BSEP-96-0343, Application for Amends to Licenses DPR-71 & DPR-62,changing Method of Detecting Reactivity Anomalies as Well as Editorial Changes to Bases for TS 3/4.1.2

Application for Amends to Licenses DPR-71 & DPR-62,changing Method of Detecting Reactivity Anomalies as Well as Editorial Changes to Bases for TS 3/4.1.2

ML20135D613
Person / Time
Site:	Brunswick
Issue date:	12/04/1996
From:	Campbell W CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20135D615	List: BSEP-96-0343, Application for Amends to Licenses DPR-71 & DPR-62,changing Method of Detecting Reactivity Anomalies as Well as Editorial Changes to Bases for TS 3/4.1.2 ML20135D630
References
BSEP-96-0343, BSEP-96-343, NUDOCS 9612100057
Download: ML20135D613 (9)

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2~CP&L Corolina Power & Light Company Williom R. Campbe!!

PO Box 10429 Vice President Southport NC 28461-0429 Brunswick Nuclear Plant DEC 041996 SERIAL: BSEP 96-0343 10 CFR 50.90 TSC 95TSB36 i

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk i Washington, DC 20555 ,

BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 {

DOCKET NOS,50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62  ;

REQUEST FOR LICENSE AMENDMENTS REACTIVITY ANOMALIES Gentlemen:

In accordance with the Code of Federal Regulationt,, Title 10, Parts 50.90 and 2.101, Carolina l Power & Light Company hereby requests a revision to the Technical Specifications for the  !

Brunswick Steam Electric Plant (BSEP), Unit Nos.1 and 2. The revisions will change the -l method of detecting a reactivity anomaly described in Technical Specification 3.1.2 and l Surveillance Requirement 4.1.2. Actual core k,will be compared to predicted core k,instead of >

comparing actual and predicted control rod density to determine if a reactivity anomaly exists.

Additionally, editorial changes to the Bases for Technical Specification 3/4.1.2 are proposed to support the Technical Specification change.  ;

The proposed revisions have been incorporated in CP&L's license amendment application for the Improved Technical Specifications; therefore, CP&L requests that the revisions provided herein he issued prior to or concurrent with approval of the Brunswick Improved Technical Specifications (ITS). In order to support orderly revision of plant procedures and training, CP&L requests the proposed license amendments be effective on the date of issuance and implementation required within 30 days following issuance.

Carolina Power & Light Company is providing, in accordance with 10 CFR 50.91(b), Mr. Dayne H.

Brown of the State of North Carolina with a copy of the proposed license amendments.

Please refer any questions regarding this submittal to Mr. Mark A. Turkal at (910) 457-3066.

Sincerely, k

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William R. Campbell 9612100057 961204 PDR i

P ADOCK 05000324 l PDR Tel 910 457 2496 Fox 910 457 2803

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~ Document Control Desk BSEP 96-0343 / Page 2 ,

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Enclosures:

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1. Basis for Change Request  ;

2. 10 CFR 50.92 Evaluation l 3. Environmental Considerations

4. Page Change Instructions

! 5. Marked-up Technical Specification Pages - Unit 1 ,

l 6. Marked-up Technical Specification Pages - Unit 2

! 7. Typed Technical Specification Pages - Unit 1 i 8. Typed Technical Specification Pages - Unit 2 William R. Campbell, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, and agents of Carolina Power & Light >

Company.

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Notary (Sel)

My commission expires:b M jo1001 l

pc: U. S. Nuclear Regulatory Commission ATTN.: Mr. Stewart D. Ebneter, Regional Administrator  !

101 Marietta Street, N.W., Suite 2900 Atlanta, GA 30323-0199 Mr. C. A. Patterson i NRC Senior Resident inspector Brunswick Units 1 and 2:

U.S. Nuclear Regulatory Commission i ATTN.: Mr. David. C. Trimble, Jr. (Mail Stop OWFN 14B20) 11555 Rockville Pike l Rockville, MD 20852-2738 The Honorable H. Wells Chairman - North Carolina Utilities Commission

! P.O. Box 29510 l Raleigh, NC 27626-0510 Mr. Dayne H. Brown North Carolina Department of Environment, Health, and Natural Resources P.O. Box 27687 Raleigh, NC 27611-7687 i

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I ENCLOSURE 1 l

l BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS REACTIVITY ANOMAllES l

BASIS FOR CHANGES l

Current Recuirement Technical Specification 3.1.2 and associated Surveillance Requirement 4.1.2 require actual control rod density to be compared with predicted control rod density to determine if a reactivity anomaly exists. The reactivity difference between actual and predicted rod density shall not exceed 1% Ak/k.

Procosed Chance

The proposed change involves replacing the words " ROD DENSITY" with " core k,," in the l Limiting Condition for Operation statement of Technical Specification 3.1.2 and in Surveillance Requirement 4.1.2. Additionally, editorial changes to the Bases for Technical Specification 3/4.1.2 are proposed to support this Technical Specification change.

l l Basis For Procosed Chance 1

The purpose of the Reactivity Anomaly Technical Specification is to ensure nuclear methods used in the safety analyses are valid during the core operating cycle and to support the results of shutdown margin demonstrations by comparing the predicted core reactivity with the actual monitored core reactivity. The use of control rod density is an outdated method of measuring core reactivity.

Computer capabilities have greatly advanced since Brunswick Units 1 and 2 were originally licensed. The original plant process computer being used when the units were licensed could not independently solve or calculate core flux distributions or core eigenvalues. The original plant process computer was limited to using parameters such as core power, core flow, reactor l pressure, and control rod density. As a result, a correlation between control rod density and core l reactivity had to be developed for each fuel reload cycle by predicting core k,, and creating a control rod density trend. This trend was compared to actual control rod density as measured by the in-plant computer. Many assumptions were used by the original process computer when calculating core reactivity using control rod density (e.g., all control rods are of equal control rod )

worth and control rod notches associated with a single control rod are of equal reactivity worth).

These assumptions were based on a control rod pattern which was predicted over the core i cycle. Additionally, the accuracy of the original plant process computer was decreased when plant operating conditions resulted in an off-normal control rod pattern (e.g., asymmetric control rod pattern). Therefore, re-calculation of a predicted control rod density trend for control rod patterns different than the targeted control rod patterns assumed at the beginning of the cycle

was required (a new correlation between control rod density and core k,, had to be developed).

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I A new computer software program, which was installed with the new plant process computers for Brunswick Units 1 and 2 in 1994, gathers plant data and directly computes an actual core k,, i

! while minimizing assumptions. As a result, the monitored core k,, more accurately reflects actual  !

nuclear conditions in the core.

J As such, the proposed change is considered acceptable since the correlation between core reactivity and control rod density depends on predicting core k,,, core k,, can be readily  :

monitored with the new plant process computer, and core k,, can more accurately detect a l reactivity anomaly in the core (assumptione are minimized). In addition, the use of core k,,in Technical Specification 3/4.1.2 is consistent with Specification 3.1.2, " Reactivity Anomalies" of  ;

NUREG-1434 (Reference 1).

References:

1. NUREG-1434, " Standard Technical Specifications General Electric Plants, BWR/6," l Revision 1, April 1995.  !

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• l ENCLOSURE 2 l BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS

REACTIVITY ANOMALIES l

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i 10 CFR 50.92 EVALUATION

( The Commission has provided standards in 10 CFR 50.92 for determining whether a significant

! hazards consideration exists. A proposed amendment to an operating license for a facility  :

f involves no significant hazards consideration if operation of the facility in accordance with the l proposed amendment would not: (1) involve a significant increase in the probability or  ;

consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Carolina Power & Light Company has reviewed these proposed license '

amendment requests and believes that their adoption would not involve a significant hazards i consideration. The basis for this determination follows.

l l l 1. The proposed amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed license amendments modify the method of detecting a reactivity anomaly. The proposed license amendments allow using core k , to detect a reactivity anomaly instead of control rod density. The correlation between core reactivity and control rod density depends on predicting core k,,.

Core k,, can be readily monitored with the new plant process computer program and core k,, can more accurately detect a reactivity anomaly in the core (assumptions are minimized). A reactivity anomaly is not considered an initiator of any previously analyzed accident. As such, changing the method of detecting a reactivity anomaly will not increase the probability of any accident previously evaluated. Although, a reactivity anomaly could impact the consequences of a previously analyzed accident, the consequences of an event occurring using the proposed method of detecting a reactivity anomaly are the same as the consequences of an event occurring using the current method of detecting a reactivity anomaly. As a result, the proposed amendments do not involve a significant increase in the consequences of any accident previously evaluated.

2. The proposed amendments would not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed license amendments do not involve a physical modification to the plant. The proposed license amendments also continue to verify that the reactivity difference between predicted and actual are such that a reactivity anomaly does not exist. In addition, core k,, can more accurately detect a reactivity anomaly in the core (assumptions are minimized) and can be readily monitored with the new plant process computer program. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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3. The proposed license amendments do not involve a significant reduction in a margin of l safety. The proposed license amendments modify the method of detecting a reactivity i anomaly. The proposed license amendments allow using core k,, to detect a reactivity ,

anomaly instead of control rod density. The correlation between core reactivity and control i rod density depends on predicting core k . Core k , can be readily monitored with the new  !

plant process computer, and core k,, can more accurately detect a reactivity anomaly in the core (assumptions are minimized). Therefore, the proposed license amendments do not involve a significant reduction in a margin of safety. .

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ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS REACTIVITY ANOMAllES 1

I ENVIRONMENTAL CONSIDERATIONS 10 CFR 51.22(c)(9) provides criterion for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed  !

amendment to an operating license for a facility requires no environmental assessment if l l operation of the facility in accordance with the proposed amendment would not: (1) involve a l significant hazards consideration, (2) reFJit in a significant change in the types or significant increase in the amounts of any efflucats that may be released offsite, or (3) result in an increase j in individual or cumulative occupational radiation exposure. Carolina Power & Light Company l has reviewed this request and believes that the proposed arpendments meet the eligibility criteria l l

for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement of environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows.

1. These proposed license amendments do not involve a significant hazards consideration, i as shown in Enclosure 2. I i

2. The proposed license amendments do not result in a significant change in the types or a significant increase in the amounts of any effluent that may be released offsite. The proposed license amendments do not introduce any new equipment nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform. The proposed license amendments do not alter the function of existing equipment and will ensure that the consequences of any previously evaluated accident do not increase. Therefore, CP&L has concluded that there will not be a significant increase in the types or amounts of any effluent that may be released offsite and, as such, does not involve irreversible environmental consequences beyond those already associated with normal operation.

3. These proposed license amendments do not result in an increase in individual or cumulative occupational radiation exposure.

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ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2

NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS I REACTIVITY ANOMAllES 1

i PAGE CHANGE INSTRUCTIONS

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! ENCLOSURE 5 l i

4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 l

NRC DOCKET NOS. 50-325 AND 50-324  :

OPERATING LICENSE NOS. DPR-71 AND DPR-62 '

REQUEST FOR LICENSE AMENDMENTS '  !

REACTIVITY ANOMAllES I i

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