ML20035D483

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Rev 0 to Grand Gulf Nuclear Station COLR,Safety-Related
ML20035D483
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 04/01/1993
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20035D482 List:
References
GGNS-MS-48.0, GGNS-MS-48.0-R, GGNS-MS-48.0-R00, NUDOCS 9304130255
Download: ML20035D483 (20)
v • d • e


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Standard No.: GGNS-MS-48.0 Revision:

0 Date:

April 1. 1993 d

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i Grand Gulf Nuclear Station Core Operating Limits Report Safety-Related 3

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Standard No.: GGNS-MS.48.0 i

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ii of vi Resision No.: 0 GRAND GULF NUCLEAR STATION NUCLEAR PLANT ENGINEERING REVIEW AND APPROVAL SHEET STANDARD NO.:

GGNS.MS.48.0 REVISION: 0 i

STANDARD TITLE: Core Operating Limits Report This document specifies items related to nuclear safety YES [X] NO [ ]

Signatures certify that the above standard was originated, verified, reviewed or waived and i

approved as noted below:

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VERIFIED BY:

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a Cognizant Group Supenisor NPE SECTION REVIEWED BY n.

REVIEW WAIVED BY DATE

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APPROVED BY:

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3 Standard No.: GGNS-MS-48.0 Page:

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Revision No.: 0 0

SAFETY EVALUATION APPLICABILITY REVIEW FORM A)

Document Evaluated: Standard GGNS-MS-48. Revision 0 B)

Description of the Proposed Change: Per GNRI-934X)8. Amendment 106 to the Grand Gulf Operatine License. Enterev committed to removine certain reactor ohrsies parameters from the Technical Specifications and placing them in a separate report orecared for each fuel cycle. Standard GGNS-MS-48 is the Core Operatine Limits Report (COLR) and establishes these varameters.

PRE-SCREENING Check the applicable boxes below. If any of the boxes are checked, neither a safety evaluation applicability rniew nor a safety evaluation is necessary and steps C, D, E, and F may be skipped. The preparer and reviewer must sign at the bottom of the form.

The change is editorial only.

10CFR50.54 applies to the change instead of 10CFR50.59.

An approved safety evaluation covering all aspects of this subject already exists.

Reference SE#

2.

The change, in its entirety, has been approved by the NRC.

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Reference:

Both the Cvele 6 operatine limits and their relocation from the Technical Foecifications has been rniewed and aooroved by the NRC. The Cvele 6 core operating limits have been approved by the NRC in GNRI-92/00114. Cycle 6 Reload. Amend. 99 to GGNS Operating License. dated May 28.1992.

The relocation of the core operatine limits from the Technical Soecifications into the COLR has been approved in GNRI-93-0008. Amend.106 to Grand Gulf Operating License. dated January 2L 1993.

The change is an T3AR change that meets the exclusion criteria outlined in Site Directive G4.803 Safety Evaluation Applicabi ity Review If any of the following questio ts are answered "yes", then a full 50.59 Safety Evaluation must be completed.

C)

Does the proposed change or activity represent a change to the Technical Specifications?

YES __

Explain:

NO _

D)

Does the proposed change or actisity represent:

(1)

A change to the facility which alters, or has the potential to alter, the information, operation, function or ability to perform the function of a system, structure or component described in the SAR?

YES _

Explain:

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i Standard No.: GGNS-MS-48 0 Page:

iv of vi Revision No.: 0 (2)'

A change to a procedure which alters, or has the potential to alter, a procedure described, outlined or sumnanzed in the SAR7 YES _ ~~

Explain:

NO _,_

(3)

A test or experiment not described in the S AR or which requires that a system be operated in an abnormal manner that is not described or prniously analyzed in the SAR?

YES__

Explain:

NO __

3M hr sloe /o PREPARER-OdU.)x NMD)

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REVIEs 0

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<y Name ich Title Date If the preparer performed an applicability rniew, the rniewer should check below to indicate by which means the independent rniew reached the same conclusions.

ON)

Reviewed the applicability review documentation.

t Completed an independent applicability rniew.

Performed a verbal rniew with the preparer.

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Standard No.: GGNS-MS-48.0

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Page:

v of vi Revision No.: 0

,!O REVISION STATUS SHEET i

I STANDARD REVISION SUMhfARY i

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i REVISION ISSUE DATE DESCRIPTION i

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Issued for use g93 i

i PAGE REVISION STATUS PAGE NO.

REVISION PAGE NO.

REVISION 1

0 11 0

2 0

12 0

3 0

13 0

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14 0

5 0

6 0

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APPENDIX /ATTACIIMENT STATUS O

Standard No.: GGNS-MS-48.0

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Revision No.: 0 O

TABLE OF CONTENTS l.0 PURPOSE...

.1 2.0 SCOPE.

.2

3.0 REFERENCES

.3 4.0 DEFINITIONS.

.6 t

5.0 GENERAL REQUIREMENTS.

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i Standard No.: GGNS-MS-48.0 Page:

1 of 14 Resision No.: 0 O

1.0 PURPOSE On October 4,1988, the NRC issued Generic Letter 88-16 (Reference 29) encouraging licensees to remove cycle-specific parameter limits from Technical Speci5 cations and to place these limits in a formal report to be prepared by the licensee. As long as the parameter limits were developed with NRC-approved methodologies, the letter indicated that this would remove unnecessary burdens on licensee and NRC resources.

On October 29,1992, Entergy Operations complied with this letter by submission of a Proposed Amendment to the Grand Gulf Operating License (Reference 30). This document requested changes to the GGNS Technical Specifications to remove certain reactor physics parameter limits that change each fuel cycle. This amendment committed to placing these operating limits in a separate Core Operating Limits Report (COLR) which will be defined in Technical Specifications.

This PCOL was approved by the NRC by SER dated January 21,1993 (Reference 31).

The developnient of this COLR is the responsibility of Design Engineering. The purpose of Standard GGNS-MS-48.0 is to develop the Core Operating Limits Report from the available supporting documents for c4J fu

  • cvele. This standard will be revised accordingly for each fuel cycle or remaining portion of a fuel cycle.

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Standard No.: GGNS-MS-48.0 Page:

2 of 14 Resision No.: 0 0

2.0 SCOPE As defined in Technical Specification 1.7a, the COLR is the GGNS document that prosides the core operating limits for the current fuel cycle. This document is prepared in accordance with Technical Specification 6.9.1.11 for each reload cycle using NRC-approved analytical methods.

The Cycle 6 core operating limits included in this report are:

1.

the Average Planar Linear Heat Generation Rate (APLHGR) limits for each fuel type for both two-loop and single-loop operation. (Technical Specification 3.2.1),

2.

the Minimum Critical Power Ratio (MCPR) operating limit including the power, flow, and exposure dependent curves. (Technical Specification 3.2.3), and i

3.

the Linear Heat Generation Rate (LHGR) limit for each fuel type including the power and flow dependent parametric adjustment factor curves, LHGRFAC and LHGRFAC,

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respectively. (Technical Specification 3.2.4)

The cycle-specific MCPR safety limits are documented in Technical Specification 2.1.2.

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Standard No.: GGNS-MS-48.0 I

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3.0 REFERENCES

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h This section contains the methodology and cycle-speci5c references used in the safety analysis of Grand Gulf Cycle 6.

The supplements and revisions of the current analytical methodology l

references are included below in accordance with Technical Specification 6.9.1.11, Core j

Operating Limits Report.

METHODOLOGY

REFERENCES:

i 1.)

XN-NF-79-71(P), Revision 2 including Supplements 1, 2, and 3, Exxon Nuclear Plant Transient Methodoloev for Boiling Water Reactors, Exxon Nuclear Company, Inc.,

i Richland, WA, November 1981. Approved by NRC letter dated October 24,1986.

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2.)

XN-NF-80-19(P)(A), Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodolouv l

for Boiling Water Reactors - Neutronic Methods for Desien and Analysis, Exxon Nuclear Company, Inc., Richland, WA, March 1983.

3.)

XN-NF-80-19(P)(A), Volume 1 Supplements 3 and 4, Advanced Nuclear Fuels Methodoloev for Boiling Water Reactors: Benchmark Results for the CASMO-i 3G/MICROBURN-B Calculation Methodoloev, Advanced Nuclear Fuels Corporation, Inc., Richland, WA, November 1990.

4.)

XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Methodoloev for Boiling f

Water Reactors THERMEX: Thermal Limits Methodolouv Summary Description, Exxon Nuclear Company, Inc., Richland, WA, January 1987.

5.)

ANF-913 (P)(A), Volume 1, Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2 A Computer Program for Boiline Water Reactor Transient Analysis, 3

j Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.

I i

6.)

ANF-1125 (P)(A) and Supplements 1 and 2, ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, Richland, WA, April 1990.

]

7.)

XN-NF-84-105(P)(A), Volume 1 and Supplements 1 and 2, XCOBRA-T: A Computer l

Code for BWR Transient Thermal Hydraulic Core Analysis, Exxon Nuclear Company, Inc., Richland, WA, February 1987.

8.)

XN-NF-573(P), RAMPEX Pellet-Clad Interaction Evaluation Code for Power Ramos, Exxon Nuclear Company, Inc., Richland, WA, May 1982. Approved by NRC letter dated August 28,1990.

4 9.)

XN-NF-81-58(P)(A) and Supplements 1 and 2, Revision 2, RODEX2: Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA, March 1984.

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Standard No.: GGNS-MS-48.0 Page:

4 of 14 Revision No.: 0 O

10.)

'N-NF-85-74(P)(A), RODEX2A (BWRh Fuel Rod Thermal-Mechanical Response aluation Model, Exxon Nuclear Company, Inc., Richland, WA, August 1986.

11.)

XN-CC-33(P)(A), Revision 1, HUXY: A Generalized Multirod Heatuo Code with 10CFR50 Anoendix K Heatuo Option, Exxon Nuclear Company, Inc., Richland, WA, November 1975.

12.)

XN-NF-825(P)(A) Supplement 2, BWR/6 Generic Rod Withdrawal Error Analysis.

MCPR for Plant Operation Within the Extended Operating Domain. Exxon Nuclear p

Company, Inc., Richland, WA, October 1986.

13.)

XN-NF-81-51(P)(A), LOCA-Seismic Structural Response of an Exxon Nuclear Comoany n

BWR Jet Pumo Fuel Assembly, Advanced Nuclear Fuels Corporation, Richland, WA, May 1986.

14.)

XN-NF-84-97(P)(A), LOCA-Seismic Structural Response of an ENC 9x9 BWR Jet Pump Fuel Assembly, Exxon Nuclear Company, Inc., Richland, WA, August 1986.

15.)

XN-NF-86-37(P), Generic LOCA Break Spectrum Analysis for BWR/6 Plants, Exxon Nuclear Company, Inc., Richland, WA, April 1986. Approved by NRC letter dated October 24,1986.

16.)

XN-NF-82-07(P)(A), Revision 1, Exxon Nuclear Company ECCS Cladding Swelling and i

Rupture Model, Exvon Nuclear Company, Inc., Richland, WA, November 1982.

17.)

XN-NF-30-19(A), Volumes 2,2A,2B, and 2C, Exxon Nuclear Methodolocv for Boiling Water Reactors EXEM BWR ECCS Evaluation Model, Exxon Nuclear Company, Inc.,

Richland, WA, September 1982.

18.)

XN-NF-79-59(P)(A), Methodoloev for Calculation of Pressure Drop in BWR Fuel Assemblies. Exxon Nuclear Company, Inc., Richland, WA, November 1983.

CYCLE 6

REFERENCES:

19.)

EMF-91-169, Revision 1, Grand Gulf Unit 1 Cycle 6 Reload Analysis Siemens Power Corporation, Richland, WA, July 1992.

i 20.)

EMF-91-168, Revision 2, Grand Gulf Unit 1 Cycle 6 Plant Transient Analysis, Siemens Power Corporation, Richland, WA, September 1992.

21.)

ANF-86-133, Revision 4, Principal ECCS and Plent Transient Analysis Parameters Grand Gulf Unit 1, Advanced Nuclear Fuels Corporation, Richland, WA, June 1991.

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Standard No.: GGNS-MS-48.0

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Revision No.: 0 22.)

ANE-91-080(P), ma lulf 1 ANF-1.5 Design Report. Mechanical. Thermal-Hydraulic.

and Neutronic Dt.gn for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies. Advanced l

Nuclear Fuels, Richland, WA, July 1991.

23.)

GNRO-91-00186, Proposed Amendment to Grand Gulf Operating License, PCOL 91/23 -

Cycle 6 Reload, December 5,1991.

i 24.)

GNRI-92/00114, Amendment 99 to Operating License NPF-29. Nuclear Regulatory Commission, May 28,1992.

j 4

25.)

EMF-91-172, Grand Gulf Unit 1 LOCA Analysis for Single Loop Operation. Siemens Power Corporation, Richland, WA, October 1991.

CYCLE 5

REFERENCES:

26.)

ANF-89-171(P), Volumes 1 and 2, Grand Gulf 1 ANF-1.4 Desit. Report. Mechanical.

Thermal-Hydraulic and Neutronic Design for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies. Advanced Nuclear Fuels, Richland, WA, January 1990.

27.)

ANF-88-152(P)(A) with Amendment I and Supplement 1, Generic Mechanical Design for Advanced Nuclear Fuels 9x9-5 BWR Reload Fuel, Advanced Nuclear Fuels Corporation, i

Richland, WA, November 1990.

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CYCLE 4

REFERENCES:

28.)

ANF-88-183(P), Supplement 1, Reload XN-1.3. Cycle 4 Mechanical Design Renort.

l Siemens Nuclear Power Corporation, Richland, WA, August 1991.

1 l

1 GENERAL REFERENCES.

29.)

MAEC-88/0313, Generic Letter 88-16, " Removal of Cycle-Specific Parameter Limits from Technical Specifications", October 4,1988.

l 30.)

GNRO-92-00093, Proposed Amendment to Grand Gulf Gperating License, PCOL-92/07, dated October 29,1992.

31.)

GNRI-93-0008, Amendment 106 to Grand Gulf Operating License, January 21,1993.

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Standard No.: GGNS-MS-48.0 Page:

6 of 14 Resision No.: 0 4.0 DEFINITIONS

'I 1.

Average Planar Linear Heat Generation Rate (APLHGR) - the APLHGR shall bc

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applicable to a specific planar height and is equal to the sum of the linear heat generation rates for all the fuel rods in the specifi-d bundle at the specified height divided by the number of fuel rods in the fuel bundle. (Technical Specification 1.3) 2.

Average Planar Exposure - the Average Planar Exposure shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel cods in the fuel bundle.

i (Technical Specification 1.2) 3.

Critical Power Ratio (CPR) - the ratio of that power in the assembly which is calculated by application of the ANFB boiling correlation (Reference 6) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

(Technical Specification 1.8) 4.

Core Onerating Limits Report (COLR) - The Grand Gulf Nuclear Station specific document that provides core operating limits for the current reload cycle in accordance with Technical Specification 6.9.1.11. (Technical Specification 1.7a) r~(

5.

Linear Heat Generation Rate (LHGR) - the LHGR shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length. (Technical Specification 1.21) 6.

Minimum Critical Power Ratio (MCPR) - the MCPR shall be the smallest CPR which exists in the core. (Technical Specification 1.25) 7.

MCPR Safety Limit - the minimum value of the C"R at which the fuel could be operated with the expected number of rods in boiling transition not exceeding 0.1% of the fuel rods in the core. (Reference 20, Section 3.4)

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1 Standard No.: GGNS-MS-48.0 Page:

7 of 14 Revision No.: 0 5.0 GENERAL REQUIREMENTS This section reports,the Grand Gulf Cycle 6 core operating limits. These limits are taken from Reference 19 Sections 5.7,6.1.3, and 7.2.3.

Averaec Planar Linear Heat Generation Rates (Technical Specification 3.2.1)

During two-loop operation, all AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits in Figure 5.1.

During single-loop operation, the APLHGR for each type of fuel as a function of average planar exposure shall not exceed the limits shown in Figure 5.1 multiplied by 0.86.

Minimum Critical Power Ratio (MCPR) (Technical Specification 3.2.3)

The MCPR shall be equal to or greater than the MCPR, MCPR, and MCPR limits at the f

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indicated core flow, thermal power, and exposure as shown in Figures 5.2, 5.3, and 5.4.

Linear Heat Generation Rate (LHGR)(Technical Specification 3.2.4)

The LHGR shall not exceed the limits shown in Figure 5.5 as multiplied by the smaller of either O.

the flow-dependent LHGR factor (LHGRFACr) of Figure 5.6 or the power-dependent LHGR factor (LHGRFAC ) of Figure 5.7.

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