ML020090172

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Request to Revise Technical Specifications: Safety Limit Minimum Critical Power Ratios (Slmcpr). Proprietary Material Enclosed
ML020090172
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 01/04/2002
From: Sumner H
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
HL-6163
Download: ML020090172 (19)
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{{#Wiki_filter:Lewis Sumner Southern Nuclear Vice President Operating Company, Inc. Hatch Project Support 40 Inverness Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 Fax 205.992.0341 SOUTHERN N COMPANY Energy to Serve Your Worid S January 4, 2002 Docket No. 50-321 HL-6163 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications: Safety Limit Minimum Critical Power Ratios (SLMCPR) Ladies and Gentlemen: In accordance with the provisions of 10 CFR 50.90, as required by 10 CFR 50.59(c)(1), Southern Nuclear Operating Company (SNC) hereby proposes a change to the Plant Hatch Unit 1 Technical Specifications, Appendix A to Operating License DPR-57. This application proposes to change the Safety Limit Minimum Critical Power Ratio (SLMCPR) for Single Loop Operation in Technical Specification (TS) 2.1.1.2 to reflect results of a cycle-specific calculation performed for Unit 1 Operating Cycle 21, using NRC-approved methodology for determining SLMCPRs. Enclosure 1 provides a description of the proposed change and an explanation of the basis for the change. Enclosure 2 details the bases for SNC's determination that the proposed change does not involve a significant hazards consideration. Enclosure 3 provides page change instructions for incorporating the proposed change. Following Enclosure 3 are the revised Technical Specifications page and the corresponding marked-up page. The information supporting this proposed change was provided by Global Nuclear Fuel and is considered to be Global Nuclear Fuel proprietary information as described in 10 CFR 2.790(a)(4) and the attached affidavit (Attachment 1). It is requested that this information be withheld from public disclosure. Proprietary text is denoted in Enclosure 1 by enclosure in double brackets. A nonproprietary version of Enclosure 1 is attached for public disclosure (Attachment 2). Southern Nuclear Operating Company requests the proposed amendment for Cycle 21 be issued with the amendment to be effective prior to the restart from the Plant Hatch Unit 1 outage currently scheduled to begin in March 2002. In accordance with the requirements of 10 CFR 50.91, the designated State official will be sent a copy of this letter and all applicable enclosures.

U.S. Nuclear Regulatory Commission Page 2 January 4, 2002 Mr. H. L. Sumner, Jr. states he is Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operating Company, and to the best of his knowledge and belief, the facts set forth in this letter are true. Respectfully submitted, H. L. Sumner, Jr. Sworn to and subscribedbefore me this day of 200L, Notary Public Commission ExpirationDate: 6-- -o 5

Enclosures:

1. Basis for Change Request
2. 10 CFR 50.92 Evaluation
3. Page Change Instructions Attachments:
1. Affidavit of Proprietary Information
2. Nonproprietary Version of the Basis for Change Request cc: Southern Nuclear Operating Company Mr. P. H. Wells, Nuclear Plant General Manager SNC Document Management (R-Type A02.001)

U.S. Nuclear Regulatory Commission, Washington, D.C. Mr. L. N. Olshan, Project Manager - Hatch U.S. Nuclear Regulatory Commission, Region II Mr. L. A. Reyes, Regional Administrator Mr. J. T. Munday, Senior Resident Inspector - Hatch State of Georgia Mr. L. C. Barrett, Commissioner - Department of Natural Resources HL-6163

Enclosure 2 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications: Safety Limit Minimum Critical Power Ratios (SLMCPR) 10 CFR 50.92 Evaluation In 10 CFR 50.92(c), the NRC provides the following standards to be used in determining the existence of a significant hazards consideration: 

         .a proposed amendment to an operating license for a facility licensed under
        §50.21(b) or §50.22 or for a testing facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety. Southern Nuclear Operating Company has reviewed the proposed license amendment request and determined its adoption does not involve a significant hazards consideration based on the following discussion. Basis for no significanthazardsconsiderationdetermination

1. The proposed Technical Specification change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The derivation of the revised SLO SLMCPR for Plant Hatch Unit 1 Cycle 21 for incorporation into the TS, and its use to determine cycle-specific thermal limits, has been performed using NRC-approved methods and procedures. The procedures incorporate cycle specific parameters and reduced power distribution uncertainties in the determination of the value for the SLMCPR. These calculations do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient. The basis of the MCPR Safety Limit is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLO SLMCPR preserves the existing margin to transition boiling and the probability of fuel damage is not increased. Therefore, the proposed change does not involve an increase in the probability or consequences of an accident previously evaluated. HL-6163 E2-1 10 CFR 50.92 Evaluation

2. The proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change is the result of a cycle-specific application of NRC-approved methods to the Unit 1 Cycle 21 core reload. This change does not involve any new method for operating the facility and does not involve any facility modifications. No new initiating events or transients result from this change. Therefore, the proposed TS change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed TS change does not involve a significant reduction in a margin of safety.

The margin of safety as defined in the TS bases will remain the same. Cycle-specific SLMCPRs are calculated using NRC-approved methods and procedures, and meet the current fuel design and licensing criteria. The SLO SLMCPR will be high enough to ensure that greater than 99.9% of all fuel rods in the core are expected to avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. Therefore, the proposed TS change does not involve a reduction in the margin of safety. The proposed change has been reviewed and recommended for approval by the Plant Hatch Plant Review Board and reviewed by the Safety Review Board. ENVIRONMENTAL IMPACT The proposed Technical Specification change was reviewed against the criteria of 10 CFR 51.22 for environmental considerations. The proposed change does not involve a significant hazards consideration, a significant increase in the amounts of effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposures. Based on the foregoing, Southern Nuclear Operating Company concludes the proposed Technical Specification meets the criteria given in 10CFR5 1.22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement. CONCLUSION Based on the evaluation above: (1) there is reasonable assurance the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the proposed amendment will not be inimical to the common defense and security or the health and safety of the public. SCHEDULE OF CHANGE This amendment is needed to support Unit 1 Operating Cycle 21 and will be implemented following refueling outage (RFO) 20, following receipt of NRC approval. HL-6163 E2-2

Enclosure 3 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications: Safety Limit Minimum Critical Powers Ratios (SLMCPR) Page Change Instructions Unit 1 Page Replace 2.0-1 2.0-1 HL-6163 E3-1

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: THERMAL POWER shall be 5 25% RTP. 2.1.1.2 With the reactor steam dome pressure Ž 785 psig and core flow ý 10% rated core flow: MCPR shall be Ž 1.07 for two recirculation loop operation or Ž 1.09 for single recirculation loop operation. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be

  • 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed: 2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods. 2.2.3 Within 24 hours, notify the plant manager, the corporate executive responsible for overall plant nuclear safety, and the offsite review committee. (continued) 2.0-1 MCPR 12/19/01 HATCH UNIT 1

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: THERMAL POWER shall be

  • 25% RTP.

2.1.1.2 With the reactor steam dome pressure Ž 785 psig and core flow 2 10% rated core flow: MCPR shall be - 1.07 for two recirculation loop operation or 4 for single recirculation loop operation. I,01 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be < 1325 psig. 2.2 SL Violations completed: With any SL violation, the following actions shall be in accordance 2.2.1 Within 1 hour, notify the NRC Operations Center, with 10 CFR 50.72. 2.2.2 Within 2 hours: 2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods. executive 2.2.3 Within 24 hours, notify the plant manager, the corporate the offsite responsible for overall plant nuclear safety, and review committee. (continued) 2.0-1 Amendment No.-4-1.- HATCH UNIT 1

ATTACHMENT 1 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications: Safety Limit Minimum Critical Power Ratios (SLMCPR) Affidavit of Proprietary Information

Global Nuclear Fuel A Joint Venturo of GE, Toshiba, & Hitachi Affidavit I, Glen A. Watford, being duly sworn, depose and state as follows: L.L.C. ("GNF-A") (1) I am Manager, Fuel Engineering Services, Global Nuclear Fuel - Americas, in paragraph (2) and have been delegated the function of reviewing the information described which is sought to be withheld, and have been authorized to apply for its withholding. Information (2) The information sought to be withheld is contained in the attachment, "Additional Regarding the Cycle Specific SLMCPR for Hatch Unit I Cycle 21," December 11, 2001. it is the owner or (3) In making this application for withholding of proprietary information of which set forth in the Freedom of licensee, GNF-A relies upon the exemption from disclosure 18 USC Sec. 1905, Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, commercial or and NRC regulations 10 CFR 9.17(a)(4) and 2.790(a)(4) for "trade secrets and (Exemption 4). The financial information obtained from a person and privileged or confidential" commercial material for which exemption from disclosure is here sought is all "confidential under the narrower definition of "trade secret," information," and some portions also qualify Exemption 4 in, respectively, within the meanings assigned to those terms for purposes of FOIA 1992), and Critical Mass Energy Project v. Nuclear Regulatory Commission 975F2d871 (DC Cir. Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983). of proprietary (4) Some examples of categories of information which fit into the definition information are: data

a. Information that discloses a process, method, or apparatus, including supporting and analyses, where prevention of its use by GNF-A' s competitors without license from GNF-A constitutes a competitive economic advantage over other companies; resources
b. Information which, if used by a competitor, would reduce his expenditure of or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; budget
c. Information which reveals cost or price information, production capacities, levels, or commercial strategies of GNF-A, its customers, or its suppliers;
d. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, of potential commercial value to GNF-A; to
e. Information which discloses patentable subject matter for which it may be desirable obtain patent protection.

set The information sought to be withheld is considered to be proprietary for the reasons forth in paragraphs (4)a. and (4)b., above. (5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, best of are as set forth in (6) and (7) following. The information sought to be withheld has, to the my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure Page 1

Affidavit All disclosures to third parties including has been made, and it is not available in public sources. or must be made, pursuant to regulatory any required transmittals to NRC, have been made, for maintenance of the information in provisions or proprietary agreements which provide confidence. is made by the manager of the originating (6) Initial approval of proprietary treatment of a document with the value and sensitivity of the component, the person most likely to be acquainted to the terms under which it was licensed information in relation to industry knowledge, or subject on a "need to know" basis. to GNF-A. Access to such documents within GNF-A is limited a document typically requires review by (7) The procedure for approval of external release of such or other equivalent authority, by the the staff manager, project manager, principal scientist and by the Legal Operation, for manager of the cognizant marketing function (or his delegate), the accuracy of the proprietary technical content, competitive effect, and determination of bodies, customers, and potential designation. Disclosures outside GNF-A are limited to regulatory others with a legitimate need for the customers, and their agents, suppliers, and licensees, and regulatory provisions or proprietary information, and then only in accordance with appropriate agreements. as proprietary because it contains details (8) The information identified in paragraph (2) is classified of GNF-A's fuel design and licensing methodology. with the testing, development and The development of the methods used in these analyses, along cost, on the order of several approval of the supporting methodology was achieved at a significant million dollars, to GNF-A or its licensor. is likely to cause substantial harm to (9) Public disclosure of the information sought to be withheld the availability of profit-making GNF-A's competitive position and foreclose or reduce is part of GNF-A's comprehensive opportunities. The fuel design and licensing methodology value extends beyond the original BWR safety and technology base, and its commercial the extensive physical database development cost. The value of the technology base goes beyond expertise to determine and apply the and analytical methodology and includes development of the includes the value derived from appropriate evaluation process. In addition, the technology base providing analyses done with NRC-approved methods. review costs comprise a substantial The research, development, engineering, analytical, and NRC investment of time and money by GNF-A or its licensor. and apply the correct analytical The precise value of the expertise to devise an evaluation process methodology is difficult to quantify, but it clearly is substantial. are able to use the results of the GNF-A's competitive advantage will be lost if its competitors or if they are able to claim an GNF-A experience to normalize or verify their own process at the same or similar conclusions. equivalent understanding by demonstrating that they can arrive information were disclosed to the The value of this information to GNF-A would be lost if the their having been required to public. Making such information available to competitors without provide competitors with a windfall, undertake a similar expenditure of resources would unfairly advantage to seek an adequate and deprive GNF-A of the opportunity to exercise its competitive very valuable analytical tools. return on its large investment in developing and obtaining these C:\U serdata\Haehl\C21\SLMCPR\SLMCPR-affidavit'doe Page 2

Affidavit State of North Carolina ) SS: County of New Hanover ) Glen A. Watford, being duly sworn, deposes and says: That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief. Executed at Wilmington, North Carolina, this _3__ day of ,20_0 

                                                                                                  ,ce Global Nuclear Fuel - Americas, LLC Subscribed and sworn before me this j_                      day of   j"*ce_.      §A4      , 20Q!

JAMES E. McGINNESS Notary Public, State of North Carolina Notary Public, State of North Carolina New Hanover County My Commission Expires My Commision Expires. C:\Userdata\Hatch \C21\SLMCPR\SLMCPR affidavit.doe Page 3

ATTACHMENT 2 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications: Safety Limit Minimum Critical Power Ratios (SLMCPR) Nonproprietary Version of the Basis for Change Request

Enclosure 1 Edwin I. Hatch Nuclear Plant - Unit 1 Request to Revise Technical Specifications: Safety Limit Minimum Critical Power Ratios (SLMCPR) Basis for Change Request PROPOSED CHANGES 1 SNC requests that the Technical Specifications (TS) contained in Appendix A to the Plant Hatch Unit to reflect a Operating License DPR-57 be amended to revise Technical Specifications Section 2.1.1.2 change in the Single Loop Operation (SLO) Safety Limit Minimum Critical Power Ratio (SLMCPR), which is based on Global Nuclear Fuel's (GNF) application of GE's NRC-approved methodology for calculating SLMCPRs. BACKGROUND The proposed change involves revising the SLO SLMCPR contained in Section 2.1.1.2 of the Plant Hatch Unit 1 TS. In the course of calculating a cycle-specific SLMCPR for another utility, it was determined that the GESTAR II (General Electric StandardApplicationfor ReactorFuel, NEDE-24011 P-A-l 11, and U. S. Supplement NEDE-2401 1-P-A- 11-US1, November 17, 1995) fuel type generic SLMCPR may be non-conservative when applied to some core and fuel designs. To rectify this deficiency, GE proposed, and the NRC accepted, a new procedure for determining cycle-specific SLMCPRs (Reference 1). GE also proposed, and the NRC has accepted, the application of reduced power distribution uncertainties in the calculation of SLMCPRs for plants using the 3D MONICORE model in the process computer for core monitoring (Reference 1). DISCUSSION OF THE PROPOSED CHANGE GNF's calculation for the plant-specific SLMCPR values for Unit 1 Cycle 21 is based upon NRC approved methods and procedures for calculating SLMCPRs each operating cycle for plants using the 3D MONICORE system. The procedures incorporate cycle-specific parameters into the analysis, including the reference loading pattern and actual bundle parameters, which are evaluated at the projected exposure distribution based on projected control blade patterns for the rodded bum through the cycle. The analysis considers the full cycle exposure range to determine the most limiting point(s). At these exposure point(s), conservative variations of the projected control blade patterns are used to maximize the number of bundles that contribute rods calculated to be susceptible to boiling transition in order to obtain a conservative calculation of the SLMCPR. The calculation also includes the application of reduced power distribution uncertainties associated with the 3D MONICORE core monitoring system. This calculation resulted in a Cycle 21 SLMCPR value of 1.07 for dual loop operation (DLO) (which is the same as the current Cycle 20 value), and 1.09 for SLO (which currently has a value of 1.08). Therefore, only the SLO SLMCPR value is being revised. Note that an increase in SLMCPR is more restrictive for plant operations; therefore, this is a conservative change. 1 Revision 11 has since been superseded by Revision 14, dated June, 2000. This revision incorporates the material contained in Reference 1. HL-6163 A2-1 Nonproprietary Version of the Basis for Change Request EVALUATION The proposed change revises the Technical Specifications to reflect a change in the SLO SLMCPR due to the plant-specific evaluation performed by GNF for Unit 1, Reload 20, Cycle 21. Cycle-specific DLO and SLO SLMCPRs were calculated using NRC-approved methods and procedures (Reference 1). The procedures incorporate plant and cycle-specific parameters which include: 1) the expected reference loading pattern, 2) conservative variations of projected control blade patterns, 3) the actual bundle parameters, 4) the full cycle exposure range, and 5) reduced power distribution uncertainties associated with the process computer system. The SLMCPR is set such that no mechanistic fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the SLMCPR is defined as the CPR in the limiting fuel assembly for which more the 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The SLMCPRs for Cycle 21 at Unit 1 are 1.07 for DLO (no change from Cycle 20) and 1.09 for SLO, which is an increase of 0.01 from the Cycle 20 value. COMPARISON OF HATCH UNIT 1 SLMCPR VALUES FOR CYCLES 21 AND 20 Table 1 summarizes the relevant input parameters and results of the SLMCPR determination for the Hatch Unit 1 Cycle 21 and 20 cores. The SLMCPR evaluations were performed using NRC approved methods and uncertainties (Reference 1). These evaluations yield the same calculated dual-loop SLMCPR values even though different inputs were used. The quantities that have been shown to have some impact on the determination of the safety limit MCPR (SLMCPR) are provided. In comparing the Hatch Unit 1 Cycle 21 and Cycle 20 SLMCPR values it is important to note the impact of the differences in the core and bundle designs. These differences are summarized in Table 1. In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle by-bundle MCPR distributions and (2) flatness of the bundle pin-by-pin power/R-factor distributions. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR. [11] The uncontrolled bundle pin-by-pin power distributions were compared between the Hatch Unit 1 Cycle 21 bundles and the Cycle 20 bundles. Pin-by-pin power distributions are characterized in terms of R factors using the NRC approved methodology (Reference 2). For the Hatch Unit 1 Cycle 21 limiting case analyzed at peak hot excess, [[ ]] the Hatch Unit 1 Cycle 21 bundle power distributions are flatter than the bundle power distributions used for the Cycle 20 SLMCPR analysis. HL-6163 A2-2 Nonproprietary Version of the Basis for Change Request

SUMMARY

on these [[ ]] have been used to compare quantities that impact the calculated SLMCPR value. Based comparisons, the conclusion is reached that the Hatch Unit 1 Cycle 20 core has a flatter core MCPR Unit 1 distribution L than what was used to perform the Cycle 21 SLMCPR evaluation; and the Hatch the Cycle 20 Cycle 21 core has a flatter in-bundle power distributions [[L] than what was used to perform SLMCPR evaluation. would The calculated 1.07 Monte Carlo SLMCPR for Hatch Unit 1 Cycle 21 is consistent with what one expect [[ ]] the 1.07 SLMCPR value is appropriate. Based on all of the facts, observations and arguments presented above, it is concluded that the calculated value SLMCPR value of 1.07 for the Hatch Unit 1 Cycle 21 core is appropriate. It is reasonable that this is same as the 1.07 value calculated for the previous cycle. For single loop operations (SLO) the calculated safety limit MCPR for the limiting case is 1.09 as determined by specific calculations for Hatch Unit 1 Cycle 21. SUPPORTING INFORMATION The following information is provided in response to NRC questions on similar submittals regarding changes in Technical Specification values of SLMCPR. NRC questions pertaining to how GE14 4. applications satisfy the conditions of the NRC SER (Reference 1) have been addressed in Reference Other generically applicable questions related to application of the GEXL 14 correlation and the applicable range for the R-factor methodology are addressed in Reference 5. Only those items that require a plant/cycle specific response are presented below since all the others are contained in the references that have already been provided to the NRC. The core loading information for Hatch Unit 1 Cycles 20 and 21 is provided in Figures 1 and 2, respectively. The impact of the fuel loading pattern differences on the calculated SLMCPR is correlated to the values of [[ ]]. The power and non-power distribution uncertainties that are used in the analyses are indicated in Table 1. The referenced document numbers have previously been reviewed and approved by the NRC. HL-6163 A2-3 Nonproprietary Version of the Basis for Change Request Table 1 Comparison of the Hatch Unit 1 Cycle 20 and Cycle 21 SLMCPR QUANTITY, DESCRIPTION Hatch Unit 1 Hatch Unit 1 Cycle 20 Cycle 21 Number of Bundles in Core 560 560 Limiting Cycle Exposure Point EOC-1.0K PHE Cycle Exposure at Limiting Point [MWcl/STU]_ 11454 9000 Reload Fuel Type GE13 GE14 32.9% 40.0% Latest Reload Batch Fraction [%] 3.98% 3.78% Latest Reload Average Batch Weight % Enrichment 40.0% 0.0% Batch Fraction for GEl4 60.0% 100.0% Batch Fraction for GEl3 3.86% 3.70% Core Average Weight % Enrichment 1.38 1.32 Core MCPR (for limiting rod pattern) [[i [[i Reduced Reduced Power distribution uncertainty NEDO-32694P-A NEDO-32694P-A Revised Revised Non-power distribution uncertainty NEDC-32601P-A NEDC-32601P-A 1.07 1.07 Calculated Safety Limit MCPRII HL-6163 A2-4 Nonproprietary Version of the Basis for Change Request Figure 1 Reference Core Loading Pattern - Cycle 20 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 A B B A B B A B B A 52 1 - - .L -I. - I. - .4.- I. - - I- *4"- 50 2 B F F F F F F F F F F B El l A 48 3 A P F: 1H Fl H F G G Fl H F H H E B A EH HH Fl G- Gb F 46 4 r* F.l F: F= n F HI B G 0 D G B H F G E F F B 44 5 BFFH H F IG F GI GF F IF F IGJFG GJF 1 H HjF F B 42 6 LAF HEC F D DGIA G DIDIG A[G DF C E EH~ FA-40 7 U I= M- F" M F: Gd Fl ( F G G FIG F G E H C H E:: B G FGG HI I 38 I3 I= r.: I:: F. I- rd c r) B B GI 0 G C H F F F G E. B 8 36 F H F A 9 A F H F G D G C G A G F G G F G A G C G D G 34 G F G A G B F B B F B G A G F G F H F F B 10 B F F H F 32 G D G B G F G G F G B G D G A G B H E B 11 B E H B G A A F F G F G F G F F F G C C G F E F G F G F G F F A 30 12 F B 28 D G B G B G C B B C G B G B G D F D G 13 B F G D F 26 B G B G C B B C G B G B G D F D G F B 14 B F G D F D G 24 G F E F G C C G F F F G F G F G F F A 15 A F F G F G F 22 A G D G B G F G G F G B G D G A G B H E B 16 B E H B G B F F H F G F G A G B F B B FIB GJA G F G F H F F B 20 17 A GF G G F G A G C G D G F H F A 18 A F H F G D G C G 18 16 19 RI F- C; Fl F F H C G D G B B G D G C H F F F G5 E B F El G B- IF 14 20 RI  : HI C H E GI F G F G G F G F G E H C H E5 B 12 21 F H E C F D G A G D D G A G D FI C E H F A A B F F H H F G F G F F G F G F H H F F B 10 22 I Ei 08 23 P Fl F E G F H B G D 0 G B H F G E F F IB BB A B 06 24 RI A B E H F H F G,G F H F HI E . B A B B A, , ,  ! 04 BI F F E FI Fl F F E F F B  ! 25 

                                                     -~~       -        -.      4-     4    -        -   4
                                                                                                       '4-..-    1                                           02 26                                            1A B BI Al BI B A B B A 01 03 05 07 09 11 13 15 17 19 21 23 25 27                                     ' 29 31 33 35 37 39 41 43 45 47 49 51 FUEL TYPE A  GE13-P9HTB355-4G5.0/6G4.0-10OT-146-T B GE 13-P9HTB355-12GZ-100T-146-T C GE13-P91-TB355-4G5.0/6G4.0-10OT-146-T D GE13-P9HTB355-12GZ-10OT-146-T E GE13-P9HTB378-6G5.0/6G4.0/lG2.0-100T-146-T F GE13-P9HTB378-6G5.0/6G4.0-100T-146-T G GE13-P9DTB3786G5.0/6G4.0-100T-146-T-2411 H GE13-P9DTB378-6G5.0/6G4.O/1G2.0-100T-146-T HL-6163                                                                            A.2-5 Nonproprietary Version of the Basis for Change Request Figure 2 Reference Core Loading Pattern - Cycle 21 1   2    3   4     5    6     7    8   9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 B B       B    BB       B   B                                           52 1                                          AI B                                                                              50 C             C    C C 2                                       R             C C                     C CI A E lA              1I                      48 RI A SFlF Fl E C                  Cl El E E E A I                     I BJ 3                                     1     ECE       E         CI   C  E   E                                                  46 RR    DIC IlC l

E E C F C F F C F C E E C D B B 4 44 5 B BBD CEDF D . C EE B ED C F+ Cj EB I~~ B ED ElEB FC D E1 C~uD 1 BB 42 6 B EDF CF, EC FICG C; F D CJ 40 7 DI P-I r* I= r, F. C't F: R F Cl C E B F C E C F" D. C, B CI Bi F E- Cs C E 38 8 r* F- SF F (Z, I- U I" i.J E C F r-- A' B 36 A A E E B E C F B E B E B B E B E B F C E B E E A IA1 9 C IB1 34 B C E C El C F1 D E C1 El D E E DI E C El DI F C El C E 10 B C El F D F B E B E A E C C E A E B E B F D F E C B 32 11 C B 30 B C1 El C El C El C E D El C F F C1 E D El C1 E C El C E 12 B C1 C1 F B IE C1 F B E C1 F B B, F1 C E BI F1 C E BI F, C C B 28 13 B C1 C1 F B B E IIC1 F B E C1 F B BI F1 CEl ED BI F] C E BI F1 C C B 26 14 B 24 B C1 El C E C El C E D El C F F1 C1 E D El C1 E C El C1 E C 15 B1 22 B C El F_ D F BI E B E Al E C1 C El A E BI E B F DI F1 E C1 16 E C E D F C E I C I E C B 20 B CE C E C F D EC E EDE D 17 C'E B IE E A A 18 A A E E B E C1 F Bj E B E BBE B E B F 18 16 19 D A I= t-l F.in F 1" FIG F F C E D F D. E C F" E A- B E El CBE DFF C E 14 20 P C I' FIG E C F B E C C E B F C E C F D C 1 12 21 B D E DI F C E C F C E E C F, C E CI F D DE4 B 22 B' C'I BJ E D4t B EFE NE FID E B DLýEB BDTDE 10 08 23 P R nI C F E C F C F F C F C E E D B B E C C F El ClB Albib 06 24 fq Rl R A F El E E I C E E E E. A B I B B 04 B B B A E E 25 B A C C C C C C1 C C1 A IB' *C 02 26 A 'I B B [_B B B B A 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 FUEL'YPE A GE13-P9-B378-6G5.0/6G4.01cG2.0-10OT-146-T B GE13-P9HB378-65.0/6G4.0-10OT-1146-T C GE13-P9DTB378-6G5.016G4.0-1OOT-146-T-2411 D GE13-P9DTB378.6G5.0/6G4.0/1GI.0-10OT-146-T E GE14-P1ODNA398-15GZ-10OT-150-T-2518 F GE14-P1ODNAB399-16GZ-1OOT-150-T-2517 HL-6163 A2-6 Nonproprietary Version of the Basis for Change Request CONCLUSION Based on all of the information presented above, it is concluded that the calculated SLMCPR values of 1.07 and 1.09, for dual loop and single loop operation, respectively, for the Hatch-1 Cycle 21 core are appropriate. REFERENCES

1. Letter, Frank Akstulewicz (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-3260 lP, Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations;NEDC-32694P, Power Distribution Uncertaintiesfor Safety Limit MCPR Evaluation; and Amendment 25 to NEDE-2401 1-P-A on Cycle Specific Safety Limit MCPR,"

(TAC Nos. M97490, M99069 and M97491), March 11, 1999.

2. Letter, Thomas H. Essig (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Report NEDC-32505P, Revision 1, R-Factor CalculationMethodfor GE]],

GEJ2 and GE]3 Fuel," (TAC No. M99070 and M9508 1), January 11, 1999.

3. GeneralElectric B WR Thermal Analysis Basis (GETAB): Data, Correlationand Design Application, NEDO-10958-A, January 1977.
4. Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to R. Pulsifer (NRC), "Confirmation of lOx 10 Fuel Design Applicability to Improved SLMCPR, Power Distribution and R-Factor Methodologies", FLN-2001-016, September 14, 2001.
5. Letter, Glen A. Watford (GNF-A) to U. S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC), "Confirmation of Applicability of the GEXL 14 Correlation and Associated R-Factor Methodology for Calculating SLMCPR Values in Cores Containing GE 14 Fuel", FLN-2001-017, October 1, 2001.

HL-6163 A2-7}}

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