2CAN051001, Response to Request for Additional Information Associated with Core Operating Limits Report References

Response to Request for Additional Information Associated with Core Operating Limits Report References

ML101400028
Person / Time
Site:	Arkansas Nuclear, Waterford
Issue date:	05/17/2010
From:	Walsh K Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN051001
Download: ML101400028 (19)

Text

rEntergy Operations, Inc.

1448 S.R. 333

~Entergy Russellville, AR 72802 Tel 479-858-3110 Kevin T. Walsh Vice President - Operations Arkansas Nuclear One 2CAN051 001 May 17, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Response to Request for Additional Information Associated With Core Operating Limits Report References Arkansas Nuclear One Unit 2 Waterford Steam Electric Station Unit 3 Docket No. 50-368 Docket No. 50-382 License No. NPF-6 License No. NPF-38

References:

1.

2CAN050902, License Amendment Request to Revise Technical Specification 6.6.5, Core Operating Limits Report, 5/15/09 [ADAMS Accession Number ML ML091350488].

2.

2CANl 10901, Response to Request for Additional Information Technical Specification 6.6.5, Core Operating Limits Report, 11/10/09 [ADAMS Accession Number ML093160202].

3.

W3F1-2009-0017, License Amendment Request NPF-38-281 to Revise Technical Specification 6.9.1.11, Core Operating Limits Report, 5/22/09

[ADAMS Accession Number ML091470271].

4.

NRC E-mail Request for Additional Information re: License Amendment Request for Administrative Changes to Technical Specification 6.6.5, "Core Operating Limits Report", 9/23/2009 [ADAMS Accession Number ML092660560].

5.

NRC Request for Additional Information, License Amendment Request, Revise TS 6.9.1.11, "Core Operating Limits Report" to Add New Analytical Methods to Support Implementation of Westinghouse Next Generation Fuel, 9/17/2009 [ADAMS Accession Number ML092520124].

Dear Sir or Madam:

In Reference 1, 2, and 3, Entergy Operations, Inc. (Entergy) proposed a change to the Arkansas Nuclear One Unit 2 (ANO-2) and Waterford Steam Electric Station Unit 3 (Waterford 3) Technical Specifications (TSs) associated with the Core Operating Limits Report (COLR) references.

2CAN051 001 Page 2 of 3 During the submittal review process, the Nuclear Regulatory Commission (NRC) determined that Requests for Additional Information (RAls) was required to complete the review of the Entergy submittals (Reference 4 and 5).

A draft of the responses to the RAIs was previously provided to the NRC. During the review of these draft responses, additional clarifications were provided in conference calls between the NRC and the two subject Entergy sites. The responses, in accordance with the aforementioned clarifications, are included in Attachment 1 to this letter. Attachments 2 and 3 contain revised markups of affected TS pages for ANO-2 and Waterford 3, respectively.

This letter contains no new commitments. The response is intended to support both the ANO-2 and Waterford 3 amendment requests.

If you have any questions or require additional information, please contact David B. Bice at (479) 858-5338 (ANO) or William J. Steelman at (504) 739-6685 (Waterford 3).

I declare under penalty of perjury that the foregoing is true and correct.

Executed on May 17, 2010.

Sincerely, KTW/wjs Attachments:

1.

Response to Request for Additional Information

2.

Proposed ANO-2 Technical Specification Changes (mark-up)

3.

Proposed Waterford 3 Technical Specification Changes (mark-up)

2CAN051 001 Page 3 of 3 cc:

Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 NRC Senior Resident Inspector Waterford Steam Electric Station Unit 3 P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Attn: Mr. Kaly Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway ATTN: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn ATTN: N.S. Reynolds 1700 K Street, NW Washington, DC 20006-3817 Morgan, Lewis & Bockius LLP ATTN: T.C. Poindexter 1111 Pennsylvania Avenue, NW Washington, DC 20004 bcc:

Waterford'3 Records Center (W-GSB-1 00) to 2CAN051001 Response to Request for Additional Information to 2CAN051001 Page 1 of 6 NRC REQUEST FOR ADDITIONAL INFORMATION The following lists Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) and the Arkansas Nuclear One Unit 2 (ANO-2) and the Waterford Steam Electric Station Unit 3 (Waterford 3) responses associated with the respective Core Operating Limits Report (COLR) reference submittals.

ANO-2 Request [Reference 61 Based on the guidance specified in Generic Letter 88-16 (GL 88-16), each approved methodology listed in Technical Specification (TS) 6.6.5.b should support a calculation for a cycle-specific core operating limit listed in TS 6.6.5.a. In other words, the methodologies listed in TS 6.6.5.b should identify its supporting role for the cycle-specific parameters listed in TS 6.6.5.a, in orderto be qualified as a part of TS 6.6.5.b.

Please provide:

(a) a description to explain why no date of approval and use of the approved methodologies are proposed for TS 6.6.5.b; (b) identification of cycle-specific parameter listed in TS 6.6.5.a corresponding to an approved methodology listed in TS 6.6.5.b with respect to a real application to the current reload analysis; and (c) a justification that all the proposed methodologies are qualified to be listed in TS 6.6.5.b.

Waterford 3 Request [Reference 71 The NRC staff has reviewed the application and determined that the following information is needed to complete its review:

1.

With respect to the list of core operating limits listed in Insert 1 in Attachment 2 to the May 22, 2009 letter, please provide the following information:

A clarification that the proposed list consists of all independent cycle-specific core operating limits; 0

A description of the relationship between TS 3.2.1, "Linear Heat Rate," and TS 3.6.1.5, "Air Temperature, Containment"; and The reason why a parenthesis (Linear Heat Rate, 3.2.1) follows TS 3.6.1.5.

2.

Based on the guidance specified in GL 88-16, each approved methodology listed in TS 6.9.1.11.1 should support a calculation for a cycle-specific core operating limit listed in the proposed Insert 1. It appears that the list of analytical methodologies in proposed Insert 2 in Attachment 2 to the May 22, 2009, letter is not in compliance with GL 88-16 guidance; therefore, please provide the following information:

Describe how each of the five methodologies listed in Insert 2 (e.g., Nuclear Engineering Methodologies) support the Insert 1 cycle-specific parameters; to 2CAN051001 Page 2 of 6 Describe the relationship between the cycle-specific parameters in proposed Insert 1 and the methodologies in proposed Insert 2; and Provide justification that all of the analytical methodologies listed in proposed Insert 2 are qualified to be listed in TS 6.9.1.11.1.

Further clarifications were obtained via several conference calls between the NRC and Entergy over the last few months. During one of the calls, following a discussion related to a draft response to address the above questions, it was agreed that the draft response would be acceptable provided Entergy Operations, Inc. (Entergy) agrees to retain the parenthetical ties currently listed in the COLR. Entergy had previously proposed to delete this information from the COLR TS. As a result of this call, Entergy agreed to formally submit the draft response letter once modified to retain the aforementioned parenthetical ties.

RESPONSE

The ANO-2 and Waterford 3 methods listed in the response are from the COLR changes submitted in letter 2CAN 110901 (ANO-2) [Reference 3] and W3F1-2009-0017 (Waterford 3)

[Reference 5].

The response to ANO-2 Request Part (a) provides the background information and justification for not including approval dates of the listed methodologies, as previously approved by the NRC [Reference 3]. However, with regard to describing the use of each methodology, Entergy agrees that this information may be retained in the TSs without undue burden to the licensee.

As a result, revised markup pages of the affected TS pages are included in Attachment 2 of this letter.

The response to Waterford 3 Request Part 2 of Question 1 associated with the relationship between containment air temperature and linear heat rate is contained within a previous submittal letter W3F1-2007-0037 [Reference 8].

W3F1-2007-0037 states that the limitation on containment minimum average air temperature ensures that the Emergency Core Cooling System (ECCS) is capable of maintaining a Peak Clad Temperature (PCT) less than or equal to 2200°F during large break Loss of Coolant Accident (LBLOCA) conditions. A higher containment average air temperature results in a higher post accident containment pressure, a higher reflood rate, and therefore a lower PCT.

With regard to Waterford 3 Request Part 1 and 3 of Question 1, the proposed list consists of all independent cycle specific core operating limits. The parenthesis (Linear Heat Rate, 3.2.1) was added for clarification because TS 3.6.1.5 is tied to the TS 3.2.1 COLR requirements.

TS 3.2.1 COLR restricts linear heat rate requirements when TS 3.6.1.5 Action A is entered.

Based upon one of the aforementioned conference calls with the NRC, the response to ANO-2 Request Part (b) and (c) and Waterford 3 Request Question 2 will be limited to a summary of the methodologies used for each parameter.

A summary of how the listed methodologies and analyses for a given parameter are used to determine the associated operating limit listed in the COLR is provided below.

to 2CAN051 001 Page 3 of 6 SHUTDOWN MARGIN The value for Shutdown Margin (SDM) is determined via the PHOENIX-P / ANC methodology.

The neutron cross sections used in ANC are calculated by the PARAGON methodology in lieu of the PHOENIX-P methodology. The primary use of PARAGON is to generate nuclear data for use in Westinghouse core simulator codes (PHOENIX-P / ANC). The NRCapproved the PARAGON topical for this purpose.

The cycle specific value for SDM is confirmed by the transient analyses performed in accordance with the CENTS methodology to meet all appropriate licensing criteria provided the COLR-specified SDM is maintained.

MODERATOR TEMPERATURE COEFFICIENT The values for Moderator Temperature Coefficient (MTC) are determined via the PHOENIX-P / ANC methodology. The neutron cross sections used in ANC are calculated by the PARAGON methodology in lieu of the PHOENIX-P methodology. The primary use of PARAGON is to generate nuclear data for use in Westinghouse core simulator codes (PHOENIX-P / ANC). The NRC approved the PARAGON topical for this purpose.

The values for MTC determined as stated above are confirmed by the large break and small break LOCA (LBLOCA and SBLOCA) and the transient code (CENTS). The Zirconium Diboride (poison), fuel cladding (ZIRLO and Optimized ZIRLO) and the design of the fuel assembly (Next Generation Fuel [NGF]) analyses have a secondary (minor) effect on the specific values and, therefore, are also used to confirm the MTC values to be acceptable throughout the new operating cycle.

The applicability of the methodologies listed above for the NGF and reactor cores incorporating NGF assemblies is documented in the NGF topical. The NRC Safety Evaluation (SE) for the NGF topical contained a condition requiring the NGF topical to be included in the COLR listing.

CEA POSITION The transient analyses performed using the CENTS methodology demonstrate that all appropriate licensing criteria is met assuming the plant is operated within the COLR-specified power reduction requirements for Control Element Assembly (CEA) position deviations.

INSERTION LIMITS The CEA insertion limits are specified following an iterative process that balances reactivity management during normal operations with the consequences of off-normal occurrences that initiate from partially rodded conditions. The transient analyses are performed in accordance with the CENTS and CEA Ejection methodologies to confirm the effects of CEA insertion.

to 2CAN051 001 Page 4 of 6 The cycle specific effects of CEA insertion on the reactor core are determined via the PHOENIX-P / ANC methodology. The neutron cross sections used in ANC are calculated by the PARAGON methodology in lieu of the PHOENIX-P methodology. The primary use of PARAGON is to generate nuclear data for use in Westinghouse core simulator codes (PHOENIX-P / ANC). The NRC approved the PARAGON topical for this purpose.

LINEAR HEAT RATE The Linear Heat Rate (LHR) limits are specified following an iterative process that balances fuel rod power peaking with the consequences of off-normal occurrences that initiate from peak fuel rod power conditions. LBLOCA and SBLOCA, which use as input certain outputs from the Fuel Rod Design analysis, demonstrate that all appropriate licensing criteria are met assuming the plant is operated below the COLR-specified LHR limit.

The Fuel Rod Design analyses (performed in accordance with the Fuel Rod Maximum Allowable Gas Pressure methodology for Waterford only) assume that the peak power fuel rod in the core may operate at the COLR LHR limit for a period of time.

The ZIRLO and'Optimized ZIRLO, and the Zirconium Diboride methodologies have a secondary (minor) impact on the limits and, therefore, are also used to confirm the LHR limit to be acceptable throughout the new operating cycle.

The applicability of the methodologies listed above for the NGF and reactor cores incorporating NGF assemblies is documented in the NGF topical. The NRC SE for the NGF topical contained a condition requiring the NGF topical to be included in the COLR listing.

AZIMUTHAL TILT Tilt limits are specified following an iterative process that balances the need to accommodate some degree of tilt with the consequences of off-normal occurrences that initiate from such tilted conditions. Transient analyses including CEA Ejection (and CENTS for Waterford),

LBLOCA, and SBLOCA demonstrate that all licensing criteria are met assuming the plant is operated within the COLR-specified tilt limits.

The ZIRLO and Optimized ZIRLO, and the Zirconium Diboride methodologies have a secondary (minor) impact on the limits and, therefore, are also used to confirm the tilt values to be acceptable throughout the new operating cycle.

The applicability of the methodologies listed above for the Next Generation Fuel (NGF) and reactor cores incorporating NGF assemblies is documented in the NGF topical. The NRC SE for the NGF topical contained a condition requiring the NGF topical to be included in the COLR listing.

to 2CAN051001 Page 5 of 6 DNBR MARGIN The Departure from Nucleate Boiling Ratio (DNBR) is calculated by using the ABB Critical Heat Flux and the WSSV methodologies for the NGF assembly. The reactor core power distributions used to determine the DNBR are determined via the PHOENIX/ANC methodology. The neutron cross sections used in ANC are calculated by the PARAGON methodology in lieu of the PHOENIX-P methodology. The primary use of PARAGON is to generate nuclear data for use in Westinghouse core simulator codes (PHOENIX-P / ANC).

The NRC approved the PARAGON topical for this purpose. The Modified Statistical Combination of Uncertainties (MSCU) methodology is used to calculate the uncertainty for the DNBR margins. These uncertainties are associated'with the Core Protection Calculators (CPCs) used to calculate DNBR during power operation. Transient analyses using the CENTS methodology provide input.into the DNBR calculations to demonstrate that the minimum DNBRs predicted to occur during all Anticipated Operational Occurrences (AOOs) remain above the DNBR limit.

The applicability of the methodologies listed above for the NGF and reactor cores incorporating NGF assemblies is documented in the NGF topical. The NRC SE for the NGF topical contained a condition requiring the NGF topical to be included in the COLR listing.

AXIAL SHAPE INDEX Axial Shape Index (ASI) limits are specified following an iterative process that balances the need to define a region of acceptable axial power distribution behavior during normal operation with the consequences of off-normal occurrences that initiated from the defined ASI range.

The MSCU methodology is used to calculate the uncertainty in the ASI range (also a CPC monitored parameter). The LBLOCA and SBLOCA analyses demonstrate that all appropriate licensing criteria are met assuming the plant is operated within the COLR-specified ASI limits.

The ZIRLO, Optimized ZIRLO, and the design of the fuel assembly methodologies have a secondary (minor) impact on the limits and, therefore, are also used to confirm the ASI values to be acceptable throughout the new operating cycle.

The full range of ASIs allowed by the COLR limit is considered in the thermal-hydraulic analyses performed in accordance with the ABB Critical Heat Flux methodology. The Zirconium Diboride (poison) methodology has a secondary (minor) effect on the values and, therefore, is also used to confirm the ASI values to be acceptable throughout the new operating cycle.

BORON DILUTION (Waterford 3 only)

The boron dilution limits are created using reactivity inputs determined via the PHOENIX-P / ANC methodology., The neutron cross sections used in ANC are calculated by the PARAGON methodology in lieu of the PHOENIX-P methodology. The primary use of PARAGON is to generate nuclear data for use in Westinghouse core simulator codes (PHOENIX-P / ANC). The NRC approved the PARAGON topical for this purpose.

to 2CAN051 001 Page 6 of 6 BORON CONCENTRATION (Waterford 3 only)

The Mode 6 RCS boron concentration limits are created using reactivity inputs determined via the PHOENIX-P / ANC methodology. The neutron cross sections used in ANC are calculated by the PARAGON methodology in lieu of the PHOENIX-P methodology. The primary use of PARAGON is to generate nuclear data for use in Westinghouse core simulator codes (PHOENIX-P / ANC). The NRC approved the PARAGON topical for this purpose.

Conclusion Based on this information, Entergy believes the listed methodologies are relevant and necessary to envelope the processes required to ensure core operating limits are derived and appropriately verified to be acceptable prior to each restart following a core design change (such as a refueling outage). Without the application of any one methodology, the associated final core operating limit may not be sufficient to protect the specified acceptable fuel design limits (SAFDLs).

REFERENCES

1.

Arkansas Nuclear One - Unit 2, Core Operating Limits Report for Cycle 21, Revision 0.

2.

WSES-3 Cycle 17 Core Operating Limits Report, Revision 0, 11/13/09 [ADAMS Accession Number ML093240083].

3.

2CAN1 10901, Response to Request for Additional Information Technical Specification 6.6.5, Core Operating Limits Report, 11/10/09 [ADAMS Accession Number ML093160202].

4.

2CAN050902, License Amendment Request to Revise Technical Specification 6.6.5, Core Operating Limits Report, 5/15/09 [ADAMS Accession Number ML091350488].

5.

W3F1-2009-0017, License Amendment Request NPF-38-281 to Revise Technical Specification 6.9.1.11, Core Operating Limits Report, 5/22/09 [ADAMS Accession Number ML091470271].

6.

NRC E-mail Request for Additional Information re: License Amendment Request for Administrative Changes to Technical Specification 6.6.5, "Core Operating Limits Report",

9/23/2009 [ADAMS Accession Number ML092660560].

7.

NRC Request for Additional Information, LicenseAmendment Request, Revise TS 6.9.1.11, "Core Operating Limits Report" to Add New Analytical Methods to Support Implementation of Westinghouse Next Generation Fuel, 9/17/2009 [ADAMS Accession Number ML092520124].

8.

W3F1-2007-0037, License Amendment Request NPF-38-271 to Support Next Generation Fuel, 8/2/07 [ADAMS Accession Number ML072180042].

to 2CAN051 001 Proposed ANO-2 Technical Specification Changes (mark-up)

ADMINISTRATIVE CONTROLS 6.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining part of a reload cycle, and shall be documented in the COLR for the following:

3.1.1.1 Shutdown Margin - Tavg > 200OF 3.1.1.2 Shutdown Margin - Tavg < 200OF 3.1.1.4 Moderator Temperature Coefficient 3.1.3.1 CEA Position 3.1.3.6 Regulating and Group P CEA Insertion Limits 3.2.1 Linear Heat Rate 3.2.3 Azimuthal Power - Tq 3.2.4 DNBR Margin 3.2.7 Axial Shape Index

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1)

"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 (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-A (Methodology for Specification 3.2.4.c and 3.2.4.d for DNBR Margin and 3.2.7 for ASI).

4)

"Calculative Methods for the CE Large BreakLOCA Evaluation Model,"

CENPD-1 32-P (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)

"Calculative Methods for the CE Small Break LOCA Evaluation Model,"

CENPD-1 37-P (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).

6)

"CESEC Digital Simla;tdin o-fa;; Coembustion ERngincering Nuclear Steam Supply System" (Methodology for Specific-ationso 3-1. 1.1 anPd-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 RegulatiEng C-EA.land-Group P IRserti;n Limits, and 3.2.4'.b for DNRl Marrg/*r

ADMINISTRATIVE CONTROLS 6.6.5 CORE OPERATING LIMITS REPORT (COLR) (Continued) 7-6)

"Technical Manual for the CENTS Code," WCAP-15996-P-A, Rev. 1 CENPD 282 P A (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).

87) "Implementation of ZIRLO Material Cladding in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A (modifies CENPD-1 32-P and CENPD-1 37-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).

8-)

"Qualification of the Two-Dimensional Transport Code PARAGON,"

WCAP-1 6045-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).

409) "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs," WCAP-16072-P-A (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Tilt, and 3.2.7 for ASI).

104) "CE 16 x 16 Next Generation Fuel Core Reference Report,"

WCAP-16500-P-A (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt, 3.2.4.b, 3.2.4.c and 3.2.4.d for DNBR Margin, and 3.2.7 for ASI).

121) "Optimized ZIRLOTM, WCAP-12610-P-A and CEN PD-404-P-A Addendum 1-A (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).

123) "Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes," WCAP-16523-P-A (Methodology for Specification 3.2.4.b, 3.2.4.c and 3.2.4.d for DNBR Margin).

134) "ABB Critical Heat Flux Correlations for PWR Fuel," CENPD-387-P-A (Methodology for Specification 3.2.4.b, 3.2.4.c and 3.2.4.d for DNBR Margin and 3.2.7 for ASI).

15)

"Calculative Methods for the CE,Nucler Poe-r Large Brcak LOCA EvaiontA Modol Improvement to 1999 Lago Break LOCA Er M Steam; Cooling Model for Less Than 1 !sec Core Reflood," CENPD 132, Supplement 4 P A, Addendum I P A (Methodology for Specifiato 3.1.1.4 for MTC, 3.2.1 for I n.ear Heat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).

c.

The core operating limits shall be determined such that all applicable limits (e.g.

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling System (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

to 2CAN051001 Proposed Waterford 3 Technical Specification Changes (mark-up)

ADMINISTRATIVE CONTROLS INDUSTRIAL SURVEY OF TOXIC"OR HAZARDOUS CHEMICALS REPORT 6.9.1.9 Surveys. and analyses of major industries in the vicinity of Waterford 3 which could have significant inventories of toxic chemicals onsite to determine impact on safety shall be performed and'submitted to the Commission at least once every 4 years, 6.9.1.10 A survey of major pipelines ( > 4 inches) within a.2-mile radius of Waterford 3, which contain explosive or flammable materials and may represent a hazard t o Waterford 3, including scaled engineering drawings, or maps which indicate.the pipeline locations, shall be performed and submitted to the Commission at least once every 4 years.

CORE OPERATING LIMITS REPORT COLR 6.9.1.11 Core operating limits shall be established and documented in the CORE OPERATlNG-....-\\.

LIMITS REPORT 'prior to each reloadcycle or any remaining part.of a reload cyclea 6.9:1.11.1 The analytical methods used to:determine the coreoperating limits shall be those previously reviewed and approved by the NRC as follows:

1)"The RO and DIT Computer C es for Nuclear Design,>CENPD-266-P-A, and "C-E Methodol r Core Designs

'taining Gadolinia-U Burnable Absorber," CENPD-275-P-.(Methodology for S ifications 3.1:1.1 and-. 1.2 for Shutdown Margins, 3.1 *.3 for MTC,.11.:3.6 f meigulating and group =*E Insertion Limits, 3 i..

Bfo ution (Calculation IBC & IBW); and 3.9. 4oron Concentration).

2) "C-E Metho or Control Element A mbty Ejection Analysis," CENPO-01 90-A.

(Methodolo for Specification 3.1.

for Regulating and group P.C,nsertion Limits and 3.2.3 for imuthal Power Tilt).

3) "bMified Statistical C9 bination of Uncertainties" CEN-. 6(V)-P-A. (Methodology for Sp'ecification 3,2.4 for BR Margin and 3.2.7 for ASI).

4) "Calculative M ods for the C-E Large Break ý,/CA Calculation Model For The Analysis of C-fandW Designed NSSS," CENpD-132, Supplement 3-P-A. (M odology for Specification 3.1.1.3'for MTC, 3.2.1 for L ar Heat Rate, 3:2.3forAzim al Power Tilt and 3.2.7 for ASI),

(Mto Olg efr Speitiatton 31'*2 MC,3..

for Aim alinoearTl

5) "Calculative Methods for the ABB* E Small Break LOCA EvaIu tn Model," CENPD-137-P, Supplement 2-P-A. (Methodology for Specification 3 Il !f MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt and.3.2.7 for ASI).

WATERFORD - UNIT 3 6-20 AMENDMENT NO. 6,12,50, iO 191

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT COLR (Continued)

[

6) "CESEC - DigitaISimulationfor a Combustion Engineering Nuclear Steam Supply System," (CE letter LD-82-001 and NRC SE to>t dated April 3. 1984). (Methodology for Specificati0

":1.1. and3.t1.1.2 for Shutdq'n Margins, 3.1.1:3 for MTC. 3:1.3.1 for Movab eControl Assemblies - CEA Position. 3.1.3.6 for Regulating and group P CEA In ieion Limits, and 3.2.3for Azi al Power Tilt).

7) "Qualification of Reacto ysicMethods for the Pressur W

dWter Reactors of the Entergy System,," ENE P: (Methodology for Spei tions 3.1.1.1 and 3.1.1.2 for Shutdown Margins

.1.1.3 for MTC, 3.1.3.6 for Regu* ng and group P CEA Insertion Limits,. 3.1.2.9 on Dilution (Calculation of CBC BW), and 3.9.1 Boron Concentrati

• 8 "u`Rod Ma~ximumAlloowable Gas P ssure,"CEN-3721.PA (Method y for Specrfication 3.2. 1, Linear Heat Ratea

9) "Technical Description Manu for the CENTS Code." WCAP-96-P-A:

(Methodology for Specificati

3. 11.1 and 3.1.1.2 for S hutdo Margins, 3.1.1.3 for MTC, 3.1.3.1 for Movable ontrol Assemblies - CEA Posi' n;.3.1.3.6 for Regulating and group P CEA Insertion mitsand3.2.3forAzimuthal wer Tilt),.

10). "Calculative M=ihods for. the CE Nuclear Po r Large Break LOCA Evaluation Modelh " CENP 132, Supplement 4-P-A; (M odology for Specification 3.1

.3 for MTC, 3.2.1' for.inrH eat Rate, 3.2.3 for Azimu al Power Tilt:and:3.2.7 for AK

11) '*i mentation of ZIRLO Materi Cladding in CE Nuclear Por FuelAssembly Desig s.* CENPD-404P-A (Mth ologyfdrSpecification3.1.1' forMTC,12 for Linear Heat Rate, 3.2.3 for Az i tha Power Tilt, and 32.7 fo ASI).

12) "Qualification of theP ENIX-P/ANC Nuclear Des'n System For Pressurized Water Reactor Cores." WCAP-1596-P-A; 'ANC: A Westinr ouse Advanced N 6dal Computer Code." WCAP-10965- -A; and "ANC: A Westing se Advanced Nodal Computer Code:

Enhancements to VC Rod Power Recovery,"

CAP-10965-P-A Addendum 1.

(Methodology for pecifications 3.1.1.1 and.1.1.2 for Shutdown Margins, 3.1.1.3 for MTC. 3.1.*3.6 f )(Regulating and group P *tA Insertion Limits. 3.1.2.9 Boron Dilution...,

(Calculation z*CBC *&IB V). and 3,9.:1,B ron C~ncentration),.

13) °Qu iiation of the Two-Dime ional TransPort Code PARAGON,'W P1604 5-P-A (Methodology for Specificatior. 3.1.1.1 and 3.1.1.2 for Shutdown M ar 'ns. 3.1.1.3 for MTC. 3.1.3.6 for Regulating and group P CEA Insertion Limits. 3.1.2*9"Boron Dilution (CalculatiOn of CBC & IBW) nd 3.9.1 Boron Concentration)./

14) 'Implehmentat loion~f,

,rconium Diboride Burnable Absorb rCoatings in CE Nuclear Power Fuel Assem bl sgs" W A PA(

eh ooyfrSpec ification 3.1. 1.L3 for MTC, 3.2.1 for Lear Hea~t Rate. 3.2.3 for Azimutha.ower Tilt, and 3.2.7 for ASI),

t5) 'CE 16 x 1&

!*5wtl~

nFUICreRf~

c ~ e o(V C P1500-P and F uial iSafely Evatrialion for W~estinghouse E lepIric Co an Wetigouse) Topical. Report jTR)

WCAP-l(3500-p. Re:,ion 0."CE lCo, ton Eii6eig Ifiei N*t Goene, a fion F

• uel [

lNGF)} Core. Refer*ence R~eport."

t{ '

odloyf or Spcifica tion 3.1

.1:3 for M TrC.

• .3.2 1 for Linear Heat Rate. 3 2.3 f0r Azimuthal Powier-Tilt. 3. 24b.b 32,.c. an]d 3 2.4.d toi

.DN BR M jrin.an dR 3 2n 7 or A S t.

.AMENDMENT NO. n2 146,i-56 WATERFORD.--UNIT 3 6-20a 102. 1.

11..

1..

200... 1..

2.1 4

ADMINISTRATIVE CONTROLS.

CORE OPERATING LIMITS REPORT COLR (Continued)

/

16) 'OptimizedZIRLO'." WCAP-126 10-P-A and CENPD-404-P-A Addendum 1-A, (Methodology for.1Specification13...13foi MTq3.21 for Linear Heat Rate. 3.2.3 for' Azimuthal Power Tilt' and 32.7 for.ASI), i--

17) "Westinghouse Correlations WS'SV and WSSV-T fdrPre icting Critical Heat Flux in' Rod Bundles with Side-Supporled Mixing Vanes." WCA'P-16523-P-A and Final Safe6 EvAluation for Westinghouse'Electric Company (Wesfinghouse) Topical Report3,KTR),

WCAP-!6523-P, "Westiglg*use Correlations WSSV and WSSV-T for Predicti'ng Critical

/Heal Flux in Rod Bundlls with Side'.SupporteoY'Aixing Vanes." (Methodolo/

for

/

Specification 3.2.4.b,3.2.4.c.and.3:2.4.d for'DNBR Margin).

18)"A8B Critical Heal Flux Correlatijo 1sfor PWR Fuel," CENPD-i-P-A (Methodology foi Specificatia '3.2 4.b, 3.2.4.c. and A

.4d for DNBR Margin and./2.7 for A SI).

19) ;C ulalive Methods rfOrtheCENuclear Power Large *reak LOCA Evatuation Model -

mpovement to 1999 Lar 9,'Break LOCA EM Steam Co ing Model for Less Thant-iidSec Core Reflood" CENPD-y32. Supplemcnt 4-.P-A, Addeduitm 1-P and Final SafetyfEvaluation

'f4(Westinghouse Elýcric Company (Westinghous.eý

.opical Report (TRJ) CE

_-132 L

I.2Supplenment 4-P-A/Addendum I-P. Calculativel tlhods for the CE [Cor is1tion Engineering] Nuel'ear Power Large Break LOCA,4valuation Model - Imp,vement to 1999 Large Break L-OCA EM Steam Cooling Mode 14or Less Than 1 in/sec ore:Reflood (Metlhodolog/yfor Specification 3. 1.1,3 for MTfC, 3.2.1 for Linear Heat Rate. 3.2.3for Azirnuthal' Power Tilt. and 3:2.7 for ASII

/!

6-9..11.2 The core operating limits shall bedetermined such that~all applicable limits (e.g., fuel thermal limits, core thermal-hydraulic limits. ECCSt limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysisare met.

6.9.1.11.3 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be'provided upon issuance; for each reload cycle, to the NRC SPECIAL REPORTS 6.9.2 Special reports shall be submitted in accordance with IO.CFR 50.4 within thetime period specified for each report.

6.10 Not Used WATERFORD - UNIT 3 6_26b AMENDMENT NO, M66, 214

INSERT 1 for the following:

3.1.1.1. SHUTDOWN MARGIN - ANY CEA WITHDRAWN 3.1.1.2 SHUTDOWN MARGIN - ALL CEAS FULLY INSERTED 3.1.1.3 MODERATOR TEMPERATURE COEFFICIENT 3&1.2.9 BORON DILUTION 3.1.3.1 CEA POSITION 3.1.3.6 REGULATING AND GROUP P CEA INSERTION LIMITS 3.2.1 LINEAR HEAT RATE 3.2.3 AZIMUTHAL POWER TILT - Tq 3.2.4 DNBR MARGIN 3.2.7 AXIAL SHAPE INDEX 3.6.1.5 AIR TEMPERATURE, CONTAINMENT (Linear Heat Rate, 3.2.1) 3.9.1 BORON CONCENTRATION INSERT 2 1-)

"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-1 0965-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, 3.2.4.b for DNBR Margin, 3.1.2.9 for Boron Dilution, and 3.9.1 for Boron Concentrations).

2)

"CE Method for Control Element Assembly Ejection Analysis,"

CENPD-0190-A (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-A (Methodology for Specification 3.2.4.c 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 (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)

"Calculative Methods for the CE Small Break LOCA Evaluation Model,"

CENPD-1 37-P (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).

6)

"Technical Manual for the CENTS Code," WCAP-1 5996-P-A, Rev. 1 (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).

7)

"Implementation of ZIRLO Material Cladding in CE Nuclear Power Fuel Assembly Designs," CENPD-404-P-A (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).

INSERT 2 Continued

8)

"Qualification of the Two-Dimensional Transport Code PARAGON,"

WCAP-1 6045-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, 3.2.4.b for DNBR Margin, 3.1.2.9 for Boron Dilution, and 3.9.1for Boron Concentrations).

9)

"Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs," WCAP-16072-P-A (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Tilt, and 3.2.7 for ASI).

10)

"CE 16 x 16 Next Generation Fuel Core Reference Report,"

WCAP-16500-P-A (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt, 3.2.4.b, 3.2.4.c and 3.2.4.d for DNBR Margin, and 3.2.7 for ASI).

11)

"Optimized ZIRLOTM, WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A (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).

12)

"Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes," WCAP-16523-P-A (Methodology for Specification 3.2.4.b, 3.2.4.c and 3.2.4.d for DNBR Margin).

13)

"ABB Critical Heat Flux Correlations for PWR Fuel," CENPD-387-P-A (Methodology for Specification 3.2.4.b, 3.2.4.c and 3.2.4.d for DNBR Margin and 3.2.7 for ASI).