ML100640459

Core Operating Limits Report - Revision 18

ML100640459
Person / Time
Site:	Palisades
Issue date:	03/05/2010
From:	Patricia Anderson Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML100640459 (14)

Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant Entergy 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Paula K. Anderson Licensing Manager March 5, 2010 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Core Operating Limits Report - Revision 18 Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. is providing revision 18 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. Attachment 1 contains a summary of changes from the previous revision. Attachment 2 contains revision 18 of the COLR.

Summary of Commitments This letter contains no new commitments and no revisions to existing commitments.

/2IY~

pka/jse Attachment(s): 1. Palisades Core Operating Limits Report Revision 18 Summary of Changes

2. Palisades Core Operating Limits Report Revision 18 CC Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ATTACHMENT 1 PALISADES CORE OPERATING LIMITS REPORT REVISION 18

SUMMARY

OF CHANGES This Core Operating Limits Report revision changes two of the five reactor core design criteria in Section 2.4, titled "Total Radial Peaking Factor."

Section 2.4 contains a requirement to perform an evaluation of the fuel assembly pin power and burnup in the reactor core design against five criteria. The purpose of the evaluation is to ensure that design margin of safety is maintained with respect to alternative source term radiological consequence analysis assumptions contained in Regulatory Guide 1.183. Two of the criteria are revised to allow more efficient utilization of nuclear fuel in future reactor core designs.

Specifically, the first criterion is revised to change the restriction on the number of rods in anyone assembly that exceed the "54/6.3" criterion in Regulatory Guide 1.183 from 21 to 42 rods. The "54/6.3" criterion refers to a restriction in the Regulatory Guide that maximum linear heat generation rate not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU.

The second criterion is revised to change the restriction on the number of fuel assemblies in any core design that contain at least one rod that exceeds the "54/6.3" criterion in the Regulatory Guide from 20 to 42 assemblies.

The revised criteria continue to preserve the design margin of safety with respect to alternative source term radiological consequence analysis assumptions.

Page 1 of 1

ATTACHMENT 2 PALISADES CORE OPERATING LIMITS REPORT REVISION 18 11 Pages Follow

Procedure No COLR Revision 18 Effective Date 2/10/10 PALISADES NUCLEAR PLANT TITLE: CORE OPERATING LIMITS REPORT Approved:~G~EJ_a_r~k_a~~ _____________________I__________________~2~/5=/~1O~

Procedure Sponsor Date IProcess Applicability Exclusion D New Procedure/Revision Summary:

Specific Changes Revision 18 See changes in Section 2.4, Total Radial Peaking Factor

Palisades COLR Revision 18 Page 1 of 10 Entergy Nuclear Palisades, LLC Entergy Nuclear Operations, Inc.

Docket No 50-255 License No DPR*20 Core Operating Limits Report

1.0 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 LCO 2.1 SHUTDOWN MARGIN (SDM) 3.1.1 3.1.6 3.9.1 2.2 Regulating Rod Group Position Limits 3.1.6 2.3 Linear Heat Rate (LHR) 3.2.1 2.4 Total Radial Peaking Factor 3.2.2 2.5 AXIAL SHAPE INDEX (ASI) 3.2.4 2.6 PCS Pressure, Temperature, and Flow 3.4.1 Departure from Nucleate Boiling (DNB) Limits

Palisades COLR Revision 18 Page 2 of 10 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 (LCO 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 (LCO 3.1.1 SHUTDOWN MARGIN) - The SDM requirement is ~ 2% for normal cooldowns and heatups.

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

2.1.4 MODE 6 (LCO 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 18 Page 3 of 10 POWER DEPENDENT INSERTION LIMITS FOUR PUMP OPERAllON 100 90 80

~ 70 I r-rr ~ I II I I 0::

w I1I1 III I I l"t-k I I I I I

~11. 60 0

0:: 50 J II I l'{ I I I I I II l-I)

~ 40 III I II ~ I I I 0::

30 I 11111 II III '~ II III 20 OPERATING REGION II II"J II I I I 10 1 111 11 11 1 1 1 11 11 IH-.,

0 r-t+- r-l I I I I I I 132 SO GROUP 4 40 o 132 GROUP 2 sb 60 I I I 132 SO GROU' 3 o CONTROL ROD POSITION (Inches withdrawn)

Figure 2.2-1 Regulating Rod Group Position Limits NOTE: A regulating rod is considered fully withdrawn at ~128 inches.

Palisades COLR Revision 18 Page 4 of 10 2.3 Linear Heat Rate (LHR)

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

Where:

Maximum allowable LHR shown in Table 2.3-1.

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

Table 2.3 Linear Heat Rate Limit Peak Rod 15.28 (kW/ft)

II 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).

1.2 Unacceptable

~1.1 Operation .

3 IX E (0.60, 1.00)

5 ~1.0 __-----------6__

~I

! '00.9 (1.00,0.93)

c

- c 0

~

Acceptable Co)

Operation 1!0.8

.LL.

0.7 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Active Fue I He Ight Figure 2.3 Allowable LHR as a Function of Peak Power Location

Palisades COLR Revision 18 Page 5 of 10 2.4 Total Radial Peaking Factor The radial peaking factor shall not exceed the following:

for P ~ 0.5 Fr:S; F~s x [1.0 +0.3 x(1-P)]

and for P < 0.5, Where:

Fr = Measured F~,

Peaking Factor Limits (Table 2.4-1),

P = Fraction of rated power.

Table 2.4 Peaking Factor Limits, F~s All Fuel Types 2.04 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 power/burnup of the core design against the following criteria:

• Fewer than 42 rods in anyone assembly violate the "54/6.3" criterion.

• Fewer than 42 assemblies in any core design contain at least one rod that violates the "54/6.3" criterion.

• All rods that violate the "54/6.3" criterion have a rod average linear heat generation rate of less than 6.7 kW/ft.

• All rods that violate the "54/6.3" criterion have a rod burnup of less than 58.5 GWD/MTU.

• In any assembly containing any rods that violate the "54/6.3" 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 18 Page 6 of 10 TABLE 2.4-2 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS LHR FrT Measurement 0.0500 0.0425 Uncertainty 2.5 AXIAL SHAPE INDEX (ASI)

The ASI limit for the Tinlet function is shown in Figure 2.5-1.

1.1 lklacceptable 1.0

... Operation I

D-0.9 O.S IIII:: 0.7 Acceptable

'0 0.6 Operation c

0

w u 0.5 l!

u. 0.4 0.3 0.2

-0.600 -0.400 -0.200 0.000 0.200 0.400 Axial Shape Index Figure 2.5 ASI Limit for Tinlet Function Break Points:

-0.550, 0.250

-0.300, 0.700

-0.080, 1.000

-0.080, 1.065

+0.400, 1.065

+0.400, 0.250

Palisades COLR Revision 18 Page 7 of 10 2.6 pes Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits

a. Pressurizer pressure ~ 2010 psia and :s; 2100 psia

b. pes cold leg temperature :s; 544 of

c. pes total flow rate ~ 352,000 gpm

Palisades COLR Revision 18 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 3.1.1, 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 3.1.1, 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)

(LCOs 3.1.6, 3.2.1,3.2.2, & 3.2.4)

6. EMF-2310 (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 18 Page 9 of 10

9. 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)

10. 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)

11. 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)

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

Advanced Nuclear Fuels Corporation, December 1991.

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.2.4)

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

14. EMF-92-116(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)

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

(LCOs 3.1.6, 3.2.1, & 3.2.2)

16. 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 118, November 15, 1988]

(LCOs 3.1.6, 3.2.1, 3.2.2, & 3.4.1)

17. EMF-1961(P)(A), Revision 0, Siemens Power Corporation, July 2000, "Statistical SetpointiTransient Methodology for Combustion Engineering Type Reactors." (LCOs 3.1.6, 3.2.1, 3.2.2, 3.2.4, & 3.4.1)

18. 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)

19. 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)

Palisades COLR Revision 18 Page 10 of 10

20. 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)

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)