ML090400968

Core Operating Limits Report, Revision 17

ML090400968
Person / Time
Site:	Palisades
Issue date:	02/09/2009
From:	Lahti L Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML090400968 (14)

Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel269 764 2000 February 9,2009 Technical Specification 5.6.5 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 Core Operating Limits Report - Revision 17

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. is providing revision 17 of the Palisades Nuclear Plant (PNP) Core Operating Limits Report (COLR). This report is submitted in accordance with the requirements of PNP Technical Specification 5.6.5.d. Enclosure 1 contains a summary of changes from the previous revision. Enclosure 2 contains revision 17 of the COLR.

Summarv of Commitments This letter contains no new commitments and no revision to existing commitments.

Laurie A. Lahti Licensing Manager Palisades Nuclear Plant Enclosures (2)

CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ENCLOSURE 1 PALISADES CORE OPERATING LIMITS REPORT REVISION 17

SUMMARY

OF CHANGES In this Core Operating Limits Report (COLR) revision, Section 3.0 is updated to add AREVA licensed topical report BAW-10240(P)-A, "lncorporation of M5 Properties in Framatome ANP Approved Methods," Revision 0, May 2004 to the list of approved analytical methods used to determine core operating limits.

This AREVA analytical methodology document was added to Technical Specification 5.6.5.b to permit the use of M5 alloy and supporting analytical methods in future reload designs. The addition of this document to the Technical Specifications was approved in Amendment No. 234 to Renewed Facility Operations License No. DPR-20 for the Palisades Nuclear Plant dated December 12, 2008.

This COLR revision incorporates the approved AREVA analytical methodology document in accordance with Technical Specification 5.6.5.

Page 1 of 1

ENCLOSURE 2 PALISADES CORE OPERATING LIMITS REPORT Revision 17 1 1 Pages Follow

Procedure No COLR Revision 17 Issued Date 2/6/2009 PALISADES NUCLEAR PLANT TITLE: CORE OPERATING LIMITS REPORT Verification Reviewer

/

/

Date Reactor Engineering Supervisor Date

Palisades COLR Revision 17 Page 1 of 10 Entergy Nuclear Palisades, LLC Entergy Nuclear Operations, Inc.

Docket No 50-255 License No DPR-20 Core Operating Limits Report 1.O INTRODUCTION This Core Operating Limits Report for Palisades has been prepared in accordance with the requirements of Technical Specification 5.6.5. The Technical Specifications Limiting Conditions for Operation (LCOs) affected by this report are listed below:

Section Title 2.1 SHUTDOWN MARGIN (SDM)

Regulating Rod Group Position Limits Linear Heat Rate (LHR)

Total Radial Peaking Factor AXIAL SHAPE INDEX (ASI)

PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

Palisades COLR Revision 17 Page 2 of 10 OPERATING LIMITS The cycle specific parameter limits for the specifications listed in Section 1 are presented in the following subsections. These limits have been developed using the NRC-approved methodologies specified in Section 3.0.

SHUTDOWN MARGIN (SDM) 2.1.1 MODES 1 and 2 (LC0 3.1.6 Regulating Rod Group Position Limits) - The minimum SDM requirement is 2% with the most reactive rod fully withdrawn. The rod insertion limit (PDIL) is discussed in Section 2.2 and shown in Figure 2.2-1.

2.1.2 MODES 3, 4 and 5, Loops Filled (LC0 3.1.1 SHUTDOWN MARGIN) - The SDM requirement is r 2% for normal cooldowns and heatups.

2.1.3 MODE 5, Loops Not Filled (LC0 3.1.1 SHUTDOWN MARGIN) - The SDM requirement is 2 3.5% assuming TaVe of 60°F.

2.1.4 MODE 6 (LC0 3.9.1 Boron Concentration) - The SDM requirement is specified in the definition of REFUELING BORON CONCENTRATION.

2.2 Regulating Rod Group Position Limits

a.

If the reactor is critical, to implement the limits on SHUTDOWN MARGIN, individual rod worth and hot channel factors, the limits on control rod regulating group insertion shall be established as shown on Figure 2.2-1.

b.

If the reactor is subcritical, the rod position at which criticality could be achieved if the control rods were withdrawn in normal sequence shall not be lower than Group 2 at 72 inches (ie, - 45% control rod insertion).

c.

The sequence of withdrawal of the regulating groups shall be 1, 2, 3, 4.

d.

An overlap of control banks in excess of 40% shall not be permitted.

Palisades COLR Revision 17 Page 3 of 10 1

POWER DEPENDENT INSERTION LIMITS 1

I I

I 132 80 I

I I

GROUP 4 40 0

132 GWUP2 80 60 1

I I

132 I

GROUP3 40 0

CONTROL ROD POSITION (Inches withdrawn)

Figure 2.2-1 Regulating Rod Group Position Limits NOTE:

A regulating rod is considered fully withdrawn at 2128 inches.

Palisades COLR Revision 17 Page 4 of 10 2.3 Linear Heat Rate (LHR)

The LHR in the peak powered fuel rod shall not exceed the following:

LHR 5 LHR~SX F* (2)

Where:

LH RTS =

Maximum allowable LHR shown in Table 2.3-1.

FA(z) =

Allowable LHR as a function of peak power location shown in Figure 2.3-1.

Table 2.3 Linear Heat Rate Limit To ensure that the design margin of safety is maintained, the determination of both the incore alarm setpoints and the Allowable Power Level takes into account the local LHR measurement uncertainty factors given in Table 2.4-2, an engineering uncertainty factor and a thermal power measurement uncertainty factor (values given in Technical Specification Basis B 3.2.1).

Peak Rod 15.28 (kWIft) 1.2 Unacceptable Operation (0.60, 1.OO)

(1.00, 0.93) 5 6 3 E0.8 0.7 Acceptable operation 0.0 0.2 0.4 0.6 0.8 1.O 1 2 Fraction of Active Fuel Height Figure 2.3 Allowable LHR as a Function of Peak Power Location

Palisades COLR Revision 17 Page 5 of 10 2.4 Total Radial Peaking Factor The radial peaking factor shall not exceed the following:

for P 2 0.5 Fr < F:~x[I.O+O.~X(I-P)]

and for P < 0.5, F, < FTsx1.15 Where:

Fr Measured F:,

FJ' Peaking Factor Limits (Table 2.4-I),

P Fraction of rated power.

Table 2.4 Peaking Factor Limits, F T ~

All Fuel Types To ensure that the design margin of safety is maintained, the determination of radial peaking factors takes into account the appropriate measurement uncertainty factors given in Table 2.4-2.

To ensure that the design margin of safety is maintained with respect to the alternative source term radiological consequence analysis assumptions as restricted by footnote 11 of Table 3 of Regulatory Guide 1.1 83, an evaluation shall be performed on the pin powerlburnup of the core design against the following criteria:

Fewer than 21 rods in any one assembly violate the "5416.3" criterion.

Fewer than 20 assemblies in any core design contain at least one rod that violates the "5416.3" criterion.

All rods that violate the "5416.3 criterion have a rod average linear heat generation rate of less than 6.7 kW1ft.

All rods that violate the "5416.3" criterion have a rod burnup of less than 58.5 GWDIMTU.

In any assembly containing any rods that violate the "5416.3" criterion there are at least four times as many rods that have total radial peaking factor of less than % of the total radial peaking factor limit of 2.04.

Palisades COLR Revision 17 Page 6 of 10 TABLE 2.4-2 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS AXIAL SHAPE INDEX (ASI)

The AS1 limit for the Tinlet function is shown in Figure 2.5-1 Measurement Uncertainty Acceptable Operation Break Points:

-0.550, 0.250

-0.300, 0.700

-0.080, 1.OOO

-0.080, 1.065

+0.400, 1.065

+0.400, 0.250 LHR 0.0500 1.2 1.1 1.0 1 0.9 P

3 O.*

a 0.7

'6 0.6 0 8 0.5 E

0.4 0.3 0. 2..

F:

0.0425 t

lhacceptable t

Operation i

j 3

-0.600

-0.400

-0.200 0.000 0.200 0.400 Axial Shape index Figure 2.5 AS1 Limit fur Tinlet Function

Palisades COLR Revision 17 Page 7 of 10 2.6 PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits

a.

Pressurizer pressure r 201 0 psia and r 21 00 psia

b.

PCS cold leg temperature 1 544 OF

c.

PCS total flow rate 2 352,000 gpm

Palisades COLR Revision 17 Page 8 of 10 3.0 ANALYTICAL METHODS The analytical methods used to determine the core operating limits are those previously reviewed and approved by the NRC, specifically those described in the Technical Specification Section 5.6.5 list of methodology documents. The Technical Specification 5.6.5 list is repeated below with revision numbers and dates added.

Specific application of these methodologies to Palisades is described in the cycle's most current safety analysis reports.

The analytical methods used to determine the radial peaking factor measurement uncertainty factors are described in FSAR, Section 3.3.2.5.

1.

EMF-96-029(P)(A) Volumes 1 and 2, "Reactor Analysis System for PWRs," Siemens Power Corporation, January 1997.

(LCOs 3.1.l, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

2.

ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events, Revision 5, July 1990," Advanced Nuclear Fuels Corporation. (Bases report not approved) (LCOs 3.1.1, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

3.

XN-NF-82-21 (P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"

Exxon Nuclear Company, Revision 1, September 1983.

(LCOs 3.2.1, 3.2.2, & 3.2.4)

4.

EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs,"

Revision 1, February 1999.

(LCOs 3.1.l, 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

5.

XN-75-32(P)(A) Supplements 1 through 4, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, October 1983.

(Bases document not approved)

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

6.

EMF-231 0 (P)(A), Revision 1, Framatome ANP, Inc., May 2004, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors."

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

7.

XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company, October 1983. (LCOs 3.1.6, 3.2.1, & 3.2.2)

8.

ANF-89-151 (P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, May 1992.

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

Palisades COLR Revision 17 Page 9 of 10 EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation, Revision 1, January 2005. (LCOs 3.2.1, 3.2.2, & 3.2.4)

XN-NF-621 (P)(A), "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, Revision 1, September 1983.

(LCOs 3.2.1, 3.2.2, & 3.2.4)

XN-NF-82-06(P)(A) Revision 1 and Supplements 2,4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company, October 1986. (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWDIMTU,"

Advanced Nuclear Fuels Corporation, December 1991.

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium DioxideIGadolinia Irradiation Examination and Thermal Conductivity Results," Exxon Nuclear Company, November 1986. (LCOs 3.1.6,3.2.1, 3.2.2, & 3.2.4)

EMF-92-1 16(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation, Revision 0, February 1999.

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

EMF-2087(P)(A), "SEMIPWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation, June 1999.

(LCOs 3.1.6, 3.2.1, & 3.2.2)

ANF-87-150 Volume 2, "Palisades Modified Reactor Protection System Report: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation, June, 1988. [Approved for use in the Palisades design during the NRC review of license Amendment 11 8, November 15, 19881 (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.4.1)

EMF-1 961 (P)(A), Revision 0, Siemens Power Corporation, July 2000, "Statistical Setpointfrransient Methodology for Combustion Engineering Type Reactors." (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4, & 3.4.1)

EMF-2328 (P)(A), Revision 0, Framatome ANP, Inc., March, 2001, "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based." (LCOs 3.1.6, 3.2.1, & 3.2.2)

BAW-2489P, "Revised Fuel Assembly Growth Correlation for Palisades, Revision 0, March 2005." (LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

EMF-2103(P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors, Revision 0, April 2003." (LCOs 3.1.6, 3.2.1,

& 3.2.2)

Palisades COLR Revision 17 Page 10 of 10

21.

BAW-10240(P)-A, Revision 0, May 2004, "Incorporation of M5 Properties in Framatome ANP Approved Methods." (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4,

& 3.4.1)