2CAN039502, Application for Amend to License NPF-6,requesting to Utilize Colr,Per GL 88-16 at Unit 2

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Application for Amend to License NPF-6,requesting to Utilize Colr,Per GL 88-16 at Unit 2
ML20081F791
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 03/17/1995
From: Yelverton J
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20081F792 List:
References
2CAN039502, 2CAN39502, GL-88-16, NUDOCS 9503220270
Download: ML20081F791 (8)
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Entergy operations,Inc.

1448 3 R 333 j

  1. = ENTERGY Ofja,80i 2 l

Jerry W. Yefverton '

Vce Prcr. dent operams mo i March 17,1995 2CAN039502 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555

Subject:

Arkansas Nuclear One- Unit 2 -

Docket No. 50-368 License No. NPF-6 Proposed Technical Specification Change Request Concerning the Core Operating Limits Report  ;

Gentlemen:

By letter dated July 22,1993 (2CAN049301), Entergy Operations requested and received a technical specification change request to utilize a Core Operating Limits Report (COLR) per Generic Letter 88-16 at Arkansas Nuclear One, Unit 2 (ANO-2). Entergy Operations has subsequently determined that the specific value listed in technical specification 3.2.4.b is a cycle-specific value. Therefore, this proposed amendment removes this value from the technical cpecifications and places it in the COLR. .

The administrative controls section is modified to include a reference to technical specification 3.2.4.b with the appropriate methodologies. In addition, a methodology listed in the administrative controls section is being superceded by another NRC-approved methodology.

The proposed ANO-2 Technical Specifications are attached.

These proposed changes have been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that these changes involve no significant hazards considerations. The bases for these determinations are included in the attached submittal.

Entergy Operations requests that the e&ctive date for this change be upon issuance. This change is associated with the reload for the upcoming fuel cycle. Although this request is neither exigent nor emergency, your prompt review is requested prior to startup from the next ANO-2 refueling outage (2R11) which is currently scheduled to begin September 22,1995.

220010 kESOSEE!!!!d?6, 'p ' '

PDR I

U. S. NRC

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March 17,1995

. 2CANO39502 Page 2 Very truly yours, 6U JWY/nbm Attachments 1

1 To the best of my knowledge and belief, the statements contained in this submittal are true. .

SUBSCRIBED AND SWORN TO before me, a Notary Public in and for MW2cw County and the State of Arkansas, this 19 day of 'lYlayL ,1995.V en % Dl cOaA orrey seu  ;

Notary P)iblic U JUANA M.TAPP My Commission Empires /l4-S000 *EIssc$u"wT" My Commesion Expres 11-8 2000 cc: Mr. Leonard J. Callan Regional Administrator U. S. Nuclear Regulatory Commission Region IV t

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One 1448 S. R. 333 Russellville, AR 72801 Mr. George Kalman NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Ms. Greta Dicus Arkansas Department ofHealth Division ofRadiation Control and Emergency Management l 4815 West Markham Street Little Rock, AR 72205 i

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2CANO39502 PROPOSED TECHNICAL SPECIFICATION

  • l RESPECTIVE SAFETY ANALYSES i

IN THE MATTER OF AMENDING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.  ;

ARKANSAS NUCLEAR ONE, UNIT TWO I DOCKET NO. 50-368  !

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( Attachment to

.t 2CANO39502 l

. Page1 of4 .l DESCRIPTION OF FROPOSED CHANGES When the core operating limit supervisory system (COLSS) is m service and neither control element assembly calculator (CEAC) is operable, technical specification 3.2.4.b requires the >

departure from nucleate boiling ratio (DNBR) limit to be maintained. This is accomplished by maintaining the COLSS calmiaW core power less than or equal to the COLSS calculated  !

core power operating limit based on DNBR decreased by 13%. The specific value of 13% of  ;

technical speci6 cation 3.2.4.b has been determined to be cycle-specific and is being relocated l to the COLR.

Technical specification 6.9.5.1 lists the analydcal methods used to determine the core operating limits addressed by the individual technical specifications. A reference to technical j

!- specification 3.2.4.b is being added to the methodologies described by technical specifications ~ '

6.9.5.1.1 and 6.9.5.1.9. As a result, a clarification is being made to specification 6.9.5.1.3 to specifically reference specifications 3.2.4.c and 3.2.4.d.  ;

The methodology specified by technical specification 6.9.5.1.3, " Statistical Combination of Uncertainties," CEN-139(A)-P is being superceded with an NRC-approved methodology, which is entitled " Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A,  !

Revision 01-P-A, dated May 1988.

l BACKGROUND I Generic Letter 88-16 dated October 4,1988 (0CNA108809), " Removal of Cycle-Specific l Parameter Limits from Technical Specifications," allowed licensees to remove cycle-specific parameters from technical specifications and place them in a Core Operating Limits Report (COLR) provided the limits are developed using an NRC-approved methodology. This l change was requested snd subsequently granted in Amendment Number 157 (2CNA049402),

dated April 11, 1994. Subsequent to this relocation of cycle-specific parameters from '!

technical specifications to the COLR, ANO has determined that the value of 13% specified in  ;

technical speci6 cation 3.2.4.b is cycle-specific.

1 DISCUSSION OF CHANGE l Each accident analysis addressed in the ANO-2 Safety Analysis Report ( SAR) is considered in  !

the reload report for each particular cycle. This is performed with respect to changes for any (

cycle-specific parameters to ensure that thermal performance during hypothetical transients is acceptable. For core reloads, the margins of safety for fuel system design, nuclear design, and thermal-hydraulic design are addressed in the reload report. The applicable limits and setpoints are determined to be within allowable limits and requirements for acceptable operation fcr a particular cycle.

. The DNBR margin-related limit of technical speci6 cation 3.2.4.b, which will be relocated to the COLR, is determined each cycle for the CEACs in a degraded condition. This limit is dependent upon the cycle-specific margin requirements for the given conditions. The margin I

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/ Attachment to 2CANO39502

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. L l- requirements are affected by the reload core design. When COLSS is in service and neither CEAC is operable, technical specification 3.2.4.b requires the DNBR limit to be maintained by maintaining the COLSS calculated core power less than or equal to the COLSS calculated  ;

core power operating limit based on DNBR decreased by 13%. In order to obtain flexibility .

in optimizing fuel cycle designs and maximize operating margins, this value may be changed from cycle to cycle. This value is derived by utilizing NRC-approved methodologies specified i in technical ap-ih=* ions 6.9.5.1.1 and 6.9.5.1.9.  !

In order to improve plant operating performance and flexibility and reduce the potential for unnecessary reactor trips by reducing the overall uncertainty factors applied in the COLSS and i core protection calculator (CPC) system, a methodology change is needed for technical i specification 6.9.5.1.3 concerning the statistical combination of uncertainties. The current ,

methodology will be replaced with the " Modified Statistical Combination of Uncertainties (MSCU)," CEN-356(V)-P-A, Revision 01-P-A, dated May 1988. This methodology has been approved for use at Palo Verde Unit 1 in a safety evaluation dated October 21,1987. This ,

methodology is also currently in use at Waterford-3 and SONGS. Rigorous, statistically justified methods are used for statistically combining uncertainties to obtain overall uncertainty ,

factors. The overall uncertainty factors are used to determine the limiting safety system setting and the linuting condition for operation for the COLSS and CPC system.

The uncertainties involved in the current statistical combination of uncertainties methodology are divided into two categories. The uncertainties within each category are combined statistically and a 95/95 probability / confidence level is generated for each category. The resultant uncertainties of the two categories are effectively combined in a deterministic manner to the 95/95 probability / confidence level.

The MSCU methodology statistically combines the uncertainty components which were previously combined deterministically. Also, the statistical treatment of several uncertainty - i components is modified so that the overall uncertainty factors can be calculated and applied as i a function of burnup, axial shape index, and power in the COLSS and CPCs. From the Palo Verde safety evaluation, the Staff determined that the resultant penalties applied to the '

COLSS power operating limit and the CPC DNBR and local power density calculations utilizing the MSCU methodology adequately incorporated all uncertainties at the 95/95 probability / confidence level and that the MSCU methodology was acceptable for use with the Palo Verde digital monitoring and protection system. The MSCU methodology will be applied at ANO-2 in a manner similar to Palo Verde.

.- Attachment to 2CANO39502

. Page 3 of 4 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION ,

I An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(c). A discussion of these standards as they relate to this amendment request follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The removal of the cycle-dependent value for the departure from nucleate boiling ratio (DNBR) reduction from technical specifications and placing it into the Core Operating Limits ,

Report (COLR) has no impact on plant operation or accident analyses. The proposed change is considered to be administrative in nature. Technical specifications will continue to require operation within the core operational limits for each cycle reload calculated by the approved 1 reload design methodologies. The appropriate actions required if limits are violated will remain in the technical specifications. The reload report presents the results of a cycle-specific evaluation of accidents and transients addressed in the ANO-2 Safety Analysis Report (SAR).

The cycle-specific evaluation demonstrates that changes in the fuel cycle design and the corresponding COLR do not involve a significant increase in the probability or consequences of an accident previously evaluated.

i The Modified Statistical Combination of Uncertainties (MSCU) methodology statistically combines uncertainties to at least a 95/95 probability / confidence level. The proposed change to reference the MSCU is administrative in nature. The currently referenced methodology is being replaced with a more recently approved methodology which has been determined to be applicable to ANO-2. The new methodology has been independently reviewed and approved by the NRC. This change does not impact either the manner in which the operating margin to limits on linear heat rate and DNBR is maintained or the manner in which the CPCs respond to transients and provide trips. Therefore, this change does not adversely impact transient ,

analysis assumptions or results. In addition, the physical design or operation of the plant is I not impacted by this change. The safety analyses will continue to be performed utilizing l NRC-approved methodologies and specific reload changes will be evaluated per 10CFR50.59.

Therefore, these changes do pm involve a significant increase in the probability or consequences of any accident previously evaluated.

Criterion 2- Does Not Create the Possibility of a New or Different Kind of Accident from any Previously Evaluated.

The proposed change to relocate the cycle-specific value for the DNBR reduction from technical specifications to the COLR is administrative in nature. No change to the design, configuration, or method of operation of the plant is made by this change. This parameter will be determined using NRC-approved methods. Technical specifications will continue to require operation within the required core operating limits and appropriate actions will be

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La Attachment to 2CAN039502 Page 4 of4

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' taken if the limits are exceeded. The relocation of a cycle-specific parameter does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to reference the NRC-approved MSCU methodology is administrative in nature. The currently referenced methodology is being replaced with a more recently approved methodology which has been determined to be r.pplicable to ANO-2. No physical alterations of plant configuration, changes to plant operating procedures, or operating parameters are proposed. The safety analyses are still performed utilizing NRC-approved methodologies and specific reload changes will be evalented per 10CFR50.59. No new equipment is being introduced, and no equipment is being operated in a manner inconsistent with its design.

Therefore, these changes do not create the possibility of a new or different kind of accident from any previously evaluated.

e Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety. .

Existing technical specification operability and surveillance requirements are not reduced by the proposed change to relocate the cycle-specific value for DNBR reduction to the COLR. '

The development of limits for a particular cycle will continue to conform to methods described in NRC-approved documentation. Technical specifications will still require that the core be operated within these limits and specify appropriate actions to be taken if the limits are i violated. The cycle-specific COLR limits for future reloads will be developed based on NRC- ,

approved methodologies. Each reload undergoes a 10CFR50.59 safety review to assure that '

operation of the unit within the cycle-speci6c limits will not involve a significant reduction in a margin of t,afety.

The proposed change to reference the MSCU methodology is administrative in nature. The currently referenced methodology is being replaced with a more recently approved methodology which has been determined to be applicable to ANO-2. The resultant overall uncertainty factors using the MSCU methodology are determined and applied to at least the same 95/95 probability / confidence level as the overall uncertainty factors using the currer.t methodology. NRC review and approval of the methodologies used to perform the cycle-specific reload analyses is not affected by this change. The safety analyses are still performed utilizing NRC-approved methodologies and specific reload changes will be evaluated per 10CFR50.59.

i Therefore, these changes do not involve a significant reduction in the margin of safety.

Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does not involve a significant hazards consideration.

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  • 6 PROPOSED TECHNICAL SPECIFICATION CHANQE_S l

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DNRR MARGIN LIMITING CONDITION POR OPERATION  :

- 3.2.4 ~The DNBR lbsit'shall be maintained by one of the following

_ __ l methods:

a. . Maintaining 00LSS calculated core power less than or. equal  ;

to COLSS calculated core power operating limit based on DNBR (when ,

COLS8 is in service, and at least one CEAC is operable); or  ;

b. Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating lLait based on DNBR i decreased by the value specified in the CORE OPERATING LIMITS REPORT (when OOLSS is in service and neither'CEAC is operable);

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c. Operating within the region of acceptable operation specified in-the CORE OPERATING LIMITS _ REPORT using any operable CPC channel (when COLSS is out of service and at least one CEAC is operable);

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d. Operating within the region of acceptable operation specified in~

the CORE OPERATING LIMITS REPORT using any operable CPC channel -

(when 00LSS is out of service and neither CEAC is operable).

APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER. ,

ACTION:

a. With COLSS in service and the DNBR limit not being maintained as indicated by COLSS calculated core power exceeding the COLSS. 'i calculated core power operating limit based on DNBR,.within 15 minutes initiate corrective action to reduce the DNBR to within the limits and either:
1. Restore the DNBR to within its lLuits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the initiating event, or
2. Reduce THERMAL POWER to less than or equal to 20% of RATED ,

THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With 00LSS out of service and the DNBR limit not being maintained as l indicated by operation outside the region of acceptable operation specified in the CORE OPERATING LIMITS REPORT, either
1. Restore the DNBR to within its limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the initiating event, or _ i
2. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ARKANSAS - UNIT 2 3/4 2-5 Amendment No. 44,49,99,494,469

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, ; i~ ADMINISTRATIVE CONTROL

  • CORE OPERATING LIMITS REPORT l]

6.9.5 The core operating limits shall be established and docume'nted in the- -i CORE OPERATING LIMITS REPORT prior to each reload cycle or any remaining part of a reload cycle.

6.9.5.1- The analytical methods used to determine the core operating limits addressed by the individual Technical specifications shall be those previously reviewed and approved by the NRC for use at ANO-2, specifically:

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1) "The ROC 8 and DIT Computer Codes for. Nuclear Design", CENPD-266-P-A, April 1983 (Methodology for Specifications 3.1.1.1 and 3.1.1.2.

for shutdown Margins, 3.1.1.4 for MTC, 3.1.3.6 for Regulating CIA Insertion Limits, and 3.2.4.b for DNBR Margin).

2) "CE Method for control Element Assembly Ejection Analysis,"

(. CENPD-0190-A, January 1976 (Methodology for Specification ~

l 3.1.3.6 for Regulating CEA Insertion Limits and 3.2.3 for

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Azimuthal Power Tilt).  !

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3) " Modified. Statistical Combination oi Uncertainties, CEN-356(V)-P-A, Revision 01-P-A, May 1988 (Methodology for specification 3.2.4.c l and 3.2.4.d for DNBR Margin and 3.2.7 for ASI).
4) " Calculative Methods for the CE.Large Break LOCA Evaluation Model,"

CENPD-132-P, August 1974 (Methodology for specification.3.1.1.4 J for MTC, 3.2.1 for Linear Meat Rate, 3.2.3 for Azimuthal Power Tilt, and.3.2.7 for ASI).

5) " Calculational Methods for the CE Large Break LOCA Evaluation Model,"

CENPD-132-P, Supplement 1, February 1975 (Methodology for Specification 3.1.1.4 for MTC,~3.2.1 for Linear Meat Rate, 3.2.3 for Azimuthal Power. Tilt, and 3.2.7 for ASI).

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6) " Calculational Methods for the CE Large Break LOCA Evaluation Model,"

CENPD-132-P, supplement 2-P, July 1975 (Methodology for specification 3.1.1.4 for MTC, 3.2.1 for Linear Meat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).

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7) " Calculative Methods for the CE Large Break LOCA Evaluation Model  !

for the Analysis of CE and W Designed NSSS," CEN-132, supplement 3-P-A, June 1985 (Methodology for Specification ,

3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).

8) " Calculational Methods for the CE Small Break LOCA Evaluation Model,"

CENPD-137-P, August 1974 (Methodology for specification 3.1.1.4 for MTC, 3.2.1 for Linear Meat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).

- ARKANSAS - UNIT 2 6-21 Amendment No. MV

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'ADMIDISTRATIVE CONTROL

. CORE OPERATING LIMITS REPORT

9)  %:ESEC-Digital Simulation of a combustion E'ngineering Nuclear Steam ~

Supply System," December-1981l(Methodology for Specifications. _

3.1.1.1 and 3.1.1.2 for Shutdown Margin, 3.1.1.4.for MTC, 3.1.3.1 for Movable Control Assemblies'- CEA Position, 3.1.3.6 for'

' Regulating CIA Insertion. Limits,'3.1.3.7 for Part Length-CEA Insertion Limits, and 3.2.4.b for DNBR Margin).

10) Letter - O.D. Parr (NRC).to F.M.' Stern-(CE), dated June'13,-1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model).' NRC approval for 6.9.5.1.4, 6.9.5.1;5, and 6.9.5.1.8 methodologies.

- 11)' Letters O.D. Parr (NRC) to A.E. Scherer (CE), dated December 9, 1975 (NRC Staff Review of the Proposed combustion Engineering ECCS Evaluation Model changes). NRC approval for 6.9.5.1.6 methodology.

12) Letters 2CNA038403, dated March 20, 1984, J.R. Miller (NRC) to J.M. Griffin (AP&L),."CESEC Code Verification." NRC spproval for 6.9.5.1.9 methodology.

I 6.9.5.2 . The core operating' limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.5.3 The CORE OPERATING LIMITS REPORT, including any mid-cycle' revisions or supplements thereto, shall be provided upon issuance to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

ARKANSAS - UNIT 2 6-21a Amendment No. 4 M

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MARKUP OF CURRENT ANO-2 TECHNICAL SPECIFICATIONS (FORINFO ONLY) l l

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