ML20084A034

Rev 3 to GGNS COLR Sr

ML20084A034
Person / Time
Site:	Catawba
Issue date:	05/22/1995
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20084A030	List: ML20084A028 ML20084A034
References
GGNS-MS-48.0, GGNS-MS-48.0-R03, GGNS-MS-48.0-R3, NUDOCS 9505300211
Download: ML20084A034 (21)

Text

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Standard No.: GGNS-MS-48.0 Revision: 3 Date:

Grand Gulf Nuclear Station Core Operating Limits Report Safety-Related

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l 9505300211 950522 3 PDR ADOCK 0500 P

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. Standard No.: GGNS-MS.48.0

, Page: il of vi l Revision No.: 3 )

GRAND GULF NUCLEAR STATION NUCLEAR PLANT ENGINEERING REVIEW AND APPROVAL SHEET l i

STANDARD NO.: ._SGNS.MS-48.0 REVISION: 3 l

STA'NDARD TITLE: Core Operatina_ Limits Report I 1

This document specifies items related to nuclear safety YES [X] NO[] 1 Signatures certify that the above standard was originated, verified, reviewed or waived and approved as noted bei S  :

ORIGINATED BY: b 2d M DATE: _ d VERIFIED BY: [ DATE: X /fS~

// / '

REVIEWED BY: /M aM DATE: M&/ff Cognizant Group Supenisor NPE SECTION _ REVIEWED BY REVIEW WAIVED BY DATE ELECTRICAllI&C ,, 4 _ e) / 9 i -

c.

t-MECHANICAUCIVIL ,_ tf Sf6 [W M (

SAFETY ANALYSIS /44 k f/#Ms-l ANII: M/A DATE:

(Insert N/A ilnot applict.ble)

I APPROVED BY: 8 M [#2O e DATE: 5-// 7/95-I Responsible Manager I

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Standard No.: GGNS-MS-48.0 Page; iii of vi Revision No.: 3 SAFETY EVALUATION APPLICABILITY REVIEW FORM A) Document Evaluated: Standard GGNS-MS-48.0. Revision 3 II) Description of the Proposed Change: Per GNRI-934)8. Amendment 106 to the GrcrdGulf Operating License. Enterav committed to removina certain rear ohysics caramaeces frota & Technical Soccifications and olacing them in a seaarate recort orecared for each fuel cycle. St=hed GGNS-MS-48 0 is the Core Occrating Limits Renart (CO1R) and establishes thee cararneters. Revisions 3 to GGNS-MS-48.0 reports the Cvele 8 core operating limilt PRE-SCREENING Check the applicable boxes below. If any of the boxes are checked, neither a safety evaluation applicability revkw i

nor a safety evaluat on 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.

2.

Ig@ approved safety evaluation covering all aspects of this subject alread Ra rence SE# The Cycle 8 core ooerating limits have been evaluated in the Cycle 8 reload safgy cyrtation. 95-0022-RQQ,

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

Referend::

_ The change is an FSAR change that meets the exclusion criteria outlined in Site Directive G4.803 Safety Evaluation Aeolicability Retiew 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 actisi;y represent:

(1) A change to the facility whict alters, or has the potential to alter, the information, operation, function or ability to perform the function of a system, structure or component desenbcd in the SAR?

YES _ Explain:

NO ._.

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. Standard No.: GGNS-MS-48.0

. Page: iv of vi Revision No.: 3 (2) A change to a procedure which alters, or has the potential to alter, a procedure described, outlined or summarized in the SAR?

YES _ Explain:

NO _

(3) A test or experiment not descrited in the SAR or which requires that a system be operated in an abnormal manner that is not described or previously analyzed in the SAR?

YES _ Explain:

NO ,_,

PREPARERV\ AbTD & bd .5khb Job'6de ' Dkte REVIEWERhb - . h Stre. Attr'. F/g/fr Name lobTitle Date If the preparer performed an applicability review, the reviewer should check Imlow to indicate by which means the independent review reached the same conclusions.

/ Reviewed the applicability review documentation.

_ Completed an ir '- g-- -f- nt applicability review.

_ Performed a verbal review with the preparer, i

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, Standard No.: GGNS-MS-48.0

, Page: v of vi Revision No.: 3 l

REVISION STATUS SHEET STANDARD REVISION

SUMMARY

4 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 3 Issued for use PAGE REVISION STATUS PAGE NO. REVISION PAGE NO. REVISION i-vi 3 8 3 1 3 9 3 2 3 10 3 3 3 11 3 4 3 12 3 5 3 13 3 6 3 14 3 7 3 15 3 APPENDIX /ATTACIIMENT STATUS

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Standard No.: GGNS-MS-48.0 e Page: vi of 'vi Revision No.: 3 TABLE OF CONTENTS 1e e e et e e444e et ce e se 6e00440 e e eteeeette0645446006046 ettete see stette 68000eteete eget este se sette see se 2e ceeed to e e4e eea 600eDee et see cepeeses eteette e etes996944069 44600e et seee Seceettee easset ette sees e

e Det see te eteet e030966043e ee teeee se e e e etee e spees ce e e egesese estegee see eseesee646900 tese Gees 4 4 ese e se sete 448esett to se e se $$eee este e e ettegeetee 960 speceseece eeece e se se esteeggeteceeeetg ee .

Se ette e es teette ce toeg spesete ese e stece 840e4e4 ee ceto seeeeete sees sees l

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. Standard No.: GGNS-MS-48.0

. Page: 1 of 15 Revision No.: 3 1.0 PURPOSE On October 4,1988, the NRC issued Generic Letter 88-16 (Reference 28) encouraging licensees 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 world 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 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).

The COLR is controlled via Mechanical Standard GGNS-MS-48.0. This standard is revised accordingly for each fuel cycle or remaining portion of a fuel cycle. Revision 3 of the COLR reports the Cycle 8 core operating limits.

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. Standard No.: GGNS-MS-48.0 Page: 2 of 15 Revision No.: 3 2.0 SCOPE A1 defined in Technical Specification 1.1, the COLR is the GGNS document that provides the l core operating limits for the current fuel cycle. This document is prepared in accordance with Technical Specification 5.6.5 for each reload cycle using NRC-approved analytical methods. l The Cycle 8 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 3.2.2), and l

3. the Linear Heat Generation Rate (LHGR) limit for each fuel type including the power and flow dependent parametric adjustment factor curves, LHGRFAC p respectively. (Technical Specification 3.2.3) and LHGRFACr, l The cycle-specific MCPR safety limits are documented in Technical Specification 2.1.1.2. l l

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Standard No.: GGNS-MS-48.0 1 Page: 3 of 15 Revision No.: 3

3.0 REFERENCES

This section contains the methodology and cycle-specific references used in the safety analysis of Grand Gulf Cycle 8. The supplements and revisions of the current analytical methodology refeiences are included below in accordance with Technical Specification 5.6.5, Core Operating Limits Report.

METHODOLOGY

REFERENCES:

1.) XN-NF-79-71(P), Revision 2 including Supplements 1, 2, and 3, Exxon Nuclear Plant Transient 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 Boiline 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 Methodolovv for Boilina Water Reactors: Benchmark Results for the CASMO-3G/MICROBURN-B Calculation Methodology. Advanced Nuclear Fuels Corporation, Inc., Richland, WA, November 1990.

4.) XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon - __;ar Methodolorv for Boiling Water Reactors THERMEX: Thermal Limits Methodoloav Summary Descriotion. Exxon Nuclear Company, Inc., Richland, WA, January 1987.

S.) ANF-913 (P)(A), Volume 1, Revision I and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Rea_q1or Transient Anaivsis.

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

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 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 Ramps.

Exxon Nuclear Company, Inc., Richland, WA, May 1982. Approved by NRC letter dated August 28,1990.

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.

. Standard No.: GGNS-MS-48.0 i

. Pege: 4 of 15 Revision No.: 3 10.) XN-NF-85-74(PXA), RODEX2A SWRk Fuel Rod Thermal-Mechanical Resoonse l Evaluation Model, Exxon Nuclear Company, Inc., Richland, WA, August 1986.

11.) XN-CC-33(PXA), Revision 1, HUXY A Genarnlimi Multirod Hantun Code with 10CFR50 Aopendix K Heatuo Ootion. Enon Nuclear Company, Inc., Richland, WA, November 1975.

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

MCPR_ for Plant Operation Within the Extended Ooeratina Domain Exxon Nuclear 2

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

13.) XN-NF-81-51(PXA), LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assemb.]y, Advanced Nuclear Fuels Corporation, Richland, WA, May 1986.

14.) XN-NF-84-97(PXA), 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(PXA), Revision 1, Exxon Nuclear Company ECCS Claddinn Swelling and Rupture Model, Exxon Nuclear Company, Inc., Richland, WA, November 1982.

17.) XN-NF-80-19(A), Volumes 2, 2A, 2B, and 2C, Exxon Nuclear Methodolony for Boilinn Water Reactors EXEM BWR ECCS Evaluation Model. Exxon Nuclear Company, Irs.,

Richland, WA, September 1982. j 18.) XN-NF-79-59(PX^), Methodolony for Calculation of Pressure Droo in BWR Fuel Assemblies. Exxon Nuclear Company, Inc., Richland, WA, November 1983.

19.) ANF-1358(PXA), Revision 1 and Correspondence, The Loss of Feedwater Heating in Boilina Water Reactors, Siemens Power Corporation, Richland, WA, September 1992. 1 CURRENT CYCLE

REFERENCES:

20.) EMF-94-187, Grand Gulf Unit 1 Cycle 8 Plant Transient Analysis, Siemens Power Corporation, Richland, WA, December 1994.

21.) EMF-94186, Grand Gulf Unit 1 Cycle 8 Reload Analysis, Siemens Power Corporation, Richland, WA, December 1994.

Standard No.: GGNS-MS-48.0

. Page: 5 of 15 .

Revision No.: 3 22.) EMF-94-105, Grand Gulf 1 ANF-1.7 Desian Report. Mechanical. Thermal-Hydranli< and l  ;

Neutronic Desian for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies, Advanced Nuclear Fuels, Richland, WA, July 1994. l 23.) ANF-86-133, Revision 4, Princioal ECCS and Plant Transient Analysis Parpeneters Grand Gulf Unit 1, Advanced Nuclear Fuels Corporation, Richland, WA, June 1991.

24.) EMF-91-172, Grand Gulf Unit 1 LOCA Analysis for Sinnie Loon Operation. Siemens Power Corporation, Richland, WA, October 1991.

25.) ANF-88-152(P)(A) with Amendment I and Supplement 1, Generic Mechanical Denien 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 (Entergy), " Transmittal of Mechanical Design Review of the 9x9-5 Fuel Design for the Higher Peak Pellet Exposure Limit",

dated July 1,1994.

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 7

REFERENCES:

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

CYCLE 6

REFERENCES:

27.) ANF-91-080(P), Grand Gulf 1 ANF-1.5 Design Report. Mechanical. Thermal-Hydraulic and Neutronic Desian for Advanced Nuclear Fuels 9X9-5 Fuel Assemblies. Advanced Nuclear Fuels, Richland, WA, July 1991.

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. l 30.) 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 15 Revision No.: 3 4.0 DEFINITIONS-

1. Averson 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

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rates for all the fuel rods in the specified bundle at the specified height divided by the ,

number of fuel rods in the fuel hund:e. l l i

2. Averane Planar Exoosure - 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 rods in the fuel bundle. l l

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. l

4. Core Operatina 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 5.6.5. l S. 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. l

6. Minimum Critical Power Ratio (MCPR) - the MCPR shall be the smallest CPR which exists in the core. l

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. l 8

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. Standard No.: GGNS-MS-48.0 of

, Page: 7 15 Revision No.: 3 5.0 GENERAL REQUIREMENTS This section reports the Grand Gulf Cycle 8 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 3.2.1, 3.2.2, and 3.2.3, these operating limits are applicable when the core thermal power is greater than 25%

of rated power.

Avernae Planar Linear Heat Generation Rates (Technical Specification 3.2.1)

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

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

Minimum Critical Power Ratio (MCPR)(Technical Specification 3.2.2) l The MCPR shall be equal to or greater than the MCPH f and MCPR p limits at the indicated core flow and thermal power, for the exposure range, as shown in Figures 5.2, 5.3, 5.4, and 5.5.

Linear Heat Generation Rate (LHGR)(Technical Specification 3.2.3) l 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 (LHGRFACp) of Figure 5.8.

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Cycle 8 Maximum Average Planar Linear Heat Generation ~

Rate 15 - -

14 --

13 - :(0,12.5)

_ (20,12.5)

=  : -

$ 12 --

E e 11 --

g I

J g 10 --

lE 9--

(55, 9.0) yg B--

p&

9 G ga R 7 l 8E Z

l l l l  : Z

,0 10 20 30 40 50 60 -o?

Average Planar Bumup (GWd/MTU) @

z Y2 K

m

' i Figure 5.1 Cycle 8 MAPLIIGR Limits 5

Cycle 8 Flow-Dependent MCPR Limit .

1.4 - -

1.35 --

1.3 -- (30,1.28)

= -

(20,1.28)

E tt:

n. 1.2 -- -  :

O (65,1.20) (105,1.20) lE 1.15 --

1.1 --

1.05 -- 2 s

s.

w 5'

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e:

1  :  :  : l l l l  :  :  :  :

$ cQ 9

0 10 20 30 40 50 60 70 80 90 100 110 9 Core Flow, Percent of Rated w o

Gz Q m

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Figure 5.2 Cycle 8 Flow-Dependent MCPR Limits .=

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P Cycle 8 Power-Dependent MCPR Limit -

BOC to EOC-30 EFPD 2.4 - -

Core Flow > 50%

2'2 -- (25, 2.20),

(40,2.10)

(25,2.05) 2--

(40,1.85)

E n.

$ 1.6 -- Core Flows 50%

iw (70,1.41)

(40,1.49) 1.4 --  ;,

(90,1.23) e = _ W ;y La 1.2 -- (70,1.25) -

y, 5'

u g

a.

1 l l l l l l l l l l 2 57 0 10 20 30 40 50 60 70 80 90 100 110 . oPm Core Power, Percent of Rated _ g Z

cn Figure 5.3 Cycle 8 Power-Dependent MCPR Limits for Exposures From BOC to EOC-30 EFPD

Cycle 8 Power-Dependent MCPR Limit EOC-30 EFPD to EOC .

2.4 - -

Core Flow > 50%

(25.2.20) i (40,2.10)

(25,2.05) 2--

_ i (40,1.85)

c. 1.8 --

E 0.

Core Flow g 50%

h 1.6 --

! E- (70,1.41) 1.4 __ (40,1.49) ,,

(90,1.26)

_ M *U M (70.1 28)  : 3, 1.2 __ (100.1.24) $

y.{-

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10 20 30 40 50 60 70 80 90 100 110

[

f' Core Power, Percent of Rated o

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Figure 5.4 Cycle 8 Power-Dependent MCPR Limits for Exposures From EOC-30 EFPD To EOC 6

Cycle 8 Power-Dependent MCPR Limit EOC to EOC+30 EFPD 2.4 - -

Core Flow > 50%

2.2 -- (25,220)

(25,2.05) (40,2.10)

E 1.8 -- (40,1.85)

E n.

Core Flow s 50%

llE 1.6 --

(40,1.49)' (70,1.41)

(95,128)

N N to 1.2 -- GO,1.30)

(80 1.28) (100,127) h,- 8.

o_a 5'

Z NZ a.

1  :  : l l l l l l  :  :  : .o o 0 10 20 30 40 50 60 70 80 90 100 110 ' ,

Core Power, Percent of Rated ~

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co b

F.:gure 5.5 Cycle 8 Power-Dependent MCPR Limits for Exposures From EOC To EOC+30 EFPD

Cycle 8 Linear Heat Generation Rate '

16 - -

15 --

14 --

- (15.5,13.1) 13 "

(0.0,13.1)

E 12 --

3 5 11 __

a:

O I 10 --

2 9 .

8-- 29 (55.0, 8.0) @y 3 [

7-- j' {

6 l l l l Z

9 Cz l l l l l

• l l o .9 0 5 10 15 20 25 30 35 40 45

~

50 55 w'o U O Average Planar Exposure (GWd/MT) z m

I m

o Figure 5.6 Cycle 8 LIIGR Limits

Cycle 8 Flow-Dependent LHGR Factor -

1.1 - -

,.. - (68, i.03 _ f 0_ (S ; S _

( o'N _

(80,1.0) (100,1.0)

(60,0.954 E 0.9 -- (50, 0.900)

O g (40,0.846)

O 3 0.8 -- (20,0.792) = _

(30, 0.792) 0.7 --

???

s. 8 g.% g.

0.6  :  :  :  :  :  :  :  :  :  :  : m 21.

0 10 20 30 40 50 60 70 80 90 100 110 f[ O -

Core Flow, Percent of Rated m' -

• o Z

Y)

I Figure 5.7 Cycle 8 Flow-Dependent LIIGR Factors [

o

c Cycle 8 Power-Dependent LHGR Factor .

1.1 - -

(70,1.0) 1--

=

(100,1.0)

$ 0.9 --

o u.

E O

] 0.8 --

(40,0.75)

(25,0.75) _ _

0.7 --

???

s. 8 g.% g-0.6  :  :  :  :  :  :  :  :  :  :  : m 3.

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

[U Z Core Power, Percent of Rated "%

o

• o z

Y' Figure 5.8 Cycle 8 Power-Dependent LIIGR Factors y

"$li O.

..