ML20077A298

GGNS COLR Safety-Related

ML20077A298
Person / Time
Site:	Grand Gulf
Issue date:	08/26/1994
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20077A297	List: ML20077A293 ML20077A298
References
GGNS-MS-48.0, GGNS-MS-48.0-R02, GGNS-MS-48.0-R2, NUDOCS 9411220260
Download: ML20077A298 (21)

Text

Standard No.: GGNS-MS-48.0 Revision:

2 Date:

August 26, 1994 Grand Gulf Nuclear Station Core Operating Limits Report Safety-Related i

9411220260 941110 PDR ADOCK 05000416

.P PDR

Standard No.: GGNS.MS.48.0 Page:

ii of vi Revision No.: _2__

GRAND GULF NUCLEAR STATION NUCLEAR PLANT ENGINEERING REVIEW AND APPROVAL SHEET STANDARD NO.:

GGNS-MS.48.0 REVISION:

2 STANDARD TITLE: Core Ooeratina Limits Reoort This document specifies items related to nuclear safety YES [X]

NO[]

Signatures certify that the above standard was originated, verified, reviewed or waived and approved as noted below:

ORIGINATED BY:

.N-DATE:

M 9Y M

DATE: F/// (V VERIFIED BY:

/ /

REVIEWED BY:

AJ DATE: 8/25/N Cognizant Group Supervisor NPE SECTION REVIEWED BY REVIEW WAIVED BY DATE k.

F!2 5/9i ELECTRICAL CIVIL h

S h, MECHANICAL u-ANII:

MA DATE:

(Insert N/A if n t applicable)

APPROVED BY:

DATE:

24 44 Q' Responsible Manager I

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

iii of vi Revision No.: _2_

l SAFETY EVALUATION APPLICABILITY REVIEW FORM j

l A)

Document Evaluated: Standard GGNS-MS-48.0. Revision 2 B)

Description of the Proposed Change: Per GNRI-93 008. Amendment 106 to the Grand Gulf Operating License. Enterry committed to removing certain reactor ohysics carameters from the Technical Specifications and olacing thtm in a separate reoort orecared for each fuel cycle. Standard GGNS-MS-48.0 is the Core Occrating Limits Report (COLR) and establishes these carameters. Revision 2 to GGNS-MS-48.0 uodates the Cvele 7 references in response to ODR 0159 94.

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

The change is editorial only.

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

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

Reference SE# The Cycle 7 core operating limits have been evaluated in the Cycle 7 reload safety evaluation. 93-0100-R01.

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

Reference:

The change is an FS AR change that meets the exclusion criteria outlined in Site Directive G4.803 Safety Evaluation Aonlicability Resiew If any of the following questions 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 activity represent:

(1)

A change to the facility which alters, or has the potential to alter, the information, operation, i

function or ability to perform the function of a system, structure or component described in the SAR?

YES _,_

Explain:

NO _

Standard No.: GGNS-MS-48.0 Page:

iv of vi Revision No.: 2 (2)

A change to a procedure which alters, or has the potential to alter, a procedure described, outlined or summarized in the 3AR7 YES__

Explain:

N O _._

(3)

A test or experiment not described in the SAR or which requires that a system be operated in an abnormal manner that is not described or previously analyzed in the SAR7 YES _

Explain:

NO _

b@I O NY PREPARER h

flF' s Na i U

Job 31e 1 Dhte fr. 6 v,

((////p REVIEWER A

Name Job Title

' ' Date If the preparer performed an applicability review, the reviewer should check below to indicate by which means the independent review reached the same conclusions.

2 Reviewed the applicability review documentation.

Completed an independent applicability review.

Performed a verbal review with the preparer.

i

~

Standard No.: GGNS-MS-48.0 Page:

v of vi Revision No.: _2_

REVISION STATUS SHEET STANDARD REVISION

SUMMARY

REVISION ISSUE DATE DESCRIPTION 0

April 1,1993 Issued for use 1

November 12,1993 Issued for use 2

August 26, 1994 Issued for use PAGE REVISION STATUS PAGE NO.

REVISION PAGE NO.

REVISION i-vi 2

9 1

1 1

10 1

2 1

11 1

3 1

12 1

4 1

13 1

5 2

14 1

6 0

15 1

7 1

8 1

APPENDIX / ATTACHMENT STATUS

1 e

Standard No.: GGNS-MS-48.0 Page:

vi of vi Revision No.: 2 TABLE OF CONTENTS 1.0 PURPOSE.,

.1 2.0 SCOPE.,......

.2

3.0 REFERENCES

.3 4.0 DEFINITIONS..

.6 5.0 GENERAL REQUIREMENTS.

.7 i

l l

Standard No.: GGNS-MS-48.0 Page:

I of 15 Revision No.: 1 1.0 PURPOSE On October 4,1988, the NRC issued Generic Letter 88-16 (Reference 28) encouraging licensees l to remove cycle-specific parameter limits from Technical Specifications 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 29). This document requested l 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 30).

l The development of this COLR is the responsibility of Design Engineering. The purpose of Standard GONS-MS-48.0 is to develop the Core Operating Limits Report from the'available supporting documents for each fuel cycle. This standard will be revised accordingly for each fuel cycle or remaining portion of a fuel cycle.

4

Standard No.: GGNS-MS-43.0 Page:

2 of 15 Revision No.: 1 2.0 SCOPE As defined in Technical Specification 1.7a, the COLR is the GGNS document that provides 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 7 core operating limits included in this report are:

l 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 (as a function of exposure) and flow dependent curves. (Technical Specification 3.2.3), and 3.

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

respectively. (Technical Specification 3.2.4)

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

Standard No.: GGNS-MS-48.0 Page:

3 of 15 Revision No.: 1

3.0 REFERENCES

This section contains the methodalogy and cycle-specific references used in the safety analysis of Grand Gulf Cycle 7.

The supplements and revisiont of the current analytical methodology l references are included below in accordance with Technical Specification 6.9.1.11, Core i

Operating Limits Report.

METHODOLOGY

REFERENCES:

1.)

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

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

2.)

XN-NF-80-19(P)(A), Volume 1 and Supplements 1 and 2, Exxon Nuclear Methodoloav for Boiling Water Reactors - Neutronic Methods for Design 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 Methodolouv for Boiline Water Reactors: Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodolouv. Advanced Nuclear Fuels Corporation, Inc., Richland, WA, November 1990.

4.)

XN NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Methodolony for Boiling Water Reactors THERMEX: Thermal Limits Methodoloey Segmary Descriotion. 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 Comouter Proerim for Boiling Water Reactor Transient Analvlis, Advanced Nuclear Fuels Corporation, Richland, WA, August 1990.

6.)

ANF-1125 (P)(A) and Supplements, 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 '!, XCOBRA-T: A Computer Code for BWR Transient Thermal Hydraulic Core Analvtis, Exxon Nuclear Company, Inc., Richland, WA, February 19RT.

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.

9.)

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

Standard No.: GGNS-MS-48.0 Page:

4 of 15 Revision No.: 1 10.)

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

I 1.)

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

12.)

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

M2Ep for Plant Ooeration Within the Extended Ooerating Domain. Exxon Nuclear Company, Inc., Richland, WA, October 1986.

13.)

XN-NF-81-51(P)(A), LOCA-Seismic Structural Resoonse of an Exxon Nuclear Company BWR Jet Pumo Fuel Assembiv. 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 Pum_p Fuel Assembly. Exxon Nuclear Company, Inc., Richland, WA, August 1986.

15.)

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

16.)

XN-NF-82-07(P)(A), Revision 1, Exxon Nuclear Comoany ECCS Cladding Swelling and BypIpre Model. Exxon Nuclear Company, Inc., Richland, WA, November 1982.

17.)

XN-NF-80-19(A), Volumes 2,2A,2B, tnd 2C, Exxon Nuclear Methodoloev for Boilina Water Reactors EXEM BWR ECCS Evaluation Model. Exxon Nuclear Company, Inc.,

Richland, WA, September 1982.

18.)

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

19.)

ANF-1358(P)(A), Revision I and Correspondence, The Loss of Feedwater Heating in Boilinu Water Reactors, Siemens Power Corporation, Richland, WA, September 1992.

CURRENT CYCLE

REFERENCES:

20.)

EMF-93-050, Grand Gulf Unit 1 Cycle 7 Plant Transient Analysis. Siemens Power Comoration, Richland, WA, June 1993.

21.)

EMF-93-051, Grand Gulf Unit 1 Cycle 7 Reload Analysis. Siemens Power Corporation, Richland, WA, June 1993.

Standard No.: GGNS-MS-48.0 Page:

5 of 15 Revision No.: 2 22.)

ANF-92-190(P), Grand Gulf 1 ANF-1.6 Design Reoort. Mechanical. Thermal-Hydraulic and Neutronic Design for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies. Advanced Nuclear Fuels, Richland, WA, December 1992.

23.)

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

l 24.)

EMF-91-172, Grand Gulf Unit 1 LOCA Analysis for Sin 21e Loon Ooeration. Siemens Power Corporation, Richland, WA, October 1991.

l 25.)

ANF-88-152(P)(A) with Amendment I and Supplement 1, Generic Mechanical Desian for Advanced Nuclear Fuels 9X9-5 BWR Reload Fuel. Advanced Nuclear Fuels Corporation, Richland, WA, November 1990.

25A.) GEXI-94/00449, S.L. Leonard (SPC) to J.B. Lee (Ente.rgy), " Transmittal of Mechanical Design Review of the 9x9-5 Fuel Design for the Higher Peak Pellet Exposure Limit",

dated July 1,1994.

I 25B.) GEXI-94/00448, D.P. Austin (SPC) to J.B. Lee (Entergy), "Retransmittal of Justification for Increasing the 9x9-5 Peak Pellet Exposure Limit", dated July 1,1994.

CYCLE 6

REFERENCES:

26.)

ANF-91-080(P), Grand Gulf 1 ANF-1.5 Design Reoort. Mechanical. Thermal-Hydraulic l

and Neutronic Design for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies. Advanced Nuclear Fuels, Richland, WA, July 1991.

4 CYCLE 5

REFERENCES:

27.)

ANF-89-171(P), Volumes I and 2, Grand Gulf 1 ANF-1.4 Desien Reoort. Mechanical.

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

GENERAL

REFERENCES:

28.)

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

29.)

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

30.)

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

Standard No.: GGNS-MS-48.0 Page:

6 of 15 Revision No.: 0 4.0 DEFINITIONS l

1.

Average Planar Linear Heat Generation Rate (APLHGR) - the APLHGR shall be 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 specified bundle at the specified height divided by the number of fuel rods in the fuel bundle. (Technical Specification 1.3) 2.

Average Planar Exoosure - the Average Plans 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 rods in the fuel bundle.

(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.

Qre Operating Limits Reoort (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) 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 CPR 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) 1 l

1

Standard No.: GGNS-MS-48.0 Page:

7 of 15 Revision No.: 1 5.0 GENERAL REQUIREMENTS This section reports the Grand Gulf Cycle 7 core operating limits. These limits are taken from Reference 21 Sections 5.7,6.1.3, and 7.2.3. As discussed in Technical Specifications 2.1.1 and 2.1.2, these operating limits are applicable when thermal power is greater than 25% of rated power.

Average 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 MCPRr and MCPR limits at the indicated core p

flow and thermal power, for the exposure range, as shown in Figures 5.2, 5.3, 5.4, and 5.5.

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

The LHGR shall not exceed the limits shown in Figure 5.6 as multiplied by the smaller of either the flow-dependent LHGR factor (LHGRFACr) of Figure 5.7 or the power-dependent LHGR factor (LHGRFAC ) of Figure 5.8.

l p

Grand Gulf Unit 1 Cycle 7 Maximum Average Planar Linear Heat Generation Rate 15 --

~

14 --

13 - - (0,12.5)

(20,12.5) 12 ex 11 --

g Z

_J g 10 --

(50, 9.5)

E g.'

(55, 9.0)

N 8--

oo y

0 10 20 30 40 50 00 Average Planar Bumup (GWd/MTU) m I

mL Figure 5.1 MAPLHGR Limits for Grand Gulf Unit 1 Cycle 7

=

O

b a

a Grand Gulf Unit 1 Cycle 7 Flow-Dependent MCPR Limit 1.4 - -

1.35 -f 1.3 -

(30, 1.28) 1.25 --

(20,1.28) c is n.

1.2 --

o (65,1.20)

(105,1.20) z 1.15 -[-

~

1.1 --

4 ta 5

.k 1.05 -,-

n o

E Z

,,o, j

~,

0 10 20 30 40 50 60 70 80 90 100 110

~

g Core Flow, Percent of Rated z

m k

m A

ao Figure 5.2 Flow Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 7 o

. - _ _- -- _ _ _ _ _ ___ ___ x,

Grand Gulf Unit 1 Cycle 7 Power-Dependent MCPR Limit BOC to EOC-30 EFPD 2.4 - -

Core Flow > 50%

2.2 -

(25, 2.20) p(40, 2.10)

~

(25, 2.05) 2-

[

NK40.1.85)

n. 1.8 E

Core Flow s50%

lE 1.8 -f 6 0,1.41) 1.4 --

(40 148)

G:

~

1. m en 2~# E 1.2 F0,1.247 8

o(.

8-(90,1.23)

(100,1.21) i Z oZ l

1 l

. o P.

~

l m

i 0

10 20 30 40 50 80 70 80 90 100 110 Q

~

i g

l Core Power, Percent of Rated g

i k

I Figure 5.3 Power Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 7 for Exposures From BOC to 5

EOC-30 EFPD

Grand Gulf Unit 1 Cycle 7 Power-Dependent MCPR Limit EOC-30 EFPD to EOC 2.4 -

Core Flow > 50%

2.2 --

(25, 2.20) 2--

(25, 2.05) l(40,1.85)

_1.8--

n.

~E h

Core Flow s50%

3 1.6 --

9 0,1.41) 1.4 --

(40,1.49)

(90,1.28) j'a.om g

8 1'2 --

60 1-

)

8 '"1 57 1

l l

l l

l l

l l

l l

l P.,, : -

~

]

0 10 20 30 40 50 60 70 80 90 100 110 Core Power, Percent of Rated 3

I oc Figure 5.4 Power Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 7 for Exposures From EOC-30 EFPD To EOC

Grand Gulf Unit 1 Cycle 7 Power-Dependent MCPR Limit EOC to EOC+30 EFPD 2.4 -

l Core Flow > 50%

2.2 --

(25,2.20) A_

l(40,2.10) 2--

(25, 2.05) l(40,1.85)

_ 1.8--

o.i Core Flow s50*4 lE 1.6 --

1 (70, 1.41) 1.4 --

(#0* I 48) 3:

[* '

'_(100,1.28)

F7m (70,1.31)::

2.m.5 1.2 --

=

E

",_. a.

- o %

~

1 l

l l

l l

l l

l l

l l

~

0 10 20 30 40 50 60 70 80 90 100 110 Core Power, Percent of Rated 3

&G h;

6 Figure 5.5 Power Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 7 for Exposures From EOC To EOC+30 EFPD

--_-__-__-__-___-_______________1_______________-__________.

Grand Gulf Unit 1 Cycle 7 Linear Heat Generation Rate 16 -

is --

~

e 14 --

J

, (15.5,13.1)

". (0.0,13.1)

^ 12 --

E3

!$.11 --

Et:

! 10 -

J g..

gmm E.k-8 ]

(55.0, 8.0) 5.

8 7

x =x o

o

.. g.,..

o a

to is to 25 so as 40 4s so as 7

Average Planar Exposure (GWdIMT) m K

m L

oo Figure 5.6 LIIGR Limits for Grand Gulf Unit 1 Cycle 7

Grand Gulf Unit 1 Cycle 7 Flow-Dependent LHGR Factor 1.1 -

(70,1.000)

(90,1.000)

(110,1.000)

(68.0,1.0)

=

=

=

=

=

1_

(s0, 1.000)

(100,1.000) g,,

g o,.'

(50, 0.900)

U

(

(40,0.846) a:

0 30.8--

(20,0.792) =

(30. 0.792) 0.7 --

g,

(1 8

5.

gry o,

{k*

0 10 20 30 40 50 60 70 80 90 100 110 b

Core Flow, Percent of Rated g Zm I

Figure 5.7 Flow Dependent LHGR Factors for Grand Gulf Unit 1 Cycle 7 h

o D

Grand Gulf Unit 1 Cycle 7 Power-Dependent LHGR Factor 1.1 -

(70.1.0) 1--

(100,1.0)

$ 0.9 --

o

• C u.

DC O

T 0.8 --

(40,0.75)

(25,0.75) -

0.7 -

N p,5 g o,

0 10 20 30 40 50 60 70 80 90 100 110 Core Power, Percent of Rated G

m I

m 1.

Figure 5.8 Power Dependent LHGR Factors for Grand Gulf Unit 1 Cycle 7 g.

L

- --- - --- --