ML041410523

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Arkansas, Unit 2, License Amendment Request to Support Cycle 18 Core Reload
ML041410523
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 05/12/2004
From: Forbes J
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
2can050405
Download: ML041410523 (40)


Text

Ak Entergy Operations, Inc.

1448 S.R. 333 E n tcergy 01 ZTel RusselMille, AR 72802 479-858-4888 Jeffrey S. Forbes Vice President Operations ArO 2CAN050405 May 12, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request To Support Cycle 18 Core Reload Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests an amendment for Arkansas Nuclear One, Unit 2 (ANO-2) to modify Technical Specification (TS) 6.9.5, Core Operating Limits Report. The proposed change will revise the analytical methods used to determine core operating limits.

In addition to the above changes, Entergy proposes to reflect the changes allowed by Technical Specification Task Force (TSTF) Traveler No. 363, "Revised Topical Report References in ITS 5.6.5, COLR." The TSTF permits the analytical methods listed in the TSs to be identified with the Topical Report number and title only.

Entergy also proposes to delete the Index from the TSs. The Index is not part of the technical content of the TSs and therefore does not need to continue to be reviewed and approved by the NRC as part of the license amendment process.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.

The proposed change includes a new commitment, which is listed in Attachment 3.

There is currently a proposed change to TS 6.9.5 under NRC review (initial letter dated June 30, 2003, "License Amendment Request Revision of Section 6.0, Administrative Controls'). The proposed change included the rearrangement of information in Section 6.0, which will result in a change to the number of TS section 6.9.5 to section 6.6.5.

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2CAN050405 Page 2 of 3 Entergy requests approval of the proposed amendment by February 28, 2005, in order to support the spring 2005 refueling outage. Once approved, the amendment shall be implemented within 60 days. Although this request is neither exigent nor emergency, your prompt review is requested.

If you have any questions or require additional information, please contact Dana Millar at 601-368-5445.

I declare under penalty of perjury that the foregoing is true and correct. Executed on May 12, 2004 Sinc qe I 7 JSF/dm Attachments:

1. Analysis of Proposed Technical Specification Change
2. Proposed Technical Specification Changes (mark-up)
3. List of Regulatory Commitments - Supplemental Information Describing the Use of the Westinghouse Nuclear Physics Code Package for Arkansas Nuclear One, Unit 2

2CAN050405 Page 3 of 3 cc: Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Drew Holland MS O-7D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205

Attachment I 2CAN050405 Analysis of Proposed Technical Specification Change to 2CAN050405 Page 1 of 11

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2). The proposed changes will modify the analytical methods referenced in Technical Specification (TS) 6.9.5.1, with a Westinghouse Nuclear Physics code package and a methodology that will support the use of ZIRLO fuel cladding.

In addition, Entergy proposes to reflect the changes allowed by Technical Specification Task Force (TSTF) Traveler No. 363. The TSTF permits the analytical methods listed in TS 6.9.5.1 to be identified with the Topical Report number and title only.

The deletion of the Index is also proposed. The Index does not include any technical information and therefore any changes to it should not require review or approval by the NRC.

Its removal is purely administrative.

Entergy desires approval of the proposed changes by February 28, 2005, in order to support the spring 2005 refueling outage.

2.0 PROPOSED CHANGE

TS 6.9.5.1, Core Operating Limits Report (COLR)

The proposed changes to TS 6.9.5.1 include the replacement of the referenced physics code package ROCS and DIT with a Westinghouse Nuclear Physics code package (PHOENIX-P/ANC); the addition of a Westinghouse Physics code (PARAGON); incorporation of TSTF, Traveler No. 363; and the addition of a topical report related to ZIRLO fuel cladding.

Westinghouse Nuclear Physics Code Package TS 6.9.5.1, Item 1 "The ROCS and DIT Computer Codes for Nuclear Design" will be deleted.

The ROCS and DIT codes will be replaced with the following Westinghouse Nuclear Physics code package:

  • WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System For Pressurized Water Reactor Cores"
  • WCAP-1 0965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code"
  • WCAP-1 0965-P-A Addendum 1, "ANC: A Westinghouse Advanced Nodal Computer Code: Enhancements to ANC Rod Power Recovery" A new item 9 will be added to TS 6.9.5.1 to include a reference to WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON." Westinghouse has not published the "-A" version as of the date of this letter; however, the NRC has approved the topical report for use. The topical report allows PARAGON to be used as a replacement for the PHOENIX-P lattice code. PARAGON may be used in the final reload analysis for the spring 2005 refueling outage.

to 2CAN050405 Page 2 of 11 TSTF Traveler No. 363, Revise Topical Report References in ITS 5.6.5, COLR The following proposed changes to TS 6.9.5.1 are in accordance with TSTF Traveler No. 363:

  • TS 6.9.5.1, Item 2 "CE Method for Control Element Assembly Ejection Analysis,"

CENPD-0190-A - the revision date, January 1976, will be deleted and relocated to the cycle specific COLR.

  • TS 6.9.5.1, Item 3 "Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A -

the revision date, May 1988, will be deleted and relocated to the cycle specific COLR.

  • TS 6.9.5.1, Items 4, 5, 6, 7, and 17 all refer to the "Calculative Methods for the CE Large Break LOCA Evaluation Model," CENPD-132-P. The revision date, supplement numbers, and revision numbers will be deleted as appropriate and relocated to the cycle specific COLR. This analytical method will only be listed once in the TSs as Item 4.
  • TS 6.9.5.1, Items 8, 9, and 10 all refer to the "Calculative Methods for the CE Small Break LOCA Evaluation Model," CENPD-1 37. The revision date, supplement numbers, and revision numbers will be deleted as appropriate and relocated to the cycle specific COLR. This analytical method will only be listed once in the TSs as Item 5.
  • TS 6.9.5.1, Item 11, "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System" - the revision date, December 1981, will be deleted and relocated to the cycle specific COLR. This analytical method will be listed in the TSs as Item 6.
  • TS 6.9.5.1, Item 12, 'Technical Manual for the CENTS Code," CENPD 282-P-A- the revision date will be deleted and relocated to the cycle specific COLR. This analytical method will be listed in the TSs as Item 7.
  • TS 6.9.5.1, Item 13, Letter: 0. D. Parr (NRC) to F. M. Stern (CE, dated June 13, 1975 -

will be deleted and relocated to the cycle specific COLR.

  • TS 6.9.5.1, Item 14, Letter: 0. D. Parr (NRC) to A. E. Scherer (CE), dated December 9, 1975 - will be deleted and relocated to the cycle specific COLR.
  • TS 6.9.5.1, Item 15, Letter: K. Kniel (NRC) to A. E. Scherer (CE), dated September 27, 1977 - will be deleted and relocated to the cycle specific COLR.
  • TS 6.9.5.1, Item 16, Letter: 2CNA038403, dated March 20, 1984, J. R. Miller (NRC) to J. M. Griffin (AP&L) - will be deleted and relocated to the cycle specific COLR.

ZIRLO Cladding A new reference (TS 6.9.5.1, item 8) will be added that will support the use of ZIRLO fuel cladding. Topical report CENPD-404-P-A, "Implementation of ZIRLO Material Cladding in CE Nuclear Power Fuel Assembly Designs" summarizes the ZIRLO material properties as they pertain to fuel rod cladding and provides an evaluation of these properties and the correlations that Westinghouse intends to use in design and licensing analysis activities. In addition, CENPD-404-P-A identifies the specific CENP topical reports that would be impacted by the implementation of ZIRLO cladding, and describes the substitutions that would be required as a result of the proposed ZIRLO implementation.

to 2CAN050405 Page 3 of 11 Index Entergy also proposes to delete the Index from the TSs. The Index is not part of the technical content of the TSs and therefore does not need to continue to be reviewed and approved by the NRC as part of the license amendment process. No further discussion is included in relationship to this administrative change.

Summary of Proposed Changes In summary, Entergy proposes a change to the analytical methodologies listed in ANO-2 TS 6.9.5.1 to allow the use of the Westinghouse Nuclear Physics code package, which includes NRC approved topical reports WCAP-11596-P-A, WCAP-1 0965-P-A, WCAP-1 0965-P-A, Addendum 1 and WCAP-16045-P-A; and to allow the use of ZIRLO fuel cladding by the addition of CENPD-404-P-A. Entergy also proposes the adoption of TSTF Traveler No. 363.

And finally, an administrative change is proposed to delete the TS Index.

3.0 BACKGROUND

TS 6.9.5.1, Core Operating Limits Report (COLR)

Westinghouse Nuclear Physics Code Package Nuclear designs for reloads and the evaluation of reload safety for ANO-2 have been performed using the ABB-CE reload methodology. To date this methodology has been executed using the tools which constitute the ABB-CE Nuclear Physics code package, DIT/ROCS as approved in CENPD-266-P-A. The basis for this change is a transition to the Westinghouse Nuclear Physics code package with the continued application of the ABB-CE reloads and safety analysis methodology. The change in the tools used to execute the reload methodology is based on the integration of technologies arising from the consolidation of the former ABB-CE nuclear entities with Westinghouse Electric Company LLC.

The Westinghouse Nuclear Physics code package is based on the ANC and PHOENIX codes, which have been reviewed and approved previously by the NRC, as described in the following topical reports:

  • WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores"
  • WCAP-1 0965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code"
  • WCAP-1 0965-P-A Addendum 1, "ANC: A Westinghouse Advanced Nodal Computer Code: Enhancements to ANC Rod Power Recovery" The Westinghouse Nuclear Physics code package has been used extensively for the design of reload cores and for evaluation of reload safety for a wide range of core sizes and fuel array sizes encompassing typical designs for Westinghouse, Combustion Engineering (CE),

and Framatome designs. The capabilities and functionality of the ANC/PHOENIX technology is well known by the NRC and the nuclear industry. Based on significant experience, including benchmarks on several CE type plants, the application of the Westinghouse Nuclear Physics code package is expected to provide predictions of key core parameters that are to 2CAN050405 Page 4 of 11 essentially the same as those obtained with the current DIT/ROCS methodology. Margins in nuclear design, based solely on the transition in nuclear physics methods, are expected to remain essentially unchanged.

The following WCAP will also be added as a new item to the list of COLR references.

WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON" This WCAP, as approved by the NRC, allows the PARAGON code to be used as a replacement for the PHOENIX-P lattice code. PARAGON may be used in the final reload analysis for the spring 2005 refueling outage.

TSTF Traveler No. 363, Revise Topical Report References in ITS 5.6.5, COLR TSTF Traveler No. 363 was written to allow the requirements in TS 6.9.5, COLR, to identify the topical report(s) by only number and title. The TSTF allows the deletion of the date, revision numbers, and any supplements from the TSs. In accordance with the TSTF this information can be relocated to the cycle specific COLR. This method of referencing topical reports allows the licensee to use current topical reports to support limits in the COLR without having to submit an amendment to the facility operating license each time the topical report is revised. With the approval of the proposed change, unnecessary expenditure of NRC and licensee resources will be eliminated and the burden of TS submittal and approval needed to license reload fuel will ease. TSTF Traveler No. 363 was approved by the NRC on April 13, 2000.

ZIRLO Cladding Ina continuing effort to improve fuel performance, ANO-2 plans to implement ZIRLO cladding material for new fuel assemblies beginning in 2005. The use of ZIRLO clad fuel rods will substantially reduce exterior corrosion and particularly the spalling experienced by some current Zircaloy-4 clad fuel rods as they approach higher burnup levels and duty cycles. The proposed TS change to TS 6.9.5.1 provides a methodology reference for the use of ZIRLO clad fuel rods in the ANO-2 reactor core.

The topical report describes the implementation of ZIRLO fuel rod cladding material properties and correlations in design and safety analysis methodologies for CE design reactors and fuel. Westinghouse has performed extensive evaluations which have concluded that the application of ZIRLO in existing CE fuel designs does not result in any undesirable changes in predicted fuel performance or safety analysis results. While modification to CE computer codes are required to implement ZIRLO material properties, no modifications are required to the NRC accepted ZIRLO properties or analysis methodologies for CE design nuclear steam supply systems and fuel designs, design performance criteria, or regulatory acceptance criteria. ANO-2 is a CE designed reactor and is supplied with CE designed nuclear fuel.

to 2CAN050405 Page 5 of 11 The ZIRLO topical report requires the use of specific versions of the Westinghouse Emergency Core Cooling System (ECCS) performance evaluation models for CE designed reactors (i.e., CENPD-1 32, "Calculative Methods for the CE Large Break LOCA Evaluation Model" and CENPD-1 37, "Calculative Methods for the CE Small Break LOCA Evaluation Model"). These are currently included in the ANO-2 TS as analytical methods used to determine the core operating limits.

4.0 TECHNICAL ANALYSIS

TS 6.9.5.1, Core Operating Limits Report (COLR)

Westinghouse Nuclear Physics Code Package To support the application of the Westinghouse Nuclear Physics code package for ANO-2, plant specific comparisons of key physics parameters for two cycles of past plant operation are provided in Enclosure 1. The comparison provided is between actual plant operating data from Cycles 15 and 16 and design calculations using the Westinghouse Physics Code package PHOENIX-P/ANC. In future ANO-2 core reload evaluations Westinghouse may use PARAGON in place of PHOENIX-P as allowed by one of the conditions in the NRC safety evaluation for WCAP-16045-P which states: 'The PARAGON code can be used as a replacement for the PHOENIX-P lattice code, wherever the PHOENIX-P code is used in NRC-approved methodologies."

Implementation of the Westinghouse Nuclear Physics code package requires no functional changes in the current reload methods. There are no changes in the safety analyses or safety methods. Changes are limited to those necessary to support effective and accurate electronic transfer of data from the Westinghouse Nuclear Physics code package to the downstream interface codes which are components of the current reload methodology.

The first reload cycle for application of the Westinghouse Nuclear Physics code package for ANO-2 is Cycle 18 (Spring 2005).

The NRC Safety Evaluation (SE) for the Westinghouse topical report WCAP-16045-P, Revision 0, "Qualification of the Two-Dimensional Transport Code PARAGON," was approved by the NRC on March 18, 2004. The SE contained two conditions and limitations which are addressed below.

Condition 1 The PARAGON code can be used as a replacement for the PHOENIX-P lattice code, wherever the PHOENIX-P code is used in the NRC-approved methodologies.

Response l This letter includes a request to use the PHOENIX-P code. As allowed by this Condition, Westinghouse may use PARAGON as a replacement code for PHOENIX-P code when evaluating the ANO-2 core reload for the spring 2005 refueling outage. Therefore, this letter also includes a request to include WCAP-1 6045-P.

to 2CAN050405 Page 6 of 11 Condition 2 The data base is insufficient to enable the staff to reach a conclusion regarding PARAGON's ability to predict depletion characteristics for a MOX fueled core at this time.

Response 2 The ANO-2 core is not a MOX fueled core and there are no immediate plans to use MOX fuel; therefore, this condition does not apply.

TSTF Traveler No. 363, Revise Topical Report References in ITS 5.6.5, COLR The proposed method of referencing topical reports will allow ANO-2 to use NRC approved topical reports to support limits in the COLR without having to submit an amendment to the TSs each time the topical report is revised. The particular approved topical reports used to determine the core limits for the particular cycle will be included in the COLR. This will eliminate unnecessary expenditure of NRC and Entergy resources, and will ease the burden of TS submittal and approval needed to license reload fuel.

ZIRLO Fuel Cladding The topical report CENPD-404-P-A was submitted to the NRC for review in January 2001. It describes the implementation of ZIRLO fuel rod cladding material properties and correlations in Westinghouse design and safety analysis methodologies for CE designed reactors and fuel. The topical report was generically accepted by the NRC for application to CE designed reactors and fuel in September 2001 subject to five conditions. ANO-2 was licensed to use ZIRLO cladding (letter from the NRC to Entergy dated May 19, 1999, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 205") before the CE topical report supporting the use of ZIRLO cladding was submitted to the NRC for approval.

To date, ZIRLO cladding has not been used as a cladding material in the ANO-2 core and the five conditions currently stated in the safety evaluation for use of the CE topical for ZIRLO cladding have not been previously addressed. ANO-2's responses to the five conditions are as follows:

Condition 1:

The corrosion limit as predicted by the best-estimate model will remain below 100 microns for all locations of the fuel.

Response 1:

During the design of each fuel cycle, the corrosion thickness is calculated using the best estimate models and methods. The calculated corrosion thickness is verified to be no greater than the maximum allowable corrosion limit of 100 microns.

to 2CAN050405 Page 7 of 11 Condition 2:

All the conditions listed in the safety evaluations for all the CENPD methodologies used for ZIRLO fuel analysis will continue to be met, except that the use of ZIRLO cladding in addition to Zircaloy-4 cladding is now approved.

Response 2:

ANO-2 will continue to abide by the conditions listed in the safety evaluations for all CENPD methodologies used in the analysis of ZIRLO fuel. This will be accomplished through the reload process that is employed.

Condition 3:

All CENP methodologies will be used only within the range for which ZIRLO data was acceptable and for which the verifications discussed in CENPD-404-P-A and responses to requests for additional information were performed.

Response 3:

Use of CENP methodologies within the accepted data ranges for ZIRLO is verified during the design and safety analysis of each fuel cycle.

Condition 4:

Until data is available demonstrating the performance of ZIRLO cladding in CENP designed plants, the fuel duty will be limited for each CENP designed plant with some provision for adequate margin to account for variations in core design (e.g., cycle length, plant operating conditions, etc.). Details of this condition will be addressed on a plant specific basis during the approval to use ZIRLO in a specific plant.

Response 4:

The modified Fuel Duty Index (mFDI) will initially be limited until data is available demonstrating the performance of ZIRLO cladding at ANO-2.

The maximum mFDI calculated based on actual 16 x 16 Combustion Engineering-designed fuel is approximately 590. To provide adequate margin to account for variations in core design, 110% of the approximate 590 value (652) is used for the majority of the ZIRLO clad fuel pins. For a fraction of ZIRLO clad fuel pins in a limited number of assemblies (no more than eight fuel assemblies), the mFDI limit is 120% of the approximate 590 value (712). The mFDI values of 652 and 712, with the aforementioned limitations will be used as upper design limits for the ANO-2 fuel.

If the mFDI and measured oxide thickness correlate as expected or is conservative relative to predictions, ANO-2 will no longer restrict the mFDI except as required to meet the 100 micron oxide limit.

to 2CAN050405 Page 8 of 11 Condition 5:

The burnup limit for this approval is 60 GWd/MTU.

Response 5:

Per ANO-2 license condition 2.C(9) Entergy Operations is authorized to operate the facility with an individual rod average fuel burnup (burnup averaged over the length of a fuel rod) not to exceed 60 megawatt-days/kilogram of uranium.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. Entergy Operations, Inc. (Entergy) has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the Technical Specifications (TSs), and do not affect conformance with any General Design Criterion (GDC) differently than described in the Safety Analysis Report (SAR).

5.2 No Significant Hazards Consideration The proposed change will modify TS 6.9.5.1, Core Operating Limits Report (COLR) to support core reload activities for Arkansas Nuclear One, Unit 2 (ANO-2).

The proposed change will delete one of the methodologies ("The ROCS and DIT Computer Codes for Nuclear Design," CENPD-266-P-A) listed in the administrative controls section of the ANO-2 TSs to allow the use of the Westinghouse Nuclear Physics code package, which includes NRC approved topical reports WCAP-1 1596-P-A ("Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores"), WCAP-10965-P-A

("ANC: A Westinghouse Advanced Nodal Computer Code"), and WCAP-1 0965-P-A Addendum 1 ("ANC: A Westinghouse Advanced Nodal Computer Code: Enhancements to ANC Rod Power Recovery").

WCAP-1 6045-P-A ("Qualification of the Two-Dimensional Transport Code PARAGON") will also be added to the list of methodologies included in the TSs. This allows the PARAGON code to be used as a replacement for the PHOENIX-P code.

Entergy also proposes to incorporate a reference to CENPD-404-P-A, "Implementation of ZIRLO Material Cladding in CE Nuclear Power Fuel Assembly Designs."

In addition, Entergy proposes the adoption of Technical Specification Task Force (TSTF)

Traveler No. 363, "Revise Topical Report References in ITS 5.6.5, COLR." This will result in the deletion of topical report dates, revision numbers, and supplement numbers and their subsequent relocation to the cycle specific core operating limits report (COLR). The cycle specific COLR will contain the complete identification of each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

to 2CAN050405 Page 9 of 11 And finally, an administrative change is proposed to delete the Index from the TSs.

Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92,

'Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

TS 6.9.5.1, Core Operating Limits Report (COLR)

The proposed amendment, in part, identifies a change in the nuclear physics codes used to confirm the values of selected cycle-specific reactor physics parameter limits and includes minor editorial changes which do not alter the intent of stated requirements. The proposed change also allows the use of methods required for the implementation of ZIRLO clad fuel rods. Inasmuch as the proposed change includes codes that have been previously approved by the NRC for CE cores, the amendment is administrative in nature and has no impact on any plant configuration or system performance relied upon to mitigate the consequences of an accident. Parameter limits specified in the COLR for this amendment are not changed from the values presently required by TSs. Future changes to the calculated values of such limits may only be made using NRC approved methodologies, must be consistent with all applicable safety analysis limits, and are controlled by the 10 CFR 50.59 process.

Assumptions used for accident initiators and/or safety analysis acceptance criteria are not altered by this change.

The proposed change also implements NRC approved TSTF Traveler No. 363. This is an administrative change that will allow specific details, such as the revision number, revision date, and supplement number of topical reports that are referenced in the TSs, to be deleted and relocated in the cycle specific COLR. This proposed change does not result in any changes to the assumptions used to evaluated accident initiators and/or safety analysis acceptance criteria.

Index The proposed deletion of the Index is purely administrative and does not impact the accident analysis.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

to 2CAN050405 Page 10 of 11

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

TS 6.9.5.1, Core Operating Limits Report (COLR)

The proposed change, in part, identifies a change in the nuclear physics codes used to confirm the values of selected cycle-specific reactor physics parameter limits. The proposed change also allows the use of methods required for the implementation of ZIRLO clad fuel rods. Neither of these changes results in a change to the physical plant or to the modes of operation defined in the facility license.

The proposed change also implements TSTF Traveler No. 363. The proposed change does not result in changes to the physical plant or to the modes of operation defined in the facility license nor does it involve the addition of new equipment or the modification of existing equipment.

Index The proposed deletion of the Index is purely administrative has no affect on existing equipment.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

TS 6.9.5.1, Core Operating Limits Report (COLR)

The proposed changes to change the nuclear physics code package and to add a topical report to support the use of ZIRLO do not amend the cycle specific parameter limits located in the COLR from the values presently required by the TS. The individual specifications continue to require operation of the plant within the bounds of the limits specified in COLR. Benchmarking has shown that uncertainties for the Westinghouse Physics code system yields are essentially the same or less than those obtained for the current ROCS/DIT methodology. Future changes to the values of these limits by the licensee may only be developed using NRC approved methodologies, must remain consistent with all applicable plant safety analysis limits addressed in the Safety Analysis Report, and are further controlled by the 10 CFR 50.59 process. The relocation of the supplement numbers, revision numbers, and approval dates of the analytical methods listed in the COLR does not affect the margin of safety. The analysis will continue to be performed using NRC approved methodology. Safety analysis acceptance criteria are not being altered by this amendment.

to 2CAN050405 Page 11 of 11 Index The proposed deletion of the Index, which is an administrative document, does not impact any TS values or safety limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 'no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE Similar changes for the use of ZIRLO have been reviewed and approved by the NRC for Calvert Cliffs Nuclear Power Plant and Palo Verde Nuclear Generating Station.

Attachment 2 2CAN050405 Proposed Technical Specification Changes (mark-up)

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IND-EX SA FE--Y-LI M ITS -AN D- M I4N G -SAF-EqY-SY-S--E  ; FE r--- N GS SE-GT4ON PAGE 2 .4--SAFE-T-Y-4AMITS Reactor-Gore ....................................................................

Reactor-Coolant--System-P4ressure . ...................................................................... 2 2 2-A 4AM IT-IN G-.SAFE-T--SY-S-TE-M-S E-TTNGS Reactor-T-rip-Setpoints ................................................................. 23 Deleted ................................................................. 2'12A BASES SECTIO0N PAGE 2-A-SAFE--TY-4MIT-S Reactor-G ore ................................................................. B21 ReaotorCoolant-System-Pressure ................................................................. B22 2-2-1,MITNG-SAFETY SY-STE-M-SE-TT4NGS Reactor-Trip-Setpoints ............................................................. B Deleted................................................................................................................. 6-2-B A - 11/

A4, .

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1ii _ rr A:mnement

_ Al _ K eNo.

I _ n no A 2.60.U I i-

INDE(X LIMITING-ONDT4QNS-FOR-OPERAT4ON-AND-SURVE-ILL-ANGE-REQUIREMENTS SE-GT40N PAGE-340-APGABILITY..........340 311.4-REACTIVITY CONTRL SYSTEMS 314.1.1 CBRAT40N-GONT-ROL Shutdown-Margin-T-, -- 9 20GF ............................................................... 33/44-4 Shutdown-Margin-Tavg- -2 OF ............................................................... 3/44-3 Boron-Dilution ............................................................... 3144-4 Moderator-T-emperature-Goefficient ........................................................... 3/144-Minimum-Temperaturefor-riticality .......................................................... 3/44-6 3/4A-3-GONT-ROL-ELEM ENT-ASSEMBLIE-S CE-A-Rosition .............................................................................................. 3114 1 17 P-ositionIndirator-Ghannels--Operating .................................................... 3/4-4-20 P-osition-lndirator-Ghannels-Shutdown.................................................... 34144-22 GEA Drop Time ............................................................... 3/1 1 23 Shutdown-GE-A-Insertion-1imit ................................................................ 3/1 1 24 Regulatinand-GroupP-CE-A4nrsertien-Limits ........................................... 3/L I A QV AKIO AC I IKIITr 1 1% I Amendment No. 37,60,157,169,229

INDEX LUMITING-GONDITIONS-F-OR-OP-E-RATN-ANDURVEIL-L-ANGE-RE-QUIREMENTS SE-GT40N PAGE 3/4.2 PO WEAR-DISTR4BUT4QN4MIWTS 3 /42 . 2-1LIN E -A ................................................................ 31 2 1 3/42 ADIAL- P-EA...............................................................

N-A 34-2-2 3 /14. 2. 3AZIM A L POT ............................................................... . 314 3 3/4.24 DNBR-MARGIN ............................................................... 3/4 314.2. 5-RGS-LGWIV-RATE .......... IA) 3/4F4Ire............................................................... 3/12... 7 3/42--REAGCTOR-GQQL-ANT-GOL-D-L-E-C--T-MP-E-RATURE ................................ 3/4-2 3/4.2.7- AXIALSHAP-E--NDEX ............................................................... 3/4-2-9 31448-P-RESSURIZER-P-RESSURE ............................................................... 3/4-240 3/4-3-4NSTRUME-NTAT4ON 3/14-3.-REAGTOR-PROT-E-GTIVE-SYST-E-MINST-RUMENTATION ...................... 34144 3M/ 2-E-NGINE-E-RE-D-SAFETY-F-E-AT-URE-AGTUATIQN-S-YSTEM INST-RUME-NTATIN. .. 3430 3/4-.3-3-MONITORING4NSTRUME-NTAr40N Radiation-Monitoring4nstrumentation ....................................................... 3/4-3-24 Remote-Shutdown-Instrumentation ....................................................... 34-3-36 Post-Acoident4nstrumentation ....................................................... 34-3-39 ARKANSAS UNIT 2 V Ameadment-No. 21,60,157,163,191, I 12o

INMDX L4MIT4NGONDIUNS-FOR-OPERAT4ON-AND-SURVEJ-LLANGE-REQUIREME-NT-S SE-CTION PAGE 3/14. RE-AGTGR-GOLANT-SYST-E-M 3/444 RE-AGT-OR COOLANT LOOPS AND GOOLANT-GIRGULATIN .............. 3/1 1 1 3/4-.42-SAF-E-TY-VAL-VE-S--SHIUTDWIN .............................................................. 3/4-4-3 3/4.1.-3SAF-E-TVALVE- ERATING............................................................... 3/4 4 4 31.1.4-P-RE-SSUR4ZER......................................................................................... 3/4-4-5 3/4A4-5-STEAM-A E-NE-RATORS ............................................................... 34-4-6 3/4.4-R EAC-TOR-GOOLANT-SYST-E-M-L-E-AKAGE Leakage-Detertion-Systems ............................................................... 314-4-13 Reaotor-Goolant-System-L-eakage ............................................................. 3/444 3/44-7-GHEMISTRY ............................................................... 314-4145 3/448-SPEGIFG-ACTIVITY ............................................................... 3/144 18 3/-.49 -.-PR E-SS UR ET-M PER AT-U RE-LIMIT-S ReactorF-oolant-System ............................................................... 3/4-4-22 Prcssurizer ............................................................... 314-4-25 3/4440-STRUGT-U RAL4NT-E-GRITY-ASME&C e-Gass 1,2and3Gmpoents ............................................... 3l41 26 3/4.4.1-RE-AGTOR-GOOLANT SYSTEM ENTS .................................................. 3/1 4 27 3/4.4.12 LOW TEMPE-R RE OVERPRESSURE-PROT-E-ON (LTGP) SYST-EM 3!4 4 28 3/4E-MERGE- NGY-GGRE-GQLNG YSTE-MS4EGGS-3/47 54-SAF-E-TY-INJE-GT4ON-TANS ............................. 3 ARKANSAS UNIT 2 V! Amendmento. N 9,60,63,191,199

INDEX LMIg4NG-GONDITIONS-FOR-OPERATION-AND-SURVELL-ANGE-REQUIREME-NT-S SECGTION tPAGE 3M4-52-EGGS-SUBSYS-T-EMS -Tavg-Ž 30002F ........................................................ 3/4-5-3 3/1.5.3 ECCS SUBSYSTEMS Tawg -300F ........................................................ 3/4 5 6 314.5A-RE-PUE-MNG WAT-E-R-TANK ............................................................... 34-3/--GQNTAINME-NT-SY-STEMS 3I4-4-P.-MARY-GONTAINME-NT Containment-Integrity ............................................................... 34-6I Gontainment-Leakagc ............................................................... 34-6-2 Gontainment-Air-Looks ............................................................... 34 6-4 internalP-ressure,-Air-Temperature-and-Relative-Humidity ........................ 3/4-6 Gontainment-Struotural Integrity ............................................................... 3--8 Gontainment-Ventilation-Syste ....................................................... 3/4-6 a 314.6;2-DEP-R ESSURIZ-AT4ONFGOOLNG-AND-PH-CONTROL-SYSTE-MS GORta.inmcnt Spray System ....................................................................... 314 6 10 Tfisodium-Phosphate*TSP) ....................................................................... -3/4-6-2 Containment-Gooling-Syste .................................................................... -3/4 6-14 3M.-6.3-GONTFAINMENT4SOL-ATION-VALES ..................................................... 646 364 COMBUST4BLE-GAS CONTROL Hydrogen-Analyzers ................................................................................... -314-6-1-8 E4eGtIiG Hydrogen Rccombiners W.......................................................... 3/4- 6 19 AAnAf A rofsA AKKJK!

^ Xrg A rs U Ift X Ss

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... 7

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LUMrnNG-GON DITIONSs-FOR-OPERA-TIN-AND-SURVE-ILL-ANGE-REQUIREMEN-TS SE-GTION PAGE 3/14-7PANT SYSTE-MS 3/4.7-.1 TURBINE-GY-GLE Safety-Valves... 34-E-nergenGy-Feedwater-Sysem .31 5 Gondensate-Storage-Tank .31477 A3ivity..3/-7-8 Main-Steam-isolation-Valves .3/4-7-40 314.7. ST-EAM-GE-NE-RATOR-P-RE-SSRE4T-E-MP-ERAT-URE-L4MITTATQN 3/1714 3/4.7.3-S E- GE-WAT-E-R-SYS E.... 311-745 314 74- E-ME-RGE-NGY-GQ L4NG-P-QND ............................................................... 31474X6 3/4-7-5-F-lOD-P-RT-E-GTION ............................................................... 3146a 3/4.-7-.ON-RTL-ROQM-E-ME-RGE-NCVEWVT4LAT4ON-AND TIR ARGNI4NING-G N D MTO I NSYSTEY -T- . .............................................................. 311 314 7 17-3/444-SHOGK-SUP-PRE-SSORS-(SNUBBE-S) ................................................... 314-7-22 3/44749-SE-AL-E-D-SOURGE CONTAMINAT1N ..................................................... 3/4-7a 314. 7.10-F.RE-SUPP-RE-SSION-SY-ST-E-MS 7

F-ire-Suppression-WateF-Syste ................................................................. 31 7 29 Spray-and/or-Sprinklef-Systems ............................................................... 34-7-33 Fire-Hose-Stations . .............................................................. 3/4-7-35 3/1.7.11 FIRE BARRIERS ............................................................... 31 7 37 3/14-712-SP-ENT-F-UE-L-P-OOL-TRUGTURAL4NTE-GRI-Y .................................... 3M-38 3 ELECTRICAL Po 31.8.1 A.G. SURGE-S Operating ........................................................ 311 8 1 Shutdown ..........................-- -- .................... 3/4-8-5 A b1/ A K1O A IKIIT . I...

A n

vil o l AKNANEAik UNI I e meRntNo. 3..t.A

INDEX L4MmTNG-GONDIT]ONSFOR-OPERATON-AND-SURVE-L-L-ANGE-REQUIREMENTS SE-GTION PAGE-314&2-NSNTE-P-OWEE-R lSTPdB tMQN SYSTEMS A G D i6tribution- Operating ............................................................... 34-M6 A-G-Distribution-Zhutdow ............................................................... 34-DG Distfibution-Opefating ................................................................ 3.4-8 D.C. Distribution-Shutdown ...............................................................34-840 Gontainment-P-enetration-GondurtorF-vercurrent-P-roteotive-DeviGes ....... 3/4-8 14 3/4-9-REF-UELNG-GPE-RATIONS 3/1494-BORON-GONGENTRNTON ............................................................... 34-94 3/42-4 NSTRUME-NTAT N ............................................................... 39 3/4-9,3-DEGA-YT4ME-AN D-SP-E-NT-FUE-t-STORAGE ............................................. 3/4-9-3 314.9.4 CONTAINME-NT-BUILDING-PE-NE-TRATNS ........................................... 31/4-9-4 3144.,5-GOMMUNIGAT4NS ............................................................... 3/4-9-6 3/4..6-REFUJELANG-MACHIN EOP-ERABILT4Y .3.................................................... 3/494 3/1.9.7 CRANE TRAVEL SPENT FUEL POOL BUILDING ................................ 3/19 8 3/4-8SHUT-DOWN-GGOO NGAND GOLANTGIRGULATION .......... .3/4 3..............

9 3/14.9.9 WAT ERLEVE-L-REACTORVESSEL ..................................................... .3/4 9 10 314.9.10-SPENT-F-U EL POL WATE-R E L........................................................ 3/49-9-44 314.9.11 FUE-L-N4"IUNG-A .EA VENTILATION SYSTEM .............. ..................... 3/1 9412 31492-FUE-L-STORAGE ............................................................... 3/4-944 3/440-SP-E-G-ALt-T-ST-E-XGE-PTI4NS 3144A04-SHUT-DOWN-MARGIN ............................................................... 3/4-104 3/440-2-GROUP-HE4GHT-FINSE-RT4N-AND4POWE-R-DISTRIBUT]ON -LIMTS... 3/440-2 3/1.44--EACTOR C OLLANT LOOPS ............................................................... 3/440-3 ARKANSAS UNIT 2 IX A ndment-Noe.29,60

INDEX LUMrnNG-GONDIUTONS-FOR-OPE-RATUON-AND-SURVEILL-ANGE-REQUIREME-NTS SE-GT4ON PAGE-314.104G E GE-A-MI&AUGNMENT ............................................................... 31 10 1 344 O.5MINIMNUM-TE-MP-EF-RATRE--FO GR-TIGALITY ....................................... 311 10 5 3/4.4-RADIOACTIVE-F-F-F E-NT-S 3/4.1 1.1 LI - QUIOU...............................................................

PATA N S. 3111 3144h2-GAS-STORAGE-TAN ............................................................... 3/44-2 3/4.4.3-E-XP-L-OSIV-E-GAS-MIXT-URE ............................................................... 3/44-3 ARKANSAS UNI' S

. ___ _ . -N. 0,93

BASES SE-GT4ON PAGE 3441-:AP-P-LICABILIT-Y............................................... B 3,1 001 3/4-4-E-AGTIVITY-GNTROL 0L-SYqEMS 3144-. BOURATON-CONTROL.............................................................................. .B-3/4 1 1 3I44-2 BORAT4ON-SYST-ES ............................................................... B 31 1 2 3I443-MOVABLE-GONTROL-ASSE-MBL4ES ....................................................... - B31441-3 3/.-2-P-OWE-R-DISTRIBUTIONI4-MITs 3/4,24-WNEAR-HE-A-T-RATE ............................................................... B-3/4-24 3/4-2-2-RADIAL-P-E-AKING-F-ACTOR ............................................................... B-342-2 3/.2.3 ZAMUTHAL-P-WE-R-T4LT......................................................................... B 314-2-2 314-2-4-DNBR-MARGIN ............................................................... B-3/4-2-3 31472-.5-RGS-F-Lz W-RAT-E ............................................................... B-3424 3/466-RE-ACTOR-C OL-AN-LANL-D LEGT-E-MPERATURE ............. ................... 2-4 3/4M2-.7-AXIAL-SHAPE-INDEX ............................................................... B-3/4-2-4 314.2.8 PRESSURIZER PRESSURE..................................................................... B 314 2 4 314-3-4NSTRUME-NTAT4ON 311.3.1 PROTECTIVE-4NSTRUMENTATION B. .3.................................

1 3/437 2-E-NGNEERE-D-SAF-ETY-F-E-ATU E-NST-RUME-NTATION .......... ............. B 4 314.3.3 MONITR4NG-INST-WMENTATION ........................................................ - B 3/4-3 2 ARKANSAS UNIT 2 Xi Amcndment No. 24.33.60.191

BASES SECTION P-AGE 3REACTORGOOLANT-SYSTEM 3114.1-REAGTORCOLANT LOOPS-AND COOLANT-GIRGULATIO ............. B 314 4-4 3/14.4.2and-3/4.4.3-SAF-ETY-VA ES...................................................................... 3B/4-344 3/1.4-.4 -- P-RE-SS-URI ER BB.........................................

3/1 1 2 3 4-ST- AM- G E-NE-RAT RS ............................................................... .3/4. B 1 4 3I4-46-RE-AGTOR-GOL-ANT-SY-STE-M-LE-AKOAGE ........................................ B 3/ 3 3/4.4-7-GHEMISTRY ............................................................... B-34-4-4 3/448-SP-E-GF4G-A TIVITY ......................... ...................................... B-3/ 4 4 3/4.49-P-RE-SSURE-ITE-MP-ERATURE-L4MITS ...................................................... B-314-4-5 B--.

314A-10-STRUGT-URAL-INT-E-GRITY ............................................................... B-14-444 3/444-RE-AGTORCOOLANTSY-ST EMVENTS .................................................. - -314 4 11 3.41.12-LOW-TE-MP-ERATURE--OVE-RP-RE-SSURE-P-ROTE-G:TON-SY-STE-N-.3/41 419 3/4-5-EME-RGE-NGY-ORE-GOOL-ING-SYTE-MS4EGGS) 3!1.5.1 SAFETYINJEGTION TANKS.................................................................... B 3/ 5 1 3/4 and3/453-EGS-SUB TFM .............................................................. 4-4 345RF UEING WATVE4R-TAN (RWT) .......................................................... B 3452 3/4-6-CONT AINME-NT-SY-STE-MS 3/4.6.-PRIMARY CONTAINMENT ............................................................. B 31 6 1 3/4-62 DE-PRE--SSURlZ-AT4CN,-GOCL4NGQAND-pH-ONTROL-SYST-EMS.. B3/4-6-3 3/4-6.3-GONTAINME-NT-ISOLATON-VALVES .............. ..................... B-3/4-6 314-.64GOMBUST4BLE-GAS-GONTROL ................................... B-3/4-6" ARKANSAS UNIT 2 xi! ArecndfnentNe4 . 29,60,63,1 94,1 9

INDEX BASES SE-GTION PAGE 3/14-7-PLANT SYSTEM 3/14.7 .1 T URBIN E C Y CLE.1...................................................................................... B 3/1 7 1 3I4.7.2-ST-E-AM-GE-NERATOR-P-RE-SURE-IT-E-MP-E-AT-URE--4MITAT-ION .B...1 7 4 3/4-7-3--SE-RV4E WATE-R-SYST-EM ............................................................... B 3/4 7 4 3/4.7.4 E-ME-RGE-N CY-CO QLNG -P-OND ............................................................... B 47-4 3/4-7-5-F-LOOD-PR ECTION .............................. ................................. B-3/14 7 4 3/4-6-GONT-ROL-ROQM-EMERGENCY-VE-NTILATON-AND AIR-GONDITININ T ......... .. B-3/4-7-5 3/1.7.8 SHOCK SUPPRESSORS (SNUBBERS) ................................................... B 3/14 7 6 3.-7-.9-SE-AL-E-D-SOURGE--GONTAMINAT4QJ ..................................................... B-34-7-7 347 FIRE-SUPPRE-SSIO YSTEMS .............................................................. 3/147 3/14-.74-P-ENE-TRATION-F4RE-BARR4ERS ............................................................. 4-7-7 314.7.12 SPENT-FUE-6POOL STRUCTURAL INTEGRITY B....................................

B3/14 7 3/48-E-L-E-GTRIGAL-P-OWE-R-SYST-F-MS ............................................................... 4 3/44.-REFUELINGGP-ERAT4ONS 3/4,91-BORON-CONGENTRATION ............................................................... B-3/4-9-4 3/"-.92-NSTRUMENTATON ............................................................... B-3/94 3/41.9.3DECAY TIME ............................................................... B-3494 3/4-9.4 CQNTAINMENTP-ENE-TRAT40NS ............................................................ B-3/4-94 ARKANSAS UNIT 2 Xil Amendment No. 60,62,206

INDEX BASES SE-G-T4ON -PAGE

'51A Q C 'RNRPAh IlMlr('ATritKIC

, . .7f. v - - g I a- . .................................................................................

2t A n% a- r-I-rI ICI PK I^ KA A11I~ktC -I MlPA MI I TV D O3A fU 2' vW'atf tNUfif4+Yf_ f............=........................................................

I ...................... E ao i 314-.97-GRANE-TRAVE-L-SP-E-NT-F-JEUSTORAGE-BUIL-DING ......................... B 3/192 3/4.9.8 COO LANT-GERCU LATI N ........................................................ B 3/1 9 2 34W9and-4-940WA -TE-R-L-E-L--RE-AGCRVE-SSE-L-AND

- S TORAGE-POL-WATEL-EEL ................................................... 3/4-9-3 3/4.944-F-UE-HANDtNG-AREA-VE-NTLAT-ICON)SYSTYEM ................................... B 3/4 9 3 31440-SP-EGIAL-TE-ST-EXGEP-TIONS 314.10.1 SHUTDOWN MARGIN B 314 10 1 3/4.-4A -GRO UP-4E-GHT-1NSE-RTGN-AND POWE-R-DISTRIBUT40N-4UM[T,-S-, -- 3/4404 31440-3--REACT-OR COOLANT LOOPS ............................................................... 4 0 3/4404-GENT-R-GE-A-MISALIGNMENT ............................................................... B 14-404 3/440-6-MINIMUM4TM ATRE F=TRFOR CRITICALITY ........................................ B3/440-14 3/1444-RADIOAGTWVE-E-F-FLUENTS 34 -UQUID-HOLDUP-TANKS ............................................................... B-344-4 34.14A2-GAS-STORAGE-TANKS ................................. .............................. B-3444-11 3/4443-E-XPL-OSIVE-GAMIXT-URE ............................................................... B-3411 1 ARKANSAS UNIT 2 xIV Amendment-No-60&-,3

DE-SIGN-FEATURES SE-GTION PAGE-5.-1Site oGration................................................................................................... 1 5.2- Rea tor-Gore..................................................................................................

5-.24-F-uel-Assemblies ................................................................

6.2.2-GontroI-E4ement-Assemblic................................................................ 1 53-Fuet-Storapc................................................................ 2 5.a.-Spent-F uel -Storage-RaG k-Criticality ..................................................... 52 5&3.2-New-Fuel-Storage-RaGk-Gritioatity ........................................................

67 3..3-Drainage ............................................................... 52 5.3.1-GaoaGitv ............................................................... 52 and A i lo A l lz as_ A Amendmemnto A ant, AKKANSAS UNI i XV

INDEX ADMfN1S-TRATIVE-GONT-ROL-S SE-GT4QNI P-AGE 6.-RFESP-ONSIBILITY .............................................. 6 1 6-.2-OGRGANIZAT40N offsite .................................................. 61 Far~ifity-Staff ............................................. 6 1 6T-3--UNIT-S-TAF-F-QUALAF-IGAT4ONS ................................... 65r 64-TRAINING ................... 65................................

6-5-DE-LE-T-ED A 1nlO ^ K1O A o I IKIIT n Ad HI - _ ..K _ - - -r wp4i i 26 Avu AMeRGMeRl NE). la.

ADMIN ISTRAkT4VE-GONTROL-S SEGTION PAGE 6&-RE-PORTABIL-E-VE-NT-AGTN ............................. ,,,... 6 42 6-.7-SA F-E-TY-UM PF-V4 OL-AT4 ON .6...................13.................

68-PRO GE-DURES-AND-PROGRAMS .................................... 643 6 RE-PORT4NG-RE-QUIRE-ME-NT-S 6".AROUT4NE-RE-PR .S.

..................................................................................................................... 6. ..............................

614 a

6-92 SP-E-GIAL-RE-PR S .. 6 46 6.9.3 RADIOACTIVE EFFLUENT RELEASE REPORT.6 18 6&94ANNUAL-RADJOL-OGIGAL-ENV4RONME-NTAL-OPERATING-REP-ORT .6-20 6-9E5--RE-OPERAT4NG-L4MFS-REP-ORT. 6-24 640-RE-GORD-RE-T-NTON ...................... 622 6.-1--RADIATION-PROT-E-FT4ON -P-ROGRAM .6-23 642-Deleted. 6-23 6-43HIGH-RADIAT4N-ARE.A......................................... 6-24 644-OFFSIE--D SEGAL-GULA4NI MANUIAL Mr6 rOC 6&5-GONTAINME-NTLE-AKAGE-RAT-E--TE-S-4NG-4ROGRAM. 6 26 ARKANSAS UNIT 2 XVII AmendmertNo. 21,60,91,94,157, 176,193

ADMINISTRATIVE CONTROL CORE OPERATING LIMITS REPORT 6.9.5 The core operating limits shall be established and documented in the 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:

1) !The ROCS-and-DT-Gomputer Godes-for-Nuolear-Design 2 r-GE-NP-D-266-P--A-April4983-"Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, "(WCAP-1 1596-P-A), "ANC: A Westinghouse Advanced Nodal Computer Code" (WCAP-10965-P-A), and "ANC: A Westinghouse Advanced Nodal Computer Code: Enhancements to ANC Rod Power Recovery" (WCAP-1 0965-P-A Addendum 1) (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 and Group P CEA Insertion Limits, and 3.2.4.b for DNBR Margin).
2) "CE Method for Control Element Assembly Ejection Analysis," CENPD-0190-A 7 JaRuary 4976 (Methodology for Specification 3.1.3.6 for Regulating and Group P CEA Insertion Limits and 3.2.3 for Azimuthal Power Tilt).
3) "Modified Statistical Combination of Uncertainties, CEN-356(V)-P-A, Revision 01-P-A7 May4988 (Methodology for Specification 3.2.4.c and 3.2.4.d for DNBR Margin and 3.2.7 for AS I).
4) "Calculative Methods for the CE Large Break LOCA Evaluation Model," CENPD-132-PT August4974 (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).

5)t "Galsuatka~eh~eds f the GE Larg Be ek-LQGA-E-valuati~nMGei," GENPD 132 P Supplenent 1, February 1975 (Methodology-foei-Speefiiation 3.1.1.1 for MTC, 3.2.1 for-L nearHeat-Rate324or-AzimuthaI-P-owerTiltT-and-. -72for-AS

6) "Caloulational-Methods-for-4he GE L~arge-reak-L-OGE-valuatiorn-Model,-~lPD 132 P.

Supplement-2-P-l-July4-97-54Methodoloogyor--Specifiration34-t-4or-M1C-3 7 2-for Linear-Heat-Rate32-.3-forAzimuthah-Power-Filt,-and-3-7-for-ASI-

7) "Cai~ulative~etheds-or-the-GE-L-arFgBreakL-OA-Evauatior-Mode~for-4he-Arlatysis of-GE-and-W-Designed-NSSS4-GE-N-43W2-Supplement-3-P-A,-June-I85 (Methodology-for-SpeGifioation 3.1.1.4 for-M r r-Linear-Heat-RateT-,3..3 for-Azimuthat-P-ower--FTt,-and-3-for-ASt)-
85) "Calculative Methods for the CE Small Break LOCA Evaluation Model," CENPD-137-PT August4974 (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).

98G-ralGulative-Methodsior-the GE mall-Break-L-O&E-valuation-Modelj-GE-PD 137, SupplementA-P-FJanuary19 Methodology-or-Speoifiration-34.1-.4for-MT-C 3.2.4for-Unear-Heat-Rato, I fo muthal-Power-Tilt,-and3.2-.74orASI).

ARKANSAS - UNIT 2 6-21 Amendment No. 157,464,469,1-9,

ADMINISTRATIVE CONTROL CORE OPERATING LIMITS REPORT

10) "Galculative-Methods4or-4he-GE-SmalBreak-L-OGA-E-valuation-ModeI-GENP-D-4,37-,

Supplement 2-P ,-dated-April 7 98 4Methodologyfor-SpeGif ration-3.41.14-for MT-CW2.--for-LUnear-Heat-Rate,-323-for-Azimuthal-Power-Tilt,-and-3.2-7-for-ASI);

146) "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System,' December 1981 (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 CEA Position, 3.1.3.6 for Regulating CEA and Group P Insertion Limits, and 3.2.4.b for DNBR Margin).

127) "Technical Manual for the CENTS Code," CENPD 282-P-Ar-February4-991-(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 CEA Position, 3.1.3.6 for Regulating and Group P Insertion Limits, and 3.2.4.b for DNBR Margin.

434-Letterr- 7 D-arr-(NRG 4o-F-4MStern-(GE-dated-JuneA3Z-497-54(NRG-Staff-Review-of the-Gombustion-Engineering-E-GGS-E-valuation-Model)-NRG-approvat-for-64.5.4 6.9.5.1.5, and 6.9.5.1.8 methedogogies.

44)-LetterB-O .-P-arr-(NRG)toA-E-Scherer-(GE4,-dated-December-9,-497(54NRG-Staff Review-of-the-P-roposed-Combustion-E-ngineering -E-GGS-Evaluation-Model-changes).

NRG-approvakfor-&.9 7 546-methodology.

4) -L-etter4-Knieq(NRG-to-A.E--Sherer4GE4-}dated-September-2-7-,4977(E-valuation-of Topioal-Reports-GE-NP-4433,-Supplement-3-P-and-CE-NP-D137,-Supplement4-P4 NRG-approval-for--54.-9-methodology.

46y-L-etter+/-&NAG38403,-dated-Maroh 20,1984-R.-Miller-(NRC4oM-Griffin-(AP&Lt=

GESEGGode-Verific~ation-NRCapprova-for methodology.-

17) "CalGlative-Methods-or-4he GENuclear-oef-arge-Beak-LOCA-E-valuation-ModelS GENP-D432-P-,Supplement-4-P-A-Revision-44Methodology-for-Specification 344-.1for-MTUC3-24for-LinearHeat-Rate-3.2-for-AzimuthakP-ower-Tilt and 32--forAS1
8) "Implementation of ZIRLO Material Cladding in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A (modifies CENPD-132-P and CENPD-137-P as 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).
9) "Qualification of the Two-Dimensional Transport Code PARAGON," WCAP-16045-P-A (may be used as a replacement for the PHOENIX-P lattice code as the 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 and Group P CEA Insertion Limits, and 3.2.4.b for DNBR Margin).

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.

ARKANSAS - UNIT 2 6-21a Amendment No. 4,5,464,4&9,4-7-9,4-82, 49,244,

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. 457,4164,459,1-7-9,482, 407,244,

Attachment 3 2CAN050405 List of Regulatory Commitments to 2CAN050405 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE (Check one) SCHEDULED COMPLETION COMMITMENT DATE (If Required)

ONE- CONTINUING TIME COMPLIANCE ACTION The cycle specific COLR will contain the complete x identification of each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

The upper design limits for ANO-2 fuel will be x limited to mFDI values of 652 for the majority of the fuel assemblies and 712 for a fraction of the fuel pins in a limited number of assemblies (no more than eight fuel assemblies).

Enclosure 1 2CAN050405 Supplemental Information Describing the use of the Westinghouse Nuclear Physics Code Package for Arkansas Nuclear One, Unit 2