Text

Core Operating Limits Report - Revision 16
	ML081500826
Person / Time
Site:	Palisades
Issue date:	05/28/2008
From:	Lahti L; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML081500826 (13)
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Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel269 764 2000 May 28,2008 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 16

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. is providing revision 16 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 16 of the COLR.

Summary 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 16

SUMMARY

OF CHANGES Documentation for all analytical methods used to determine the core operating limits currently listed in Technical Specification Section 5.6.5 is added to COLR Section 3.0, with documentation revision numbers and dates.

Page 1 of 1

ENCLOSURE 2 PALISADES CORE OPERATING LIMITS REPORT REVISION 16 10 Pages Follow

Procedure No COLR Revision 16 Issued Date 5/1/2008 PALISADES NUCLEAR PLANT TITLE: CORE OPERATING LIMITS REPORT

Palisades COLR Revision 16 Page 1 of 9 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 16 Page 2 of 9 2.0 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.

2.1 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 2 2% for normal cooldowns and heatups.

2.1.3 MODE 5, Loops Not Filled (LC0 3.1 .I SHUTDOWN MARGIN) - The SDM requirement is 2 3.5% assuming ,T ,, 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 16 Page 3 of 9 POWER DEPENDENT INSERTION LIMITS FOUR P U M P OPERATION r I 8 I 1 132 80 40 0 132 60 GROUP 4 GROUP:! 80 CONTROL ROD POSITION (inches withdrawn)

Figure 2.2-1 Regulating Rod Group Position Limits NOTE: A regulating rod is considered fully withdrawn at 212%inches.

Palisades COLR Revision 16 Page 4 of 9 2.3 Linear Heat Rate (LHR)

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

LHR LHR~~x F~( 2 )

Where:

LHRTS= 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 2.3-1 Linear Heat Rate Limit Peak Rod 15.28 (kW/ft)

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).

Unacceptable Operation Acceptable Operation Fraction of Active Fuel Height Figure 2.3-1 Allowable LWR as a Function of Peak Power Location I

Palisades COLR Revision 16 Page 5 of 9 2.4 Total Radial Peaking Factor The radial peaking factor shall not exceed the following:

for P 2 0.5 F, < ~ ; ~ x [ 1 . 0 + 0 . 3 ~ ( l - P ) ]

and for P < 0.5, F, < F;'x~ .15 Where:

- Measured FJ, Fr FJS = Peaking Factor Limits (Table 2.4-1),

P -

- Fraction of rated power.

Table 2.4-1 - Peaking Factor Limits, 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 1.183, an evaluation shall be performed on the pin powerfburnup 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 GWD/MTU.

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 16 Page 6 of 9 TABLE 2.4-2 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS LHR F, '

Measurement 0.0500 0.0425 Uncertainty AXIAL SHAPE INDEX (ASI)

The AS1 limit for the Tinletfunction is shown in Figure 2.5-1 Acceptable Operation Axial Shape Index 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

Palisades COLR Revision 16 Page 7 of 9 2.6 PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits

a. Pressurizer pressure 2 2010 psia and 5 2100 psia

b. PCS cold leg temperature 5 544 OF

c. PCS total flow rate 2 352,000 gpm

Palisades COLR Revision 16 Page 8 of 9 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. I , 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,jl Revision 1, February 1999.

(LCOs 3.1.1, 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)