2CAN070402, License Amendment Request to Support Cycle 18 Core Reload Arkansas Nuclear One, Unit 2
| ML041960419 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/08/2004 |
| From: | Forbes J Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2CAN070402 | |
| Download: ML041960419 (80) | |
Text
A^ Entergy Operations, Inc.
E41448 E ntKe vo O X S.R. 333 Russellvlle. All 72802 Tel 479-858-4888 Jeffrey S. Forbes Vice President Operations ANO 2CAN070402 July 8, 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
REFERENCE:
- 1. Entergy letter to the NRC dated May 12,2004, License Amendment Request to Support Cycle 18 Core Reload" (2CAN050405)
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 new commitments, which are 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.
2CAN070402 Page 2 of 3 By letter dated May 12, 2004 (Reference 1), Entergy, in part, proposed changes to the list of analytical methods included in TS 6.9.5. One additional analytical method is proposed to be reflected in TS 6.9.5. The May 12, 2004 letter (Reference 1) is superseded in its entirety by this letter. Revisions bars are included in the attachments to reflect the changes to the May 12, 2004, letter (Reference 1) needed to support the additional reference.
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 July 8, 2004 Sincerely, JSF/dm Attachments:
- 1. Analysis of Proposed Technical Specification Change
- 2. Proposed Technical Specification Changes (mark-up)
- 3. List of Regulatory Commitments Enclosure I - Supplemental Information Describing the Use of the Westinghouse Nuclear Physics Code Package for Arkansas Nuclear One, Unit 2
2CAN070402 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. Box310 London, AR 72847 U. S. Nuclear Regulatory Commission Aftn: 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 2CAN070402 Analysis of Proposed Technical Specification Change to 2CAN070402 Page I of 14
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 supplement the list of the analytical methods referenced in Technical Specification (TS) 6.9.5.1 with a Westinghouse Nuclear Physics code package, a methodology that will support the use of ZIRLO fuel cladding, and a methodology that will allow the use of zirconium diboride (ZrB 2) burnable absorber coating on fuel pellets.
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; the addition of a topical report related to ZIRLO fuel cladding; and the addition of a reference to the topical report that supports the use of ZrB2 burnable absorber coating on fuel pellets.
Westinghouse Nuclear Physics Code Package TS 6.9.5.1, Item I '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 2CAN070402 Page 2 of 14 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-01 90-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-137. 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 forthe 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 2CAN070402 Page 3 of 14 Zirconium Diboride Burnable Absorber Coating A change is proposed to TS 6.9.5.1, Core Operating Limits Report. The proposed change will add WCAP-1 6072-P-A to the list of referenced topical reports (item #10). The NRC Safety Evaluation (SE) for the topical report was dated May 6,2004 (TAC No. MB8721). "-A" will be included in this application, however, the "-A" version has not yet been published by Westinghouse.
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-1 1596-P-A, WCAP-1 0965-P-A, WCAP-10965-P-A, Addendum 1 and WCAP-16045-P-A; to allow the use of ZIRLO fuel cladding by the addition of CENPD-404-P-A; and the addition of WCAP-16072-P-A, to allow the use of ZrB2 as a burnable absorber coating on the fuel pellets. 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, ROCS and DIT 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" to 2CAN070402 Page 4 of 14 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 essentially the same as those obtained with the current ROCS and DIT 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 In a 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 2CAN070402 Page 5 of 14 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-132, "Calculative Methods for the CE Large Break LOCA Evaluation Model" and CENPD-137, 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.
Zirconium Diboride Burnable Absorber Coating A change in the burnable absorbers is planned for ANO-2 Cycle 18 (spring 2005). The change reflects a transition from Erbia integral burnable absorber, which is mixed in the fuel, to a burnable absorber using ZrB2 coating. Zirconium diboride burnable absorbers were introduced to Westinghouse cores in the mid-1980's as Integral Fuel Burnable Absorbers (IFBA). The description of IFBA was originally documented in WCAP-1 0444-P-A, "Reference Core Report Vantage 5 Assembly" September 1985. The use of ZrB2 burnable absorbers in Combustion Engineering (CE) fuel assemblies is addressed separately in WCAP-16072-P-A.
The topical report, WCAP-16072-P, includes a detailed description of the ZrB2 burnable absorber and summarizes the evaluation of the effects of the ZrB 2 material characteristics on design and safety analyses, including loss of coolant accident (LOCA) analyses. The current reload methodologies and those requested for use in this letter can be used with the implementation of the ZrB2 burnable absorber. Changes in the safety analyses or safety methods are addressed directly or via reference through WCAP-1 6072-P-A.
ANO-2 TS 2.1.1.2 requires that the peak fuel centerline temperature be maintained less than 50800F (decreasing by 580F per 10,000 MWD/MTU for burnup and adjusting for burnable poisons per approved topical reports). Therefore, the impact of ZrB2 was evaluated in relationship to the fuel centerline temperature. By design ZrB2 is coated onto the outer surface of the uranium dioxide (UO2) fuel pellets prior to loading into the fuel rod cladding tubes rather than being mixed with the U02 directly as is done with other integral fuel burnable absorber materials. Therefore, ZrB 2 does not impact the properties of U02 and does not affect the peak fuel centerline temperature.
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."
Attachment I to 2CAN070402 Page 6 of 14 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 1 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.
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.
to 2CAN070402 Page 7 of 14 ZIRLO Fuel Cladding The topical report CENPD404-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.
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.
to 2CAN070402 Page 8 of 14 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.
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.
Zirconium Diboride Burnable Absorber Coating The final NRC SE for WCAP-16072-P includes five conditions and limitations which require response. The conditions and ANO-2's responses are below.
Condition 1 A license amendment is required to add this TR to the Core Operating Limits Report analytical methods listed in the licensee's TS.
Response I This letter includes the TS change required by this condition.
Attachment I to 2CAN070402 Page 9 of 14 Condition 2 Plant-specific core design guidelines or cycle-specific calculation shall be used to verify that required power margins in the axial cutback regions are maintained within safety analysis limitations.
Response 2 Cycle specific evaluations will be performed as part of the reload efforts to verify that required power margins in the axial cutback regions are maintained within the safety analysis limitations.
Condition 3 Plant TS SRs on MTC [moderator temperature coefficient] validate the physics predictions and ensure that plant operations remain within allowable limits. In addition to current SRs, licensees shall confirm that the peak positive HFP MTC is within the TS limits at the highest RCS soluble boron concentration predicted during full power operation. The peak positive HFP MTC shall be derived by adjusting the measured MTC at HFP BOC conditions to the maximum HFP soluble boron concentration expected during the cycle. In order to ensure a conservative adjustment, a direct measurement of MTC is required at the highest RCS [reactor coolant system] soluble boron concentration predicted during full power operation. This direct measurement is only required for the first application of ZrB 2 IFBA in a CE [Combustion Engineering]
14 x 14 or 16 x 16 fuel assembly design. During the first cycle implementation, Westinghouse shall provide the staff with a letter containing the following information:
- i. Measured HFP BOC MTC (TS SR),
ii. Measured HFP MTC at highest RCS soluble boron concentration, iii. Calculated HFP MTC at highest RCS soluble boron concentration, and iv. Demonstrated accuracy of the calculated HFP MTC within current analytical uncertainties.
In addition, plant procedures used to perform MTC surveillances shall be updated, where appropriate, to reflect the calculated peak positive HFP MTC along with ZrB2 IFBA's distinctive trend in RCS critical boron concentration.
Response 3 ANO-2 intends to fully meet this condition by performing both a hot zero power and a hot full power (HFP) MTC test early in Cycle 18, in accordance with ANO-2 TS. ANO-2 will perform an additional MTC test within seven days of reaching the highest RCS soluble boron concentration predicted during full power operation to confirm that the MTC is within the TS limits at the peak boron concentration. (The allowance to perform the third MTC test +/- 7 day of achieving the highest predicted RCS soluble boron concentration was discussed with the NRC staff per telecom on May 26, 2004.) The appropriate information from the confirmatory test will be sent to Westinghouse to report to the NRC as required by this condition.
Plant procedures will be modified as needed to reflect the calculated peak HFP MTC along with ZrB2 IFBAs distinctive trend in RCS critical boron concentration.
Attachment I to 2CAN070402 Page 10 of 14 Condition 4 Prior to startup following a Condition III or IV event, licensees must evaluate clad hydriding to ensure that hydrides have not precipitated in the radial direction (in accordance with section 3.2 of this SE).
Response 4 Inthe event of a Condition IlIl or IV event at ANO-2, an evaluation of fuel structural integrity with respect to radial hydriding will be performed prior to power ascension.
Condition 5 CEN-372-P-A constraints and limitations with regard to rod internal pressure and DNB propagation must continue to be met. In addition, licensees must ensure that the following two conditions are satisfied:
- a. For Condition I (normal), Condition II (moderate frequency), and Condition IlIl (infrequent) events, fuel cladding burst must be precluded for ZrB2 IFBA fuel rods.
Using models and methods approved for CE fuel designs, licensees must demonstrate that the total calculated stress remains below cladding burst stress at the cladding temperatures experienced during any potential Condition II or Condition III event. Within the confines of the plant's licensing basis, licensees must evaluate all Condition II events in combination with any credible, single active failure to ensure that fuel rod burst is precluded.
- b. For Condition IV non-LOCA events which predict clad burst, the potential impacts of fuel rod ballooning and bursting need to be specifically address with regard to coolable geometry, RCS pressure, and radiological source term.
Response 5 The constraints and limitations of CEN-372-P-A will continue to be met. Analyses, as part of the ANO-2 reload efforts, will be performed in support of the generic implementation of ZrB2 fuel to ensure that cladding bursts are precluded for Conditions 1,11, III and IV events.
5.0 REGULATORY ANALYSIS
5.3 Applicable Regulatorv 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).
to 2CAN070402 Page 11 of 14 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-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"), and WCAP-10965-P-A Addendum 1 ("ANC: A Westinghouse Advanced Nodal Computer Code: Enhancements to ANC Rod Power Recovery").
WCAP-16045-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."
WCAP-16072-P-A, 'Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs" will also be added to the list of analytical methods listed in TS 6.9.5.1.
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).
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 to 2CAN070402 Page 12 of 14 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 will add an NRC approved topical report, WCAP-16072-P-A, to the list of referenced topical reports. The topical report has been previously approved by the NRC for use in Combustion Engineering core designs and as such, the proposed change is administrative in nature and has no impact on any plant configurations or on system performance that is relied upon to mitigate the consequences of an accident. In addition, prior to the use of the ZrB 2 burnable absorber coating, fuel design will be analyzed with applicable NRC staff approved codes and methods.
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.
- 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 adds a reference to the topical report that allows the use of ZrB 2 as a burnable absorber coating on the fuel pellet. The topical report has been previously approved by the NRC for use in Combustion Engineering core designs and as such, the proposed change is administrative in nature and has no impact on any plant to 2CAN070402 Page 13 of 14 configurations or on system performance that is relied upon to mitigate the consequences of an accident. In addition, prior to the use of the ZrB2 burnable absorber coating, fuel design will be analyzed with applicable NRC staff approved codes and methods. This change is administrative in nature and does not create a new or different type of accident than previously evaluated because the design requirements for the facility remain the same.
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 and DIT methodology. Future changes to the values of these limits by the licensee may only be developed using NRC'approved mnethodologies, 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.
The proposed change will add WCAP-16072-P-A to the list of referenced topical reports.
The topical report has been previously approved by the NRC for use in Combustion Engineering core designs and as such, the proposed change is administrative in nature and has no impact on any plant configurations or on system performance that is relied upon to mitigate the consequences of an accident. In addition, prior to the use of the ZrB2 burnable absorber coating, fuel design will be analyzed with applicable NRC staff approved codes and methods.
to 2CAN070402 Page 14 of 14 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 uno 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 2CAN070402 Proposed Technical Specification Changes (mark-up)
WDEX DE-FINITIONS SE-GT4QN' -PAGE-
- 1. EFINTIONS D efinedD ei-n.............................................................................................................................................................
rT rTn edoe r mA. 1 114-Thermal Power ............................................................. 1i1 Rated Thermal Power .............................................................. 1.1 Operational Made Mode ......................................................................................... I Action .............................................................. 1.1 Operabic O perability .............................................................. 1.1 Reportablc urrence .............................................................. 1.1 Contaiment4ntegfity .............................................................. 12 Channel alib ation. ............................................................ 1 Channel Check ............................................................. 12 Channcl Functional Test.............................................................. 13 Core Alteration .............................................................. 13 Shutdown Margin .1...................................................3 IdentifledLeakage .............................................................. 13 Unidentifted-Leakage .............................................................. 1 .
Pre66ure Boundary Loakage .................... .. 14 Azim uthal Power Tilt Tq ......................................................................... 1.1 Dose Equivalent 1 131 .............................................................. 14 E A vcragcDisintegration Energy .............................................................. 1.1 Staggered Test-Basi .1..1 Frequency Notation 1......................................
A ia-ShapeIndex ............................................ ................... 5 Reato-ip-S em-ResponseTme ............................................................. 15 Engineered Safety Feature Response Time ............................................................ 15 PhysiG-S Tests ............... ,,,,,,,,,,.,,.1.5 Software 1..........1 5
Planar Radial Peaking Factor Fxy .............................................................. 15 A n .1 A . I-c A C ..1211 T l __
1_ I _ .1 . - I A I 7 iNKrldkFqWkb UN! i ad v Amenoment No. 24,91,457 I
DE-4NIMOINS SECTION -PAGE Uquktiadwasreatment- sy t :m ......................................................................... -45 Member(s) of the Public............... ............................................................................. 16 Purge Purging............................ ............................................................................. - 6 ExcIusion Area ............................. -46 Unrestricted Area ......................... ............................................................................. - 6 Core Operating Limits Rcport ...... ............................................................................. 6 A
rl.1 A K1-A U" I-1ig~ii A~r~p~l~b I All A~~~
A\menameni . . _ . . .
1 No. -- __
-- I -
uihiAY.41 - - ^^
INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS -
SECTION PAGE 2.1 SAFE-TY4LMO Rea tCor Core ....................................................................... ................................... -24 Reactor Coolant System Prccfure.
Rear~~~tOF~ t y te rO~a P es u ........................................ ..................
.......................................... ad _
2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Setpoint ......................................................... .................................. 3 Deleted ................................................................................ ................................... 4 BASES SE-G-TION -PAGE 2.1 SAFETY LIMITS Reactor Core ........................................................... B2 1 Raost or o antS e Pressure .................................... ....................... B2 2.2 LIMITING SAFETY SYSTEM SET-TINGS Reat - - t s............................. ..............................................................
A, U LULI8 .....
A DVAKI2A c I IKIIT III Amendment No. 24,60,77
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3---O-APPLCA ILI ....................
- .............................................................. 3/.0-1 3144-REACTIVITY CONTROL SYSTEMS 3/1.1.1 BORATION CONTROL Shutdown Margin Tas 2000 F ..................................................... 3/1 1 1 Shutdown M argin Ta 00 F........................................................... 3/.1 3 Boron Dilution ..................................................... 3/1 1 1 ModeratoF-Temperature-Goeffient ..................................................... 3/1 5 Minimum ateai . ........................................ 311 6 3/1.1.3 CONTROL ELEMENT ASSEMBLIES CEA Position..................................................... 3/1 1 17 PoGition M ator Channels Operating .................................................... 3/1 1-20 CEA Drop Time ................................ 3/1 1 23 Shutdown CEA Insertion Limit ................................ 3/1 1 21 Regulating and Group P CEA Insertion Limits ................................ 3/1 1 25 ARKANSAS UNIT-2 IV Amendment No. 37,60,157,16 ,229
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 341.2 POW ER-hSTR4BUTION-WMITS 3/14..1 LINEAR-HEATRE ............................................................. 31 2 1 3/1.2.2 RADIAL PEAKING FACTORS .............................................................. 3/1 2 2 3/1.2.3 AZIMUTHAL POWER TILT........................... .......................... 314 2 3 3!1.4.5 RCS FLOWRATE............................................................. 3/1 2 7 314.2.6 REACTOR COGLA.NT COLD LEG TEMPERATURE .............. 28 3/1.2.7 AXIAL SHAPE INDEX ............................................................. 3/1 2 9 3/14.2.8 PRESSURIZER PRSSTUMAIRE ............................................................. 3/1 2 10 3/14.3 !NSTRUMENTAT40N 314.3.2 EGINE-ERE-DSAF-ETY FEATUACTUATION SYSTEM iNSTRUMENTATION .............................................................. 3/1 3 10 314.3.3 MONITORING INSTRUMENTATION Radiation Monitoring instrumenttion ........................................................ 3/1 3 21 FRemotde Shutdown +4nttrumentation......................................................... 3/143 36 Pot Acdnnstrumcntation ...................................................... 3/1 3 39 ARKANSAS UNIT 2 V Amendmont No.124,60a157,163,191, 1a3
NOEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314.4 REACTOR-CGLANT- SYSTEM 314.4.1 REACTOR COOLNT LOOPS AND COOLANT CIRGULATION ............. 31 41 3M.1.2 SAFETY VALVES SHUTDOWN .............................................................. 3!4 43 nIA A e ^ A Dr-Tr 'IfA I I Df- d'A T'% PIA A AI J.1. r I LV I I..............................................................
3M1.4.1 PRESSURIZER .............................................................. 3/11 5 3/1.1.5 STEAM GENERATORS ............................ ................................. 31 16 314.4.6 REACTOR COOLA.NT SY-STEM LEAKAGE Leakage DeteG ion Systemr.3...................................................................... 1 4,13 Reactor Coolant Syctem Leakage ..................................................... 3/1 4 14 31447-CHEMISTRY ................ 3/1 15 34A8-SERGIC ACTIVI .................... ......................................... 3/1 1 18 344-9E-SSURE/EMtPERATURE-ULMS ReactorCoolant ............................................................. 3/14 1 '22 PreSSUrizer ................................................................................................. 3/1 1 25 3/14.4.10 STRUCTURAL INTEGRITY ASME Code Claos 1, 2 and 3 Components ............................................... 3/ 4 26 3M/.4.11 REACTOR COOLANT SYSTEM VENTS .................................................. 3M 4 27 3/1.1.12 LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM 3/1 4 28 314.5 MTRG _WV GRE GOt INIG 77 3/4,.4-SAFETY MNJE-CTION--TANKS ............................................................. 3/154 ARKANSAS UNIT 2 VI Amendment No. 29,60,63,191,193
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314.5.2 ECCS SUBSYSTEMS -Ta' TF 300°F ........................................................ 3/1 5 3 314.5.3 ECCS SUBSYSTEMS T-wg-300OF......................................................... 3 56 314.5.1 REFUELING WATER TANK .............................................................. /1 5 7 3/1.6 CONTAINMENT SYSTEMS 3/1.6.1 PRIMARY CONTAINMENT Gontaimenntegrity ............................................................. 3/1 64 GGntairmentbLeakage ............................................................. 3/1 6 2 GontainretAir Locks................3..............................................................11. 3 /164 ante+nal Pressure, Air Temperatre and Relative Humidity. .31 6 6 Umct I owciuruAidaI IoIeIlIy ......................................................... ' v 0 Qontainm ent Ve ntilation SSIx em ------------------------------------ ------------------ 3/
_1 6 - a 314.6.2 DEPRESSURIZATION, COOLING, AND pH CONTROL SYSTEMS Containment Spray System ..............- 3/146 10
- fiso u r Phosphate (TSP)....................................................................... 314 6 12 Containment Cooling System ............................................................. 3/1 6 11 3I4.6.3 CONTAINME-NT4OLAT4QN VALVES ...................................................... 31 646 3/4;6.4-GOMBBUSTBLE GAS CONTROL Hydrogen-Analyzer ............................................................. 3/4 6148 EAiRKy SA UNIT 2 en Reombiners W ..........................................................
311 4612 ARKANSAS UNIT 2 Vll Amendment No. 60,454,494-,245
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3114.7-4PtLA NSYSTMS 3/14.7.4 TURBLE-GYLE-Safety Valves ..................................................... 3/1 71 Emergenriy-Feedater ..................................................... 31 7 5 OJr__!* Kat
_ IA v7 7 ICE)UH~t %tal 0i [1li.....................................
af3i f f ActiGi........................................................................................................
Main SteamIsolation Valver ...................................................... 3/1 7 10 314.7.2 STEAMGE-NERATOR PRESSUREgEMPERATURE LIMITATION.311 7_1 314.7.3 SERVICE WATER SYSTEM ............................................................. 31 7 15 3/14.7.EMERGE-NGY GOL1NGPND ............................................................. 716 3147 5 FLOO ECRO TIO N ...................................................................... 3/4~zlr"I 3I44-6GONTEROLMR-E-MERGENCY VENTILAT40"AND AIR CONDITIONING SYSTEM ............................................................ 3/1 7 17 JC4..8 SHGGK-SUPPRESSORS (SNUBBERS) ................................................... 3!1 7 22 311.7. SEALED-SOURGE CONTAMINATION ..................................................... 3/1 7 27 314.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System ................................................................ 3/1 7 29 Spray and/or Sprinkler System6 ............................................................... 3/4 7- 33 Fire Hese-Stations .................. ........................................... 31 7 35 3/1.7.11 FIRE BARRIERS ............................................................. 3/1 7 37 3/14742-SPENT-FUEL POOL STRUCTURAL INTEGRITY ..................................... 3/4.- 7-2q3 314-8ELECTRIGAL POWER STEMS Y 314.8.1 A.C. SOURCES pefating ................................................................................................... 3/1 8 1 S tdown..............................................................................................3/18 5 ARKANSAS UNIT 2 VW1l Amendment No. 30.60.62.99.206
NDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4-82ONS 1 E R ST-RBUTION-SYSTE-MS A.G. DistributienR-Operating.......... A.C.
Distnbution Operating~~~.
......................................................i..... 3183!1 8-6 A.G. DistributinShutdown ...................................................... 314-84 D.C. Distribution Operating ....................................................... 311 8 D.C. Distribution Shutdown ...................................................... 3/1 8 10 Gontainment-Penetration Conductor Overcurrent ProteG ive Devicc ... 31 8 11 341.9 REFUELING OPERATIONS 311.9.1 BORON CONCENTRATION ............................................................. 314 9 1 311.9.2 INSTRUMENTATION .. 9 2..............................
2 314.9.3 DECAY TIME AND SPENT FUEL STORAGE ........................................... 311
- 34. G MMUN IPAT.ONS ....................... 311 6 3/4.9.5 R E F UE ING AT CHLIN OPRS
.................................................................................................. 314 9-6 31.9.6 REFUELING MAHINE PERABILI.............................................. 31 9 7 314.9.7 CRANE TRAVEL -SPENT FUEL POOL BUILQNG ................. 314 9 8 31.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION ........................ 31 9 9 311.9.9 WATER LEVEL REACTOR VESSEL ..................................................... 314 9 10 314.9.10 SPENT FUEL POOL WATER LEVEL ........................................................ 311 9 11 314.9.11 FUEL HANDLING AREA VENTILATION SYSTEM.3. . 9 12 31.9.12 FUEL STORAGE .3.. .9............................. 3 114 314.10 SPECIAL TEST EXCEPTIONS 3/44041SHUTQOWN-MARGtN 314 10 1 3 GROUP-HERGHT-,4NSERTION AND PC) 610.2 DISTRIBUTION4LMITS .... 314 10 2 3.10.3 REACTOR COOLANT LOOPS .. 31 10 3 ARKANSAS UNIT 2 IX Amendment No. 29,60
INDEX LIMITING CONDITIONS FOR OPERATION AND SURV.EILLANCE REQUIREMENTS SECTION 4PAGE 3 0E T- GEA MISAGNMENT........................ ....................................... -314 40 4 3/4AO-6MINNIMUME-MPERAT-URE-F-R GRITIGALTY- ................................ .............................
"IIA 4A n
-Mu 3/1.11 RADIOACTIVE EFFLUENTS 3/1.11.1 LIQUID HOLDUP TANKS ................................... ....................................... 3/4 1AA1 3/1.11.2 GAS STORAGE TANKS..................................... , ....................................... 3J44 1-2 314.11.3 EXPLOSIVE GAS MIXTURE. . . . .. . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . .. .
'nIA AA.
at I i m m q3 15 A -V K1O A M~r0Nr4WW) - I- IKI.
ur1im m ^
A Amendment s NO. bU,446 l _ An, He
INDEX BASES SECTION PAGE 31 .0 APP ABILI ............................................................................................ B 31 0 1 314.1 REATIVI ONTR-SYSTMS 311.1.1 BORATION CONTROL.............................................................................. B 31 1 3/1.1.2 BORATION SYSTEMS.............................................................................. B 3/1 1 2 314.1.3 MOVABLE CONTROL ASSEMBLIES B 311 1 3 314.2 POWER DISTRIBUTION LIMITS 3/1.2.1 LINEAR HEAT RATE................................................................................. B 3/1 2 1 3/1.2.2 RADIAL PEAKING FACTORS................................................................... B 3/1 2 2 3/1.2.3 AZIMUTHAL POWER TILT........................................................................ B 31 2 2 3/14.2. DNB -MA GIN......................................................................................... B 3423 314.2.5- RCSF4OW-ATE ........................................... B342 4 314R.2.- AGATOR-COOLANT COLD LEG T-EMPERATURE................................ B 314 2 3/1.2.7 AXIAL SHAPE INDEX ................................................................................ B 3/1 2 1 314.2.8 PRESSURIZER PESR. ................................... B 3/142-4 314.3 INSTRUMENTATION 311.3.1 PROTECTIVE INSTRUMENTATION......................................................... B 314 3 1 314.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION............. B 314 3 1 314.3.3 MONITORING INSTRUMENTATION........................................................ B 31 3 2 ARKANSAS UNIT 2 Xl Amendment No. 21,33,60,191
INDEX BASES SECTION PAGE 3/4.4RE-AGTOR-GOLANT-SYSTE-M 3144REACTORCOANT LOOPS-AND GOOLANT-GCIRGULAAONNB 31 4 1 3/4..2 and 3/1.1.3 SAFETY VALVES....................................................................... B 31 11 3/1.1.5 STEAM GENERATORS ............................................................. B 31 1 2 314.4.6 3M14.4 REACT-OR STEAM GECOOLA.NT-ER TORSSYSTEM LEAKA.GE........................
..................................................................... B 314 3/14 4 32 314.4.7 CHEMISTRY ................................................ B 314 4 4 3/4.4.8SPEGIRC ACTIVITY ..................... ........................................ B 3!1 4 1 3/1.4.9 PRESSURE-TEMPERATURE LIMITS ....................................................... B 3/1 1 5 3/1.4.10 STRUCTURAL- ITEGRITY....................................................................... B 314 4 11 3/14.4.11 RE-AGTL-G LASYSTEM VENTS BB44 3/1 1............................
3/1.1.12 LOW TE-MP-EAT-UROERA REt-SREREG ON i tSYSE.B.......12 2-/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS 3/14.5.1 SAFETY INJECTION TANKS.................................................................... B 3/1 5 1 3/1.5.2 and 3/1.5.3 ECCS SUBSYSTEMS ............................................................. B 31 5 1 3/1.5.1 REFUELING WATER TANK (RA9T) B.........................................
1 3/1 5 2 314.6 CONTAINMENT SYSTEMS 314.6.1 PRIMARY CONTAINMENT ............................................................. B 3/1 6 1 3/1.6.2 DEPRESSURIZATION, COOLING, AND pH CONTROL SYSTEMS ........ B 3/1 6 3 31.6.3 CONTAINMENT4ISOLTION-VA LVEST4 ...................................................... B 3.. 6 6 3/44 -GOMBUST4BLELGAS-CONTROL ............................................................. BB 6 ARKANSAS UNIT 2 Xi! Amendment No. 29,60,63,101,100
INDEX BASES SECTION PAGE 314.7.1 TURBINE CYGLE ............................................................. - B 1 314.7.2 STEAM GENERATOR PRESSUREITEMPERATURE LIMITATION ......... B3 71 3/1.7.3 SERVICE WATER SYSTEM .......................... B 314 7 1 314.7.1 EMERGE-NGY COOLING POND ............................................................. B 314 7 1 314.7.5 FLOOD PROTECTION.............................................................................. B 314 7 4 314.7.6 CONTROL ROOM EMERGENCY VENTILATION AND AIR CONDITIONING SYSTEM ............................................................ B 3/17 5 3M.7.8 SHOCK SUPPRESSORS (SNUBBERS) ................................................... B 314 7- 6 31.7.9 SEALED SOURCE CONTAMINATION ....................................... B 311 7 7 3/4.7-.10 PE-NE P4R N-FNRESBA IERS ............................... .. B 4 314.7-.12 SPENT FUEL POOL ST-RUCT-URAL= INT-EGRIT.... B 3141 77 314.7. PELECTRITCA L POWEB RR SYSEM.............................................................. B...............
3/181 314.8 EL=ECTRICAL= POWER SYSTEMS ................................. B 314 8 1 311.0 REFUELING OPORATIONS 314.9.1 BOR CONCENTRATION.B 314 9 1 3W4.9.2 INSTRUMENTATION /19 4 1..............................
1 3/1.9.3 DECAY TIME ............................... B 31 9 1 314.9.4 CONTAINMENT PENETRATIONS............................................................ B 31 9 1 ARKANSAS UNIT 2 XII Amendment No. 60,62,206
NDEX BASES SECTION PAGE 3I4.965GQMMUNICATIONS ...................... ....................................... B 314 9 2 31.6 REFUE-NG MAGHIN E-PARII.B. 311 9 2 3!1.9.7- GRANE TRAVEL SPENT FUEL STORAGE BUILDING.............. B 314 9 2 311.9.8 COOL.ANT CIRCULATION ............................................................. B 31 9 2 3I4.9.9and 314.9.1 0 WATER LEVEL -REACTOGR VESSEL AND STORAGE POOL WATER LEVEL ............................... B 31 9 3 311.9.11 FUEL HANDLING AREA VENTILATION SYSTEM ............... .................... B 3/1 9 3 314.10 SPECIAL TEST EXCEPTIONS 314.10.1 SHUTDOWN MARGIN ............................................................. B 311 10 1 3/1.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS ... B 3/1 10 1 3 /-4RE- FOR LANT-LOOPS ............................................................. B 34 101 3/14.0. G ENT E SA GN NT................................................................ / 10 1 3/1.10.5 MINIMUM TEMPERATURE FOR CRITICALIP4 ....................................... B 3/1 10 1 3/1.11 RADIOACTIVE EFFLUENTS 3/1.11.1 LIQUID HOLDUPTANKS.......................................................................... B 3/1111 3/1.11.2 GAS STORAGETANKS .....................................................
3/1.11.3 EXPLOSIVE GAS MIXTURE ............................................................. B 3/1 11-1 A NV~A Kgt A ^ I IkiIT e Vl \/
IJFmm3 Amenamen N~0&- a. 6 0,61,4 93 D as MuVrlata MY
^ _ _ J A
!NID-X DESIGN FEATURES SECT-ION P1AGEF-5.4-Site oration...- ............................................................................................ 5-5.2 Readtor CoFe. ............................................................ ;. . . .. . . . . .. . . . .
5.2.1 Fuel Ascrer mil er ............................................................. 5 1 5.2.2 Control Eleiment ASsem blies ................................................................ 5 1 Sh Fel.Strage...... ......................................................................................... -64 5 i 5.3.1 Spent Fuel -ZH )rage Rack Criticality {..................................................... 5 2 5.3.2 New Fuel S4Gorage Rack Criticality-. ....................................................... 5 2 2
5.3.3 Drainage-... ............................................................................................. 5 2 5.3. apaGey-..h G .............................................................................................
ARKANS5AASZ- IJM!Tq XV Amendelnntt 12No. 60181,205l
INDEX ADMINISTRATIVE CONTROLS SECT-ION PAGE 6 RESPONSiB TY ................................................... 1 6-.2-ORGANIZAT40N Offcito .................................................... 61 Facility Staff ................................................... 64 6.3 UNIT STAFF QUALIFICATIONS ........................................................ 6 5 6.1 TR AININ G ......................................................... 6 5 6.5 DELETED A LIMA K1O A o l @K2@T ^
IFkKWkFYWk)
. .. .... ... ... .. _ . .. . _ UFYI I i
_ ... ... . _ Av i Amendment No . 6O0a6
WQEX ADMINISTRATIVE CONTROLS SE-GTION PAGE-6 EPORT-ABE -VENT- ATION ................................................................
61-2 6.7 SAFETY LIMIT VIOLATION ....... ................................................................ 643 6.8 PROCEDURES AND PROGRAMA-&................................................ :...... - - -
6.9 REPORTING -REQUIREMENTS
^, t^ A Dnel i-ilkor m=DneDrpr V.ty.Ii F% U ;IHI M Erk-l a ......................... ................
........................................................... 6 1a 6.9.2 SPECIAL-RE-PORTS ...................................................... ..............................
6.9.3 RADIOACTIVE EFFLUENT RELEASE REPORT. .............................. f)48 6.9.4 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERA TIN REORT..
6.9.5- RE--ORE RAT4NG-4M R-PORT......................... .............................. 6-24 6 R E-G O -E -E .ION ..................................................... ..............................
642 6.11 RADIATION PROTECTION PROGRAM......................... .............................
6.12 Deleted.............................................................................. ..............................
6.13 HIG RADIATION AREA. .............................. 6-24 6.11 OFFSITE DOSE CALCULATION MANUAL (ODCM4g,) ........ ..............................
rr KTK FA IT l4Ve DATF TQTlktr- - -
DDnnDAR^A
- - - - ............ I......... .........
H - Fi
..... _ A,....
Amendlment No. 4,60,4,94, 5
_,., ^ _ ^
A 'ii
^ A AKAhrdN~iA~ UNI! 2
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 ROGS and DIT Computer Codes for Nuclear Design",CENPD 266 P-A, Ap 4l -Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, u(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-10965-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-01 90-A 7 January 1976 (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-AT May-1-9B (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-1 32-P7 August 4974 (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).
Supplement 1, February 1975 (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for-Lhiear HeatRate,-32.3foeAzimthaW[owef-tW"and 3.2.7 for ASi .
- 6) "Gal-GulafionalMethods for the CE Large Break-LOGA-E-valuation Model," CENPD 132 P, Supplent-2 uly §76-(methodology-for-Speoifieation 3.1.1.4 for MTC, 32.1 o-LineaIrHeat-Rate, 3.2.3 for AzimuthaPower-TAt-and .2. foS)
- 7) "Calculative Methods for the CE Large Brca L-CA Evaluation Model for the Analycis of CE and WDesigned NSSS," CEN 132, Supplement 3 P A, June 1985 (Methodale e for Specfiation 3.1.1.1 for MTC, 3.2.1 for Linear Heat Rate, 3.2.3 Aimuthl Power Tilt, and 3.2.7 for AS!).
- 85) "Calculative Methods for the CE Small Break LOCA Evaluation Model," CENPD-1 37-PT August 1974 (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)- leulative Methods for the CE Small Break LOCA Evaluation Model," cENPD 137, Supplement I P, January' 1977 (Methodology for Speification 3.1.1.4 forAMTG.,
3.2.1 for Linear Heat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7-for AS.
ARKANSAS - UNIT 2 6-21 Amendment No. 147,464,469,47-Q,
ADMINISTRATIVE CONTROL CORE OPERATING LIMITS REPORT
- 10) "Calculative Methods for the CE Small Break LOCA Evaluation Model," CENPD 137, Supplement 2 P A, dated April, 1998 (Methodology for Specification 3.1.1.4 for MTC, 3.2.1 for Lincar Heat Rate, 3.2.3 for Azimuthal Power Tilt, and 3.2.7 for ASI).
446) "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System-" Deremb _984-(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).
427) "Technical Manual for the CENTS Code," CENPD 282-P-A- F-ebuaiY4 94 (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.
- 13) Lotter: G.D. Parr (NRC) to F.M. SteFrn (CE), dated june 13, 1975 (NRC Staff Review o the-G ustien-E-ngineering ECCS EsvatuatinMdt).-NR4ieproaI fr 69514, 6 4-a5an6.9.5..8-4chodologies d
44)-Letter: O.D. P aFt(NRG)-to-AE. Scherer (CE), dated December 9,1975 (NRC-Staff Reviesw4hePp0sedGombusonEngineering- SEuatien-ModeI-ehanges}-
NRC approval for 6.9.5.1.6 mnethodology
- 15) Letter: K. Kniel (NRC) to A.E. Scherer (CE), dated September 27,1977 (Evaluation of Topieal-Reports CENPD 133, Supplement 3 P and-GENPD 137, Supplement I P).
NRG-approval for 6.9.5.1.0 methodology.
- 16) Letter: 2CNA038403, dated March 20,1984, J.R. Miller (NRC) to J.M. Griffin (APSL),
"CESEC Code Vr~ification." INICR approVal for 6.9.5.1.11 methodology.
CENPD 132 P, Supplement 4 P A, Revision 1 (Methodology for Speitft 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) "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-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).
- 10) "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs," WCAP-1 6045-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).
ARKANSAS - UNIT 2 6-21a Amendment No. 4657,464,469,470,482, 497,244,
6.9.5.2 The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
6.9.5.3 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
ARKANSAS - UNIT 2 6-21a Amendment No. 4&-,464,46-,4-79,482, 49,244,
Attachment 3 2CAN070402 List of Regulatory Commitments to 2CAN070402 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.
SCHEDULED TYPE COMPLETION COMMITMENT (Check one) 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). If the mFDI and measured oxide thickness correlate as expected or is conservative relative to predictions, mFDI will no longer be restricted except as required to meet the 100 micron oxide limit.
Cycle specific evaluations will be used to verify that x required power margins in the axial cutback regions are maintained within the safety analysis limitations. l Hot full power (HFP) MTC test within seven days of x I reaching the highest RCS soluble boron concentration predicted during full power operation will be performed. _
Provide the necessary information from the HFP test x I at peak boron concentration conditions to Westinghouse.
Plant procedures will be modified as needed to x I reflect the calculated peak HFP MTC along with ZrB2 IFBAs distinctive trend in RCS critical boron concentration.
In the event of a Condition Ill or IV event at ANO-2, x an evaluation of fuel structural integrity with respect to radial hydriding will be performed prior to power ascension.
Analyses, as part of the ANO-2 reload efforts, will be x performed in support of the generic implementation of ZrB2 fuel to ensure that cladding bursts are precluded for Conditions I, II, IlIl and IV events.
Prior to the use of ZrB2 burnable absorber coatings, x I the fuel design will be analyzed with applicable NRC staff approved codes and methods.
Enclosure I 2CAN070402 Supplemental Information Describing the use of the Westinghouse Nuclear Physics Code Package for Arkansas Nuclear One, Unit 2 to 2CAN070402 Page 1 of 38 Supplemental Information Describing the use of the Westinghouse Nuclear Physics Code Package for Arkansas Nuclear One, Unit 2 April 2004 ABSTRACT This document provides benchmarking data for Arkansas Nuclear One, Unit 2 (ANO-2).
The computer programs used are part of the Westinghouse Nuclear Physics code package (ANC/PHOENIX) and were obtained from Westinghouse Electric Company LLC. The calculations were performed by Westinghouse Electric Company LLC. The results of these calculations were compared to operating data from ANO-2. These comparisons further demonstrate the applicability of the Westinghouse Nuclear Physics code package to perform reload design calculations for ANO-2.
to 2CAN070402 Page 2 of 38 TABLE OF CONTENTS SECTION PAGE
- 1. INTRODUCTION AND CONCLUSIONS ................................... 6 1.1 OBJECTIVE ................................... 6 1.2 SCOPE ................................... 6
1.3 CONCLUSION
S ................................... 6
- 2. PHYSICS MODEL VERIFICATION FOR ANO-2 ................................... 7 2.1 CYCLE DESCRIPTION ....................... 7; 7
2.2 ZERO POWER PHYSICS TESTS ........................ 7 2.2.1 CRITICAL BORON CONCENTRATION .7 2.2.2 MODERATOR TEMPERATURE COEFFICIENT .8 2.2.3 CONTROL ROD WORTH .8 2.2.4 DIFFERENTIAL BORON WORTH .8 2.3 POWER OPERATION .8 2.3.1 BORON LETDOWN CURVES .8 2.3.2 AXIAL POWER DISTRIBUTIONS .8 2.4
SUMMARY
.8 to 2CAN070402 Page 3 of 38 LIST OF TABLES TABLE PAGE 2.2-1 Arkansas Nuclear One Unit 2 Cycles 15 and 16 HZP Critical Boron Concentration Comparison Between Measurement and Prediction .................. 10 2.2-2 Arkansas Nuclear One Unit 2 Cycles 15 and 16 HZP Moderator Temperature Coefficient Comparison Between Measurement and Prediction ..... 11 2.2-3 Arkansas Nuclear One Unit 2 Cycles 15 and 16 Control Rod Worth Comparison Between Measurement and Prediction .................................... 12 2.2-4 Arkansas Nuclear One Unit 2 Cycles 15 and 16 HZP Differential Boron Worth Comparison Between Measurement and Prediction .............................. 13 2.3-1 Arkansas Nuclear One Unit 2 Cycles 15 and 16 Boron Letdown Comparison Between Measurement and Prediction .................................... 14 to 2CAN070402 Page 4 of 38 LIST OF FIGURES FIGURE PAGE 2.1-1 Arkansas Nuclear One Unit 2 Cycle 15 Core Loading Pattern .................... 15 2.1-2 Arkansas Nuclear One Unit 2 Cycle 16 Core Loading Pattern .................... 16 2.2-1 Arkansas Nuclear One Unit 2 Cycle 15 Measured versus Predicted Reference Bank Integral Rod Worth ..................................................... 17 2.2-2 Arkansas Nuclear One Unit 2 Cycle 16 Measured versus Predicted Reference Bank Integral Rod Worth ..................................................... 18 2.3-1 Arkansas Nuclear One Unit 2 Cycle 15 Boron Letdown Comparison Between Measurement and Prediction ..................................................... 19 2.3-2 Arkansas Nuclear One Unit 2 Cycle 16 Boron Letdown Comparison Between Measurement and Prediction ..................................................... 20 2.3-3 Arkansas Nuclear One Unit 2 Cycle 15 Axial Power Distribution Comparison Between Plant measurement and ANC - 25.4 EFPD ............. 21 2.3-4 Arkansas Nuclear One Unit 2 Cycle 15 Axial Power Distribution Comparison Between Plant measurement and ANC - 246.9 EFPD ........... 22 2.3-5 Arkansas Nuclear One Unit 2 Cycle 15 Axial Power Distribution Comparison Between Plant measurement and ANC - 449.1 EFPD ........... 23 2.3-6 Arkansas Nuclear One Unit 2 Cycle 16 Axial Power Distribution Comparison Between Plant measurement and ANC - 27.0 EFPD ............. 24 2.3-7 Arkansas Nuclear One Unit 2 Cycle 16 Axial Power Distribution Comparison Between Plant measurement and ANC - 251.5 EFPD ........... 25 2.3-8 Arkansas Nuclear One Unit 2 Cycle 16 Axial Power Distribution Comparison Between plant measurement and ANC - 445.0 EFPD ........... 26 2.3-9 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ............ 27 ANC - 25.4 EFPD 2.3-10 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ............ 28 ANC - 246.9 EFPD 2.3-11 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assemble Average Power Distribution Comparison Between Plant measurement and ............ 29 ANC -449.1 EFPD 2.3-12 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ............ 30 ANC - 27.0 EFPD 2.3-13 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 251.5 EFPD ..................................................... 31 to 2CAN070402 Page 5 of 38 LIST OF FIGURES (continued)
FIGURE PAGE 2.3-14 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Average Power Distribution Comparison Between Plant measure and ANC - 445.0 EFPD ........................................................ 32 2.3-15 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 25.4 EFPD ............. 33 2.3-16 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 246.9 EFPD ........... 34 2.3-17 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 449.1 EFPD ........... 35 2.3-18 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 27.0 EFPD ............. 36 2.3-19 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 251.5 EFPD ........... 37 2.3-20 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 445.0 EFPD ........... 38 to 2CAN070402 Page 6 of 38
1.0 INTRODUCTION
AND CONCLUSIONS This report provides comparisons between predictions and operating data as a further demonstration of the applicability of the Westinghouse Nuclear Physics code package to perform reload design calculations for Arkansas Nuclear One, Unit 2 (ANO-2).
1.1 OBJECTIVE The objective of this report is to further demonstrate the applicability of the Westinghouse Nuclear Physics code package to perform reload design calculations for ANO-2. To this end, extensive design calculations have been performed for ANO-2 Cycles 15 and 16, and the results are compared to actual plant operating data.
1.2 SCOPE Westinghouse has performed core design calculations for comparison with core follow data for ANO-2. Comparisons between measurements and predictions for ANO-2 Cycles 15 and 16 are presented in Section 2. All methods used to generate the results detailed in this report (computer programs, model development, and data processing) are standard licensed methods used by Westinghouse Electric Company LLC.
1.3 CONCLUSION
S This report describes the use of the Westinghouse Nuclear Physics code package as applied to model the ANO-2 core. Calculations were performed for ANO-2 Cycles 15 and 16 as described in Section 2. The results from these comparisons further demonstrate the applicability of the Westinghouse Nuclear Physics code package to perform reload design calculations for ANO-2.
to 2CAN070402 Page 7 of 38
- 2. PHYSICS MODEL VERIFICATION FOR ANO-2 Core physics model verification for ANO-2 will include comparisons between measurement and predictions for Cycles 15 and 16. ANO-2 is currently in its 17t cycle of operation. In this section, predictions made using the Westinghouse Nuclear Physics code package are compared to zero power physics test measurements and at power operating data. As stated in Section 1,the methods employed to generate the predictions reported in this section are standard licensed methods used by Westinghouse Electric Company LLC.
ANO-2 is a Combustion Engineering (CE) reactor with a thermal rating of 3026 megawatts. The core consists of 177 assemblies of the CE 16 x 16 design.
2.1 CYCLE DESCRIPTIONS ANO-2 Cycle 15 began operation in December 2000 and shutdown in April 2002 after a 471 Effective Full Power Days (EFPD) cycle. Cycle 15 used Guardian Grid fuel assemblies with an active fuel length of 150 inches. Gadolinia was used as the burnable absorbers. The core loading pattern for Cycle 15 including a description of the fresh fuel and the locations of control rods is shown in Figure 2.1-1.
ANO-2 Cycle 16 began operation in May 2002 and shutdown in September 2003 after a 493 Effective Full Power Days (EFPD) cycle. Cycle 16 used Guardian Grid fuel assemblies with an active fuel length of 150 inches. Erbia was used as the burnable absorbers. The core loading pattern for Cycle 16 including a description of the fresh fuel and the locations of control rods is shown in Figure 2.1-2.
2.2 ZERO POWER PHYSICS TESTS After each refueling, startup physics tests are conducted to verify that the nuclear characteristics of the core are consistent with design predictions. While the reactor is maintained at hot zero power (HZP) conditions, the following physics parameters are measured:
- Critical Boron Concentrations,
- Moderator Temperature Coefficient,
- Control Rod Worth, and
- Differential boron worth.
2.2.1 CRITICAL BORON CONCENTRATION Table 2.2-1 provides the comparisons between HZP critical boron concentrations measurements and predictions for Cycles 15 and 16. The values represent all rods out (ARO) and Reference Bank in conditions. As shown, excellent agreement is demonstrated for each case with all differences well within the +/- 100 ppm review criteria.
to 2CAN070402 Page 8 of 38 2.2.2 MODERATOR TEMPERATURE COEFFICIENT Table 2.2-2 provides the comparisons between HZP Moderator Temperature Coefficient measurements and predictions for Cycles 15 and 16. Excellent agreement is demonstrated with all differences being well within the review criteria of +/- 3 pcm/0F.
2.2.3 CONTROL ROD WORTH Table 2.2-3 provides the Control Rod Worth comparisons between measurement and prediction for Cycles 15 and 16. In all cases, the agreement is within acceptance criteria. Figures 2.2-1 and 2.2-2 show the integral rod worth comparisons for the Reference Bank. The predicted rod worth and integral worth were calculated at the exact conditions which were present during the measurement. Excellent agreement is observed between measured and predicted integral worth.
2.2.4 DIFFERENTIAL BORON WORTH Table 2.2-4 provides the Differential Boron Worth comparisons between measurement and predictions for Cycles 15 and 16. Both the measured and predicted values are obtained using the worth of the Reference Bank in pcm divided by the change in boron concentration from ARO to Reference Bank inserted. All differences are within the review criteria of +/- 5%.
2.3 POWER OPERATION 2.3.1 BORON LETDOWN CURVES Reactor coolant system boron concentrations are measured daily at the plant. Critical boron concentrations measured at or very close to hot full power, all rods out, equilibrium xenon and samarium conditions are compared to the predicted boron letdown curves for Cycles 15 and 16 in Figures 2.3-1 and 2.3-2. The predicted curves were obtained from design depletions with the three-dimensional ANC model. Table 2.3-1 shows the difference in ppm between measurement and ANC at various cycle exposures. The mean difference between measured and predicted critical boron concentration for both cycles is 21 ppm with a standard deviation of 6 ppm.
2.3.2 AXIAL POWER DISTRIBUTIONS Measured core average axial power distributions for Beginning-of-Cycle (BOC), Middle-of-Cycle (MOC) and End-of-Cycle (EOC) obtained with the incore monitoring codes using incore detector "snapshots" were compared to predicted axial distributions in Figures 2.3-3 through 2.3-8. The predicted distributions were obtained from three-dimensional ANC calculations performed for core conditions similar to those at the time of the usnapshots.' Overall, the comparisons show excellent agreement between measured and predicted axial power distributions.
to 2CAN070402 Page 9 of 38 2.4
SUMMARY
In this section, predictions made using the Westinghouse Nuclear Physics code package are compared to zero power physics test measurements and at power operating data from ANO-2, Cycles 15 and 16. In all cases, the predictions agree well with the measurements and produce results that are essentially the same as the current ROCS and DIT system. The agreement between the predictions and the measurements reported here further demonstrates the applicability of the Westinghouse Nuclear Physics code package to perform reload design calculations for ANO-2.
Enclosure I to 2CAN070402 Page 10 of 38 TABLE 2.2-1 Arkansas Nuclear One Unit 2 Cycles 15 and 16 HZP Critical Boron Concentration Comparison Between Measurement and Prediction I CYCLE l BORON CONCENTRATION (PPM)
CONFIGURATION MEASURED PREDICTED DIFF.
(M) (P) (M-P) 15 ARO 1907 1930 -23 BANK B IN 1664 1681 -15 16 ARO 1937 1959 -22 BANK B IN 1665 1685 -20 Review Criteria is +/- 100 ppm Enclosure I to 2CAN070402 Page 11 of 38 TABLE 2.2-2 Arkansas Nuclear One Unit 2 Cycles 15 and 16 HZP Moderator Temperature Coefficient Comparison Between Measurement and Prediction CYCLE MODERATOR TEMPERATURE COEFFICIENT (PCM/ 0 F)
CONFIGURATION MEASURED PREDICTED DIFF.
(M) (P) (M-P) 15 ARO +1.77 +2.98 -1.21 16 ARO +0.70 +1.53 -0.83 Review Criteria is i 3 pcm/ 0F to 2CAN070402 Page 12 of 38 TABLE 2.2-3 Arkansas Nuclear One Unit 2 Cycles 15 and 16 Control Rod Worth Comparison Between Measurement and Prediction CYCLE CONTROL ROD WORTH (PCM)
CEA MEASURED PREDICTED % DIFFER 1ENCE GROUP (M) (P) (M/P- 1))*1 00 15 3 &6 1039 1052 -1.2 2 &P 1189 1182 0.6 A&5 1088 1092 -0.4 1&4 1211 1232 -1.7 1835 1846 -0.6 ITOTAL(2) 6362 6404 -0.7 16 3 &4 968 978 -1.0 P& 5 1360 1322 2.9 1& 6 1339 1322 1.3 A&2 1360 1344 1.2 2119 2029 4.4 7146 6995 2.2 Acceptance Criteria is +/-15% or 100 pcm which ever is greater
(') Reference Bank - Acceptance Criteria is +/-10%
(2) Sum of all measured banks within +/-10%
to 2CAN070402 Page 13 of 38 TABLE 2.2-4 Arkansas Nuclear One Unit 2 Cycles 15 and 16 HZP Differential Boron Worth Comparison Between Measurement and Prediction CYCLE S DIFFERENTIAL BORON WORTH (PCM/PPM)
CONFIGURATION MEASURED PREDICTED % DIFFERENCE (M) (P) (MP-1) *100 15 AVERAGE OVER BANK B 7.53 7.48 0.7 INSERTION 16 AVERAGE OVER BANK B 7.82 7.48 4.5 INSERTION Review Criteria is +/- 5%.
to 2CAN070402 Page 14 of 38 TABLE 2.3-1 Arkansas Nuclear One Unit 2 Cycles 15 and 16 Boron Letdown Comparison Between Measurement and Prediction I CYCLE JHFP BORON LETDOWN (PPM)
EXPOSURE MEASURED PREDICTED DIFFERENCE l l (MWDIMTU) (M) (P) im-15 50 1224 1240 -16 75 1170 1183 -13 100 1112 1127 -15 125 1060 1076 -17
______150 1009 1029 -20 175 962 982 -20
______200 915 935 -20 225 851 881 -30 250 787 813 -26
______275 710 742 -32 300 627 651 -25 325 538 561 -23 350 448 471 -23 375 356 381 -25
______400 267 291 -24 425 178 202 -24 434 146 169 -24 16 50 1247 1269 -22 75 1211 1219 -7 100 1149 1168 -19
______125 1084 1100 -16 150 1013 1032 -19
______175 940 956 -16 200 860 881 -20 225 778 801 -23 250 696 721 -25 275 614 638 -24 300 527 554 -27 325 446 470 -24 350 363 385 -22
_______375 278 300 -22 400 194 215 -21 425 112 130 -18 452 25 29 -4 Enclosure I to 2CAN070402 Page 15 of 38 FIGURE 2.1-1 Arkansas Nuclear One Unit 2 Cycle 15 Core Loading Pattern H J K L M N P R M2 Si T4 S4 S4 T3 T4 54 6 2 p 5 Si S2 S2 'T4 S4 S4 TI R4 B S4I B T3 A T4 S2 S4 T3 $4 T4 TI R2 2 _ _ 4 _ _ B _ _ 3_ _
S4 T4 T3 S4 S3 T2 Si B p A S4 S4 S4 S3 Si TI R4 T3 S4 T4 T2 TI RI 4
I + 4 A I 4 2
T4 TI T1 Si R4 5 3 S4 R4 R2 A Legend Batch ID jCEAJ to 2CAN070402 Page 16 of 38 FIGURE 2.1-2 Arkansas Nuclear One Unit 2 Cycle 16 Core Loading Pattern H J K L M N P R 8 M2 TI T4 U4 T1 T1 U2 T4 6 2 P 5 9 T1 TI U4 T4 U4 T2 U2 T4 B B I A 10 T4 U4 T3 .U3 TI V3 VI S2 2 1 4 B I 3 11 U4 T4 U3 T3 U3 U2 Si B P A 12 TI U4 TI U3 T3 U1 Si
_ _ _ B 16 1 13 TI T2 U3 U2 U1 S4 I A 2 I t I-14 U2 U2 U1 Si Si 5 3 15 T4 T4 S2 A Legend Batch ID CEA Group ID to 2CAN070402 Page 17 of 38 FIGURE 2.2-1 Arkansas Nuclear One Unit 2 Cycle 15 Measured versus Predicted Reference Bank Integral Rod Worth 2000- .
1800 1600 -
1400 \
1200 0
AA-800 JiEASURED (ID A [
l_-REDICTED (P)
A AN A
400 -
200 = -
0 0 40 80 120 160 RODPOSITIN (INCHE3 WKhtRAWN) to 2CAN070402 Page 18 of 38 FIGURE 2.2-2 Arkansas Nuclear One Unit 2 Cycle 16 Measured versus Predicted Reference Bank Integral Rod Worth 2200X A
2000 -1_ _ _ _
A 1800 1800- - _ _ _ _
1400 1200 - ~
I 1000 I _ _ _ _ _ _
.1I -MEASURED (M) oo0 0---_-PREDICTED (P)_
600 -
400 \
oo 20 1 1X 0 40 s0 120 180 RODPOSMON (IN04ES WTHDRAWNM to 2CAN070402 Page 19 of 38 FIGURE 2.3-1 2.3-1 Arkansas Nuclear One Unit 2 Cycle 15 Boron Letdown Comparison Between Measurement and Prediction
- - - - -ANC Predicted
- Measured (B.10 Corrected for 19.80a~b) I 1800 r I 11111[ 1111 1
600 - ---.-- =
500- -
- 100 1: 1
-81 C
300 soo l lU =
. T TT1'11111 1 2
200 - - - -= . -
8 800 - i 400 l ll l l 1 '
300 - .- - .- -- - - -
0.0 50.0 100.0 1500 200.0 250.0 300.0 350.0 400.0 450.0 500.0 EFPD to 2CAN070402 Page 20 of 38 FIGURE 2.3-2 Arkansas Nuclear One Unit 2 Cycle 16 Boron Letdown Comparison Between Measurement and Prediction I- - - ANCPredided
- Measured (B-10 Corrected for 19.80 aAD)I 1800 - I [ 1 1700 - - 1 1500 1400 1300
- "S 1200
. .AA I 1000 . .
C 900
-2 800 -
500- ___K - l l ,L ]_
500 -- - '
400 - - I. I 200.
o 0.0 50.0
-=100.0 150.0 S_
200.0 250.0 300.0 350.0 400.0. 450.0 500.0 EFPD to 2CAN070402 Page 21 of 38 FIGURE 2.3-3 Arkansas Nuclear One Unit 2 Cycle 15 Axial Power Distribution Comparison Between Plant measurement and ANC - 25.4 EFPD
-Predicted A Measured osEE 1.0 ln.Il __ __EL 0.9 - _A-1 KJ 0.8 - /_ l L 9 . _ __
I.
0.7 +/- +/- __
0 a.
6 0.6 -1Lz H_ _z W0 0.4 -_Lz zH_ _H 0.3 ___-z-- l z__
0.2 z 11zz__
0.1 I L H H 0.0 0.0
=
10.0 20.0
+
30.0 zzH 40.0 50.0 60.0 z_
70.0 80.0 90.0 5 100.0 Core Helght (%)
to 2CAN070402 Page 22 of 38 FIGURE 2.3-4 Arkansas Nuclear One Unit 2 Cycle 15 Axial Power Distribution Comparison Between Plant measurement and ANC - 246.9 EFPD l-Predicted a Measuredl 1.2 _ __ __
1.1 __ I ___f i 1.0 0.8 -L L _
0.7 08 7 1 1 1 1__11 I__iTI_
___II \
0o - i7 _ _
1 4 __ _
0U - I -I-0.4 _____
04--_ l 2 _ _
- 0. _ T1 T__
0.40 _ 1 W __
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Core Height (%)
Enclosure I to 2CAN070402 Page 23 of 38 FIGURE 2.3-5 Arkansas Nuclear One Unit 2 Cycle 15 Axial Power Distribution Comparison Between Plant measurement and ANC - 449.1 EFPD
-- Predicted A Measured 1.2
- A A llg!
A A A a Al AI
. t-_ - _ I I_ _ _
0 _8 a/ 7 I I I _ \
0 0.7 -
a.
0.6 ==
0 0.4 - I i l _
0.2 . _-.
__= I __ =__ -.
-- 1-
__ 1 = =__
00 = = _ I 1 =
___ = =
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Core Height (%)
to 2CAN070402 Page 24 of 38 FIGURE 2.3-6 Arkansas Nuclear One Unit 2 Cycle 16 Axial Power Distribution Comparison Between Plant measurement and ANC - 27.0 EFPD l-Prrdicted Measuredl 1.3r I II I I I II I 1.2 - _ J Ii . W 1.0 _ +/- L +/- i 0.9 f TL-i L-0.8 lL +/-H__
0 a.
z 01 0.6 -=08i^ i K Z+/-i1 _ _ T 122 M
0.4 - -= _ +/-1 0.3 _
0.2 _ - J L K 0_1 _ _ z +/-= _ _ +/-=
v.v 0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Core Height (%)
Enclosure I to 2CAN070402 Page 25 of 38 FIGURE 2.3-7 Arkansas Nuclear One Unit 2 Cycle 16 Axial Power Distribution Comparison Between Plant measurement and ANC - 251.5 EFPD I-Predicted & Measured 1.21 III I II I 1I I I
, i I A 1tIII 4 4 *
- a. ..................... .
1.0 t.O H+
~~l Li II>
0.9 A_ u H z H z H _ _ _
0 .8-_ zu ti +/- _ _
310 0.7 --
0 0-0.1 0.6 _
0.3 -=_r
- 0. 2 ==2l l 0.2 i_ 1 _z -
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Core Height (%)
to 2CAN070402 Page 26 of 38 FIGURE 2.3-8 2.3-8 Arkansas Nuclear One Unit 2 Cycle 16 Axial Power Distribution Comparison Between Plant measurement and ANC - 445.0 EFPD I-Predicted A Measured 1.2 -
1.1 TI 1L. I 1.1 l,& A 'Aa I IlIlIla d
AA.
0.8 . i _ ___
0 0 .7_ _ _ _ _ _ _ - - - - _ _ _ _ _ _ _ _ _ -
0.06 ______T____ l I1I 1 I _____ ____ I____
- 0 0.5 ==
0.4 I T 1 I 7
._ _ _ __ _ _ I - __ _ I __ _ _ _ _ _
0.3- - __ = = -
0.2 1 _ I - I I 0.1 r _ I I I 0.0=
0.0 10.0 20.0 30.0 40.0 50.0 60.0 70.0 80.0 90.0 100.0 Core Height (%)
Enclosure I to 2CAN070402 Page 27 of 38 FIGURE 2.3-9 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 25.4 EFPD Legend: ANC Predicted Assembly Average Power Measured Data
% Difference 0.952 1.246 1.336 1.125 1.064 1.262 1.068 0.457 0.921 1.235 1.335 1.125 1.059 1.267 1.070 0.452 3.388 0.858 0.052 0.018 0.510 -0.355 -0.224 1.173 1.246 1.210 1.221 1.309 1.080 1.090 1.146 0.376 1.250 1.215 1.226 1.309 1.075 1.081 1.166 0.373
-0.320 -0.436 -0.408 -0.008 0.484 0.795 -1.707 0.858 1.336 1.222 1.179 1.363 1.111 1.243 1.061 0.286 1.330 1.221 1.174 1.366 1.101 1.222 1.064 0.289 0.466 0.123 0.434 -0.190 0.881 1.743 -0.291 -0.867 1.125 1.309 1.363 1.136 1.125 1.250 0.696 1.110 1.303 1.364 1.134 1.129 1.260 0.702 1.333 0.491 -0.095 0.168 -0.310 -0.794 -0.841 1.064 1.080 1.111 1.124 1.131 1.013 0.335 1.052 1.073 1.105 1.119 1.145 1.047 0.341 1.160 0.652 0.570 0.456 -1.249 -3.257 -1.644 1.262 1.090 1.242 1.249 1.013 0.394 1.271 1.095 1.250 1.263 1.030 0.397
-0.708 -0.457 -0.624 -1.124 -1.660 -0.831 1.068 1.146 1.059 0.693 0.335 1.081 1.192 1.081 0.711 0.338
-1.193 -3.875 -2.044 -2.573 -0.917 0.457 0.376 0.286 0.455 0.375 0.279 0.484 0.401 2.436 to 2CAN070402 Page 28 of 38 FIGURE 2.3-10 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 246.9 EFPD Legend: ANC Predicted Assembly Average Power Measured Data
% Difference 0.806 1.077 1.369 1.121 1.052 1.368 1.236 0.529 0.781 1.067 1.365 1.113 1.054 1.380 1.244 0.531 3.188 0.937 0.300 0.692 -0.228 -0.898 -0.659 -0.283 1.077 1.063 1.148 1.403 1.071 1.099 1.209 0.420 1.073 1.057 1.145 1.407 1.073 1.102 1.224 0.425 0.335 0.539 0.244 -0.263 -0.168 -0.245 -1.258 -1.060 1.369 1.148 1.114 1.396 1.082 1.342 1.098 0.311 1.353 1.139 1.109 1.404 1.084 1.349 1.111 0.320 1.190 0.755 0.469 -0.591 -0.157 -0.489 -1.206 -2.843 1.121 1.403 1.395 1.059 1.036 1.235 0.701 1.101 1.392 1.391 1.057 1.043 1.252 0.718 1.835 0.826 0.266 0.189 -0.671 -1.334 -2.327 1.052 1.070 1.082 1.036 1.005 0.956 0.336 1.036 1.054 1.070 1.024 1.019 0.981 0.349 1.574 1.489 1.112 1.152 -1.355 -2.558 -3.697 1.368 1.099 1.341 1.235 0.956 0.376 1.365 1.094 1.342 1.241 0.971 0.387 0.249 0.494 -0.045 -0.475 -1.545 -2.893 1.236 1.209 1.097 0.699 0.336 1.237 1.226 1.107 0.718 0.348
-0.057 -1.395 -0.903 -2.578 -3.448
+ I-0.529 0.419 0.311 0.526 0.420 0.311 0.551 -0.191 -0.064 Enclosure I to 2CAN070402 Page 29 of 38 FIGURE 2.3-11 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 449.1 EFPD Legend: ANC Predicted Assembly Average Power Measured Data
% Difference 0.836 1.068 1.301 1.078 1.037 1.328 1.250 0.594 0.804 1.058 1.295 1.072 1.026 1.332 1.259 0.601 4.045 0.964 0.432 0.569 1.092 -0.308 -0.746 -1.132 1.068 1.049 1.104 1.322 1.053 1.085 1.208 0.472 1.055 1.035 1.094 1.322 1.044 1.080 1.222 0.485 1.232 1.333 0.923 0.038 0.852 0.491 -1.162 -2.660 1.301 1.104 1.070 1.322 1.067 1.313 1.114 0.352 1.285 1.094 1.057 1.323 1.059 1.324 1.133 0.372 1.269 0.896 1.211 -0.053 0.765 -0.801 -1.677 -5.427 1.078 1.322 1.322 1.048 1.050 1.242 0.748 1.054 1.312 1.314 1.038 1.048 1.258 0.776 2.287 0.739 0.647 0.993 0.229 -1.233 -3.571 1.037 1.053 1.067 1.051 1.046 1.029 0.388 1.015 1.038 1.052 1.040 1.055 1.051 0.413 2.198 1.484 1.416 1.087 -0.815 -2.075 -5.985 1.328 1.085 1.313 1.242 1.029 0.436 1.321 1.074 1.316 1.251 1.044 0.454 0.507 1.043 -0.190 -0.719 -1.390 -4.007 1.250 1.208 1.114 0.746 0.388 1.251 1.221 1.127 0.775 0.411
-0.088 -1.073 -1.127 -3.754 -5.527 0.594 0.471 0.352 0.594 0.477 0.360 0.084 -1.216 -2.304 to 2CAN070402 Page 30 of 38 FIGURE 2.3-12 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 27.0 EFPD Legend: ANC Predicted Assembly Average Power Measured Data
% Difference 0.938 1.183 1.177 1.293 1.232 1.240 1.185 0.528 0.875 1.155 1.167 1.303 1.242 1.248 1.187 0.515 7.261 2.442 0:840 -0.745 -0.781 -0.665 -0.126 2.624 1.183 1.225 1.286 1.165 1.297 1.162 1.131 0.483 1.168 1.212 1.287 1.166 1.310 1.161 1.125 0.467 1.302 1.089 -0.101 -0.069 -0.970 0.103 0.533 3.448 1.177 1.286 1.156 1.300 1.199 1.225 0.982 0.296 1.168 1.290 1.155 1.325 1.206 1.230 0.970 0.273 0.814 -0.271 0.104 -1.902 -0.547 -0.431 1.195 8.306 1.293 1.165 1.301 1.156 1.249 1.147 0.529 1.296 1.160 1.311 1.160 1.259 1.149 0.522
-0.239 0.466 -0.740 -0.319 -0.794 -0.174 1.302 1.232 1.297 1.200 1.250 0.990 0.887 0.291 1.235 1.301 1.197 1.256 0.983 0.890 0.292
-0.211 -0.323 0.251 -0.446 0.733 -0.371 -0.411 1.240 1.168 1.228 1.149 0.887 0.367 1.246 1.164 1.221 1.140 0.884 0.366
-0.458 0.370 0.557 0.789 0.339 0.383 1.185 1.135 0.984 0.530 0.295 1.186 1.133 0.969 0.509 0.294
-0.067 0.194 1.558 4.167 0.511 0.528 0.484 0.296 0.515 0.471 0.284 2.544 2.673 4.409 to 2CAN070402 Page 31 of 38 FIGURE 2.3-13 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 251.5 EFPD Legend: ANC Predicted Assembly Average Power Measured Data
% Difference 0.820 1.048 1.102 1.317 1.164 1.135 1.145 0.540 0.790 1.049 1.111 1.337 1.178 1.145 1.151 0.532 3.863 -0.124 -0.765 -1.466 -1.180 -0.847 -0.530 1.504 1.048 1.111 1.292 1.129 1.318 1.102 1.117 0.503 1.039 1.111 1.306 1.140 1.338 1.106 1.119 0.493 0.886 -0.027 -1.072 -0.948 -1.517 -0.317 -0.152 2.070 1.102 1.291 1.121 1.335 1.174 1.275 1.004 0.319 1.099 1.306 1.127 1.362 1.187 1.290 1.005 0.308 0.273 -1.118 -0.488 -1.961 -1.120 -1.140 -0.060 3.706 1.317 1.128 1.336 1.139 1.312 1.194 0.566 1.319 1.131 1.352 1.148 1.331 1.208 0.573
-0.152 -0.239 -1.154 -0.801 -1.450 -1.143 -1.135 1.164 j 1.318 1.174 1.313 1.005 0.938 0.326 1.160 1.325 1.178 1.325 1.006 0.952 0.337 0.353 -0.543 -0.348 -0.891 -0.099 -1.502 -3.379 1.135 1.104 1.276 1.195 0.938 0.402 1.130 1.099 1.278 1.198 0.945 0.413 0.460 0.437 -0.164 -0.217 -0.688 -2.640 I- I t I 1.145 1.119 1.004 0.566 0.330 1.139 1.118 0.999 0.561 0.339 0.491 0.098 0.501 0.873 -2.511 4- t -
0.540 0.503 0.319 0.526 0.493 0.314 2.603 2.132 1.722 to 2CAN070402 Page 32 of 38 FIGURE 2.3-14 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Average Power Distribution Comparison Between Plant measurement and ANC - 445.0 EFPD Legend: ANC Predicted Assembly Average Power Measured Data
% Difference 0.831 1.034 1.073 1.282 1.125 1.110 1.156 0.589 0.797 1.023 1.074 1.289 1.130 1.109 1.156 0.589 4.279 1.075 -0.075 -0.574 -0.434 0.099 0.035 0.034 1.034 1.086 1.265 1.091 1.285 1.084 1.136 0.553 1.019 1.081 1.269 1.098 1.294 1.082 1.135 0.551 1.442 0.425 -0.347 -0.619 -0.688 0.194 0.053 0.308 1.073 1.264 1.091 1.298 1.138 1.268 1.034 0.359 1.071 1.280 1.090 1.313 1.146 1.277 1.038 0.358 0.196 -1.242 0.083 -1.157 -0.655 -0.681 -0.395 0.251 1.282 1.091 1.298 1.109 1.294 1.198 0.610 1.284 1.098 1.307 1.111 1.303 1.208 0.628
-0.125 -0.610 -0.651 -0.189 -0.721 -0.820 -2.913 1.125 1.285 1.138 1.295 1.016 0.980 0.366 1.124 1.297 1.144 1.307 1.012 0.990 0.392 0.125 -0.910 -0.507 -0.888 0.375 -1.040 -6.633 1.110 1.086 1.268 1.198 0.980 0.447 1.099 1.078 1.269 1.203 0.992 0.473 1.001 0.733 -0.095 -0.449 -1.210 -5.417 1.034 1.156 0.609 0.370 1.137 1.143 1.126 1.030 0.613 0.394 1.129 1.022 0.379 -0.653 -6.044
+ 4 9 0.589 0.553 0.359 0.581 0.549 0.365 1.325 0.820_
-1.752_
to 2CAN070402 Page 33 of 38 FIGURE 2.3-15 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 25.4 EFPD Legend: ANC Predicted Assembly Fr Measured Data
% Difference 1.058 1.332 1.537 1.225 1.140 1.455 1.332 0.747 1.063 1.299 1.514 1.222 1.127 1.449 1.334 0.730
-0.452 2.509 1.519 0.237 1.127 0.442 -0.157 2.371 1.332 1.300 1.313 1.510 1.187 1.185 1.392 0.674 1.314 1.302 1.306 1.490 1.188 1.167 1.412 0.645 1.378 -0.131 0.574 1.335 -0.084 1.525 -1.423 4.529 1.537 1.317 1.265 1.577 1.216 1.426 1.383 0.605 1.508 1.305 1.248 1.566 1.206 1.382 1.385 0.618 1.910 0.935 1.346 0.702 0.829 3.176 -0.130 -2.088 1.225 1.510 1.576 1.251 1.232 1.446 1.131 1.207 1.483 1.565 1.259 1.221 1.449 1.116 1.458 1.855 0.709 -0.643 0.942 -0.200 1.362 1.140 1.186 1.216 1.232 1.206 1.364 0.726 1.117 1.187 1.210 1.210 1.216 1.403 0.690 2.068 -0.093 0.488 1.852 -0.790 -2.807 5.263 1.455 1.184 1.425 1.445 1.364 0.815 1.454 1.182 1.414 1.453 1.382 0.823 0.048 0.169 0.792 -0.551 -1.274 -0.984 4 4 .4 4- 1 1.332 1.391 1.382 1.123 0.726 1.348 1.443 1.407 1.125 0.684
-1.172 -3.610 -1.798 -0.204 6.156 0.747 0.673 0.605 0.734 0.652 0.600 1.840 3.237 0.917 to 2CAN070402 Page 34 of 38 FIGURE 2.3-16 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 246.9 EFPD Legend: ANC Predicted Assembly Fr Measured Data
% Difference 0.893 1.175 1.500 1.211 1.125 1.489 1.455 0.844 0.893 1.153 1.482 1.194 1.117 1.497 1.467 0.831
-0.022 1.873 1.187 1.449 0.725 -0.548 -0.838 1.516 1.175 1.155 1.227 1.543 1.180 1.182 1.408 0.703 1.158 1.145 1.226 1.532 1.175 1.182 1.429 0.694 1.494 0.917 0.073 0.698 0.426 -0.008 -1.497 1.355 1.500 1.227 1.205 1.526 1.179 1.470 1.372 0.625 1.470 1.217 1.186 1.523 1.172 1.467 1.401 0.646 2.076 0.805 1.636 0.223 0.632 0.211 -2.070 -3.281 1.211 1.543 1.526 1.178 1.136 1.398 1.116 1.180 1.517 1.509 1.169 1.142 1.406 1.131 2.610 1.721 1.127 0.761 -0.517 -0.583 -1.326 1.125 1.179 1.179 1.135 1.065 1.223 0.693 1.097 1.156 1.156 1.121 1.078 1.264 0.679 2.515 2.025 2.034 1.222 -1.206 -3.244 2.137 1.489 1.183 1.469 1.397 1.223 0.722 1.479 1.173 1.459 1.395 1.252 0.742 0.656 0.827 0.692 0.172 -2.308 -2.709 1.455 1.408 1.371 1.110 0.694 1.457 1.431 1.395 1.127 0.676
-0.165 -1.600 -1.735 -1.508 2.648 0.844 0.702 0.625 0.824 0.689 0.630 2.390 1.857 -0.762 to 2CAN070402 Page 35 of 38 FIGURE 2.3-17 Arkansas Nuclear One Unit 2 Cycle 15 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 449.1 EFPD Legend: ANC Predicted Assembly Fr Measured Data
% Difference 0.927 1.137 1.392 1.145 1.096 1.422 1.410 0.896 0.912 1.116 1.382 1.133 1.078 1.421 1.426 0.898 1.611 1.927 0.760 1.050 1.632 0.049 -1.108 -0.234 1.137 1.132 1.159 1.416 1.126 1.155 1.360 0.745 1.112 1.110 1.144 1.412 1.117 1.143 1.381 0.753 2.276 1.982 1.302 0.297 0.779 1.085 -1.535 -1.062 1.392 1.159 1.137 1.408 1.132 1.409 1.331 0.668 1.370 1.144 1.115 1.406 1.124 1.414 1.364 0.706 1.591 1.347 1.973 0.164 0.703 -0.382 -2.405 -5.409 1.145 1.416 1.408 1.124 1.128 1.364 1.125 1.115 1.402 1.397 1.115 1.122 1.374 1.147 2.737 1.013 0.823 0.825 0.544 -0.735 -1.935 1.096 1.125 1.134 1.129 1.094 1.244 0.754 1.067 1.111 1.117 1.115 1.106 1.280 0.756 2.728 1.278 1.495 1.301 -1.076 -2.805 -0.265 1.422 1.154 1.409 1.364 1.245 0.790 1.410 1.136 1.406 1.367 1.273 0.812 0.837 1.558 0.228 -0.241 -2.161 -2.673 4- 4 -.--. l. 4 1.410 1.360 1.331 1.119 0.755 1.417 1.380 1.356 1.140 0.753
-0.459 -1.428 -1.851 -1.851 0.252 0.896 0.744 0.669 0.887 0.745 0.687 0.981 -0.094 -2.677 Enclosure I to 2CAN070402 Page 36 of 38 FIGURE 2.3-18 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 27.0 EFPD Legend: ANC Predicted Assembly Fr Measured Data
% Difference 1.043 1.260 1.276 1.373 1.326 1.336 1.374 0.855 1.011 1.220 1.249 1.382 1.322 1.319 1.386 0.831 3.145 3.296 2.170 -0.651 0.303 1.327 -0.844 2.913 1.260 1.308 1.362 1.237 1.383 1.268 1.344 0.832 1.226 1.279 1.368 1.227 1.400 1.264 1.361 0.806 2.807 2.275 -0.409 0.807 -1.207 0.324 -1.249 3.200 1.276 1.362 1.225 1.369 1.275 1.350 1.293 0.614 1.249 1.369 1.216 1.400 1.274 1.361 1.282 0.579 2.137 -0.504 0.715 -2.207 0.047 -0.786 0.866 6.137 1.373 1.237 1.371 1.228 1.360 1.356 0.894 1.376 1.221 1.385 1.225 1.380 1.364 0.848
-0.196 1.286 -1.032 0.253 -1.449 -0.594 5.437 1.326 1.382 1.277 1.362 1.145 1.211 0.633 1.315 1.387 1.265 1.378 1.131 1.224 0.604 0.821 -0.389 0.917 -1.132 1.283 -1.062 4.819 1.336 1.272 1.352 1.359 1.212 0.733 1.317 1.268 1.350 1.352 1.217 0.708 1.450 0.355 0.126 0.503 -0.411 3.590
+ 4- 4 4-1.374 1.351 1.296 0.895 0.651 1.384 1.377 1.281 0.824 0.625
-0.723 -1.902 1.195 8.682 4.127 0.855 0.833 0.616 0.832 0.815 0.594 2.715 2.221 3.651 to 2CAN070402 Page 37 of 38 FIGURE 2.3-19 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 251.5 EFPD Legend: ANC Predicted Assembly Fr Measured Data
% Difference 0.906 1.102 1.208 1.388 1.226 1.201 1.284 0.829 0.904 1.099 1.211 1.397 1.234 1.204 1.299 0.814 0.199 0.236 -0.223 -0.666 -0.608 -0.282 -1.132 1.868 1.102 1.155 1.364 1.188 1.390 1.183 1.272 0.812 1.090 1.160 1.371 1.196 1.401 1.188 1.288 0.800 1.129 -0.388 -0.511 -0.627 -0.792 -0.379 -1.258 1.513 1.208 1.365 1.182 1.401 1.246 1.380 1.286 0.625 1.199 1.371 1.186 1.419 1.252 1.399 1.283 0.611 0.767 -0.452 -0.304 -1.268 -0.503 -1.372 0.257 2.375 1.388 1.188 1.402 1.193 1.391 1.375 0.903 1.379 1.186 1.409 1.201 1.411 1.401 0.881 0.689 0.143 -0.511 -0.625 -1.445 -1.849 2.462 1.226 1.390 1.248 1.391 1.139 1.238 0.669 1.215 1.387 1.243 1.405 1.140 1.280 0.660 0.914 0.209 0.394 -0.961 -0.079 -3.311 1.440 1.201 1.185 1.381 1.376 1.239 0.751 1.189 1.179 1.387 1.389 1.272 0.749 1.026 0.526 -0.418 -0.943 -2.602 0.254 1.284 1.277 1.287 0.904 0.682 1.286 1.290 1.277 0.862 0.679
-0.132 -1.015 0.823 4.824 0.383 0.829 0.813 0.625 0.807 0.800 0.617 2.777 1.574 1.379 to 2CAN070402 Page 38 of 38 FIGURE 2.3-20 Arkansas Nuclear One Unit 2 Cycle 16 Radial Assembly Fr Distribution Comparison Between Plant measurement and ANC - 445.0 EFPD Legend: ANC Predicted Assembly Fr Measured Data
% Difference 0.920 1.100 1.169 1.343 1.189 1.184 1.267 0.862 0.905 1.066 1.165 1.343 1.181 1.171 1.266 0.855 1.702 3.199 0.361 -0.030 0.643 1.102 0.111 0.831 1.100 1.122 1.329 1.140 1.347 1.152 1.269 0.853 1.062 1.121 1.327 1.145 1.349 1.147 1.268 0.846 3.598 0.089 0.188 -0.419 -0.163 0.462 0.110 0.839 1.169 1.329 1.142 1.355 1.206 1.354 1.274 0.670 1.162 1.338 1.139 1.363 1.208 1.356 1.272 0.667 0.602 -0.643 0.263 -0.594 -0.141 -0.111 0.141 0.390 1.343 1.140 1.355 1.154 1.361 1.342 0.926 1.338 1.145 1.356 1.155 1.362 1.355 0.911 0.389 -0.393 -0.103 -0.104 -0.095 -0.952 1.669 1.189 1.346 1.207 1.362 1.122 1.239 0.716 1.175 1.353 1.206 1.365 1.119 1.262 0.719 1.183 -0.488 0.075 -0.227 0.304 -1.854 -0.403 1.184 1.154 1.353 1.342 1.239 0.794 1.160 1.142 1.348 1.350 1.266 0.800 2.069 1.095 0.401 -0.615 -2.140 -0.725 1.267 1.270 1.274 0.926 0.730 1.253 1.257 1.263 0.889 0.740 1.142 1.042 0.839 4.150 -1.378 0.862 0.853 0.671 0.847 0.842 0.674 1.807 1.270 -0.416