exxon Nuclear Company Evaluation blodel EXEM/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 1 Swelling and Rupture Model " Exxon Nuclear Company, dated November 1982. ,
- c. XN-NF 81-58(A), Rev. 2 and Supplements 1 through 4, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," E:non Nuclear Company. Rev. 2 and Supplements I and 2 dated March
-1984, Supplem:nts 3 and 4 dated June 1990.
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- d. XN-NP-85-16(A), Volume ! through Supplement-3; Volume 2, ReeJ1-:
and Supplement 1f"PWR- 17x17 Fuel Cooling Tests Program, _ Exxon l Nuclear Company. All reports dated February 1990. o
- e. -XN-NP.-85-lo:i(A), Rev. O and Supplement 1, " Scaling.of itCTF Based Reflood Heat 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 Reac(ors," Exxon Nuclear Company,' dated October 1983. : 3.9 XN-NF-621(A), Rev.1, " Exxon Nuclear DNB Correlation of PWR Fuei Design," Exxon Nuclear Company, dated September 1983. An acceptable Millstone 2 specific application of these analytical methodologies is described in ANP-88-126, " Millstone Unit 2 Cycle 10 Safety ' Analysis Report," , dated October 1988. h
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