NLS2002129, License Amendment Request to Revise TS-Safety Limit Minimum Critical Power Ratio, Cooper Nuclear Station

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License Amendment Request to Revise TS-Safety Limit Minimum Critical Power Ratio, Cooper Nuclear Station
ML023240456
Person / Time
Site: Cooper Entergy icon.png
Issue date: 11/15/2002
From: Coyle M
Nebraska Public Power District (NPPD)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2002129
Download: ML023240456 (26)
v • d • e


Text

H Nebraska Public Power District Nebraska's Energy Leader 50.90 NLS2002129 November 15, 2002 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

License Amendment Request to Revise Technical Specifications - Safety Limit Minimum Critical Power Ratio Cooper Nuclear Station, Docket 50-298, DPR-46 The purpose of this letter is for the Nebraska Public Power District (NPPD) to request an amendment to Facility Operating License DPR-46 in accordance with the provisions of 10 CFR 50.4 and 10 CFR 50.90 to revise the Cooper Nuclear Station (CNS) Technical Specifications (TS). This proposed TS change will revise dual recirculation loop and single recirculation loop Safety Limit Minimum Critical Power Ratio values to reflect results of a cycle specific calculation This change is needed to support Cycle 22 operations. Completion of the Cycle 21 Refueling Outage and startup in Cycle 22 is scheduled for early April 2003. Therefore, NPPD requests Nuclear Regulatory Commission approval of the proposed TS change and issue of the requested license amendment by March 1, 2003. Once approved, the amendment will be implemented within 30 days.

Attachment I provides a description of the TS change, the basis for the amendment, the no significant hazards consideration evaluation pursuant to 10 CFR 50.9 1(a)(1), and the environmental impact evaluation pursuant to 10 CFR 51.22. Attachment 2 provides the specific changes to the current CNS TS on marked up pages. Attachment 3 provides the final, clean typed versions of the affected TS page. Attachment 4, List of Regulatory commitments, reflects that there are no commitments in this submittal. No Bases pages are affected by this amendment request.

The information supporting this proposed change was prepared by Global Nuclear Fuel - Americas (GNF-A) and is considered to be GNF-A proprietary information as described in 10 CFR 2.790(a)(4) The proprietary information is provided as Attachment 5, with specific proprietary text enclosed within double brackets. It is requested that this information be withheld from public disclosure. The affidavit required by 10 CFR 2.790(b)(1) is provided as Attachment 6. A nonproprietary version of Attachment 5 for public disclosure is provided as Attachment 7.

Cooper Nuclear Station P0 Box 98 / Brownville, NE 68321-0098 0 J Telephone: (402) 825-3811 / Fax: (402) 825-5211 http //www nppd corn

NLS2002129 Page 2 of 3 This proposed TS change has been reviewed by the necessary safety review committees (Station Operations Review Committee and Safety Review and Audit Board). Amendments to the CNS Facility Operating License through Amendment 195 issued September 30, 2002, have been incorporated into this request. NPPD has concluded that the proposed changes do not involve a significant hazards consideration and that they satisfy the categorical exclusion criteria of 10 CFR 51.22(c).

This request is submitted under oath pursuant to 10 CFR 50.30(b). By copy of this letter and its attachments, the appropriate State of Nebraska official is notified in accordance with 10 CFR 50.91 (b)(1). Copies to the NRC Region IV office and the CNS Resident Inspector are also being provided in accordance with 10 CFR 50.4(b)(1).

Should you have any questions concerning this matter, please contact Mr. Paul Fleming at (402) 825-2774.

Sincerely, Michael T. Coyle Site Vice President

/cb Attachments cc: Regional Administrator w/ attachments USNRC - Region IV Senior Project Manager w/ attachments USNRC - NRR Project Directorate IV-l Senior Resident Inspector w/ attachments USNRC Nebraska Health and Human Services w/ attachments Department of Regulation and Licensure NPG Distribution w/o attachments Records w/ attachments

NLS2002129 Page 3 of 3 Affidavit STATE OF NEBRASKA )

)

NEMAHA COUNTY )

Michael T. Coyle, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this correspondence on behalf of Nebraska Public Power District; and that the statements contained herein are true to the best of his knowledge and belief.

Michael T. Coyle Subscribed in my presence and sworn to before me this /.0'dayof zee r- 2002.

NOTARY PUBLIC

NLS2002129 Attachment I Page 1 of 7 NPPD's Evaluation 1.0 Introduction 2.0 Description of Proposed Amendment 3.0 Background 4.0 Regulatory Requirements and Guidance 5.0 Technical Analysis 6.0 Regulatory Analysis 7.0 No Significant Hazards Consideration (NSHC) 8.0 Environmental Consideration 9.0 Precedents 10.0 References

NLS2002129 Attachment I Page 2 of 7 License Amendment Request to Revise Technical Specifications Safety Limit Minimum Critical Power Ratio Cooper Nuclear Station, NRC Docket 50-298, DPR-46 Revised Pages 2.0-1 (No Bases pages are being changed) 1.0 Introduction This letter is a request to amend Operating License (OL) DPR-46 for Cooper Nuclear Station (CNS).

The proposed changes would revise the OL to change the Safety Limit Minimum Critical Power Ratio (SLMCPR) for both two recirculation (dual) loop operation and single recirculation loop operation in Technical Specification (TS) 2.1.1.2 to reflect results of a cycle specific calculation performed for CNS operation Cycle 22, using Nuclear Regulatory Commission (NRC) approved methodology. Completion of the Cycle 21 Refueling Outage and startup in Cycle 22 is scheduled for early April 2003. Therefore, NPPD requests Nuclear Regulatory Commission approval of the proposed TS change and issue of the requested license amendment by March 1, 2003. Once approved, the amendment will be implemented within 30 days.

2.0 Description of Proposed Amendment This license amendment request proposes the following changes:

TS 2.1.1.2 will be revised to change the two recirculation loop MCPR from a 1.08 to > 1.09 and the single recirculation loop MCPR from > 1.09 to > 1.11.

These proposed changes are necessary to reflect cycle specific calculations for CNS Cycle 22 operations.

There are no changes to the associated TS Bases.

3.0 Background The CNS Cycle 22 core has 548 fuel assemblies, consisting of 128 fresh General Electric GEl4 fuel bundles, 256 irradiated GEl4 fuel bundles, and 164 irradiated GE9B fuel bundles.

Calculations by Global Nuclear Fuels - Americas (GNF-A) for the CNS Cycle 22 SLMCPR values are based on NRC approved methods and procedures.

NLS2002129 Page 3 of 7 4.0 Regulatory Requirements and Guidance 10 CFR 50 Appendix A Criterion 10 - Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

The fuel cladding must not sustain damage as a result of normal operation and abnormal operational transients. The reactor core safety limits are established to preclude violation of the fuel design criterion that a SLMCPR is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.

5.0 Technical Analysis Design Basis:

The safety design basis provided in the Updated Safety Analysis Report (USAR) (USAR 111-7) is that the thermal hydraulic design of the core shall establish a thermal hydraulic safety limit for use in evaluating the safety margin relating the consequences of fuel barrier failure to public safety. To ensure that adequate margin is maintained, a design requirement based on a statistical analysis was selected as follows. Moderate frequency transients caused by a single operator error or equipment malfunction shall be limited such that, considering uncertainties in manufacturing and monitoring the core operating state, at least 99.9% of the fuel rods would be expected to avoid boiling transition. The lowest allowable transient minimum critical power ratio (MCPR) limit which meets the design requirement is termed the fuel cladding integrity SLMCPR.

A plant unique operating limit MCPR is established to provide adequate assurance that the fuel cladding integrity SLMCPR is not exceeded for any anticipated operational transients.

The operating limit MCPR is obtained by adding the maximum delta critical power ratio (CPR) value for the most limiting transient postulated to occur at the plant to the fuel cladding integrity SLMCPR. Cycle specific delta critical power ratio values are determined as part of the reload analysis and are reported in the Supplemental Reload Licensing Report.

Analyses have been performed which show that at least 99.9% of the fuel rods in the core are expected to avoid boiling transition (and, therefore, cladding damage due to overheating) if the MCPR is equal to or greater than the fuel cladding integrity SLMCPR.

NLS2002129 Page 4 of 7 6.0 Regulatory Analysis As part of a reload core design, cycle specific transient analyses are performed to determine the required SLMCPR and the delta CPR for specific transients. Attachment 5, "Additional Information Regarding the Cycle Specific SLMCPR for Cooper Cycle 22", compares the SLMCPR value for Cycle 22 with the SLMCPR value for the current operating cycle, Cycle

21. Attachment 5 documents that the SLMCPR evaluations were performed using NRC approved methods and uncertainties. Included is supporting information that documents prior communications by GNF-A to the NRC that deal with NRC questions pertaining to how GE14 applications satisfy the conditions of the NRC SER accepting GNF-A methods and uncertainties, generically applicable questions related to application of the GEXLI-4 correlation and the applicable range for the R-factor methodology. This attachment also provides the core loading information for CNS Cycle 21 and 22.

By using the sum of the maximum delta CPR and cycle specific SLMCPR to determine operating limit MCPR, compliance with General Design Criteria 10, Reactor Design, is preserved.

7.0 No Significant Hazards Consideration 10 CFR 50.91 (a)(1) requires that licensee requests for operating license amendments be accompanied by an evaluation of significant hazard posed by issuance of an amendment.

NPPD has evaluated this proposed amendment with respect to the criteria given in 10 CFR 50.92 (c).

The proposed changes would revise the Cooper Nuclear Station Operating License to increase the values of the Safety Limit Minimum Critical Power Ratio for both two recirculation (dual) loop operation and single recirculation loop operation in Technical Specification 2.1.1.2. The changes reflect results of a cycle specific calculation performed for Cooper Nuclear Station Cycle 22 operation, using Nuclear Regulatory Commission (NRC) approved methodology.

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

The probability of an evaluated accident is derived from the probabilities of the individual precursors to that accident. The consequences of an evaluated accident are determined by the operability of plant systems designed to mitigate those consequences.

Limits have been established, consistent with NRC approved methods, to ensure that fuel performance during normal, transient, and accident conditions is acceptable. The proposed change conservatively establishes the safety limit for the minimum critical power ratio (SLMCPR) for Cooper Nuclear Station Cycle 22 such that the fuel is

NLS2002129 Attachment I Page 5 of 7 protected during normal operation and during any plant transients or anticipated operational occurrences.

Changing the SLMCPR does not increase the probability of an evaluated accident.

The change does not require any physical plant modifications, physically affect any plant components, or entail changes in plant operation. Therefore, no individual precursors of an accident are affected.

The proposed change revises the SLMCPR to protect the fuel during normal operation as well as during any transients or anticipated operational occurrences. Operational limits (MCPR) are established based on the proposed SLMCPR to ensure that the SLMCPR is not violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation and anticipated operational occurrences) is met. Since the operability of plant systems designed to mitigate any consequences of accidents has not changed, the consequences of an accident previously evaluated are not expected to increase.

Based on the above NPPD concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Creation of the possibility of a new or different kind of accident would require the creation of one or more new precursors of that accident. New accident precursors may be created by modifications of the plant configuration, including changes in allowable modes of operation. The proposed change does not involve any modifications of the plant configuration or allowable modes of operation. The proposed change to the SLMCPR assures that safety criteria are maintained for Cycle 22.

Based on the above NPPD concludes that the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in the margin of safety?

The value of the proposed SLMCPR provides a margin of safety by ensuring that no more than 0. 1% of the rods are expected to be in boiling transition if the MCPR limits is

NLS2002129 Attachment I Page 6 of 7 violated during all modes of operation. This will ensure that the fuel design safety criteria (i.e., that at least 99.9% of the fuel rods do not experience transition boiling during normal operation as well as anticipated operational occurrences) are met.

Based on the above NPPD concludes that the proposed changes do not involve a significant reduction in a margin of safety.

From the above discussions, NPPD concludes that the proposed amendment involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of"no significant hazards consideration" is justified.

8.0 Environmental Consideration 10 CFR 51.22(c)(9) provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment.

A proposed amendment to an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amount of any effluents that may be released off-site, or (3) result in an increase in individual or cumulative occupational radiation exposure. NPPD has reviewed the proposed license amendment and concludes that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the proposed license changes. The basis for this determination is as follows:

1. The proposed license amendment does not involve significant hazards as described previously in the No Significant Hazards Consideration Evaluation.
2. This proposed change does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released off-site.

The proposed license amendment does not introduce any new equipment, nor does it require any existing equipment or systems to perform a different type of function than they are presently designed to perform. NPPD has concluded that there will not be a significant increase in the types or amounts of any effluents that may be released off-site and these changes do not involve irreversible environmental consequences beyond those already associated with normal operation.

3. These changes do not adversely affect plant systems or operation and therefore, do not significantly increase individual or cumulative occupational radiation exposure beyond that already associated with normal operation.

NLS2002129 Attachment I Page 7 of 7 9.0 Precedents Similar amendments have been granted by the NRC for (not all inclusive):

1. Edwin I. Hatch Nuclear Plant, Unit 1, Docket No. 50-321, License No. DPR-57, Amendment 229, dated April 5, 2002.
2. Browns Ferry Nuclear Plant, Unit 3, Docket No. 50-296, License No. DPR-68, Amendment 234, dated March 29, 2002.
3. Brunswick Steam Electric Plant, Unit 1, Docket No. 50-325, License No. DPR-71, Amendment 220, dated March 22, 2002.
4. Limerick Generating Station, Unit 1, Docket No. 50-352, License No. NPF-57, Amendment 156, dated March 12, 2002.
5. Monticello Nuclear Generating Plant, Docket No. 50-263, License No. DPR-22, Amendment 125, December 6, 2001.

Although none of these are identical to this request, in that none of the core designs is exactly like the CNS Cycle 22 core design, all share several points in common. First, all are mixed core designs using GEl4 fuel with some other fuel type(s). Second, due to the use of GE14 fuel all were reviewed for applicability of methods, uncertainties and the GEXL14 correlation. Each was determined to not involve a significant hazards consideration and were approved by the NRC.

10 References (None)

NLS2002129 Page 1 of 2 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS REVISIONS MARKUP FORMAT COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Listing of Revised Pages TS Pages 2.0-1

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 > 785 psig and core flow

> 10% rated core flow:

MCPR shall bec> oor two recirculation loop operation or_>

for single recirculaRo 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 Pressure SL Reactor steam dome pressure shall be < 1337 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

Amendment 1 2.0-1

NLS2002129 Page 1 of 2 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS REVISIONS FINAL TYPED FORMAT COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46 Listing of Revised Pages TS Pages 2.0-1

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 > 785 psig and core flow

> 10% rated core flow:

MCPR shall be > 1.09 for two recirculation loop operation or > 1.11 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 Pressure SL Reactor steam dome pressure shall be < 1337 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

Amendment 2.0-1

NLS2002129 Page 1 of 2 ATTACHMENT 4 LIST OF REGULATORY COMMITMENTS COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46

ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS Correspondence Number: NLS2002129

] The following table identifies those actions committed to by Nebraska Public Power District

] (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None

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4

-4 I PROCEDURE 0.42 1 REVISION 11 1 PAGE 13 OF 16

NLS2002129 Page 1 of4 ATTACHMENT 6 AFFIDAVIT COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46

Global Nuclear Fuel A Joint Venture of GE, Toshiba, & Hitachi Affidavit 1, Glen A. Watford, state as follows:

(1) I am Manager, Fuel Engineering Services, Global Nuclear Fuel - Americas, L.L.C.

("GNF-A") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the attachment, "Additional Information Regarding the Cycle Specific SLMCPR for Cooper Cycle 22," October 4, 2002.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4) and 2.790(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information," and some portions also qualify under the narrower definition of "trade secret," within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data 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;
b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;
c. Information which reveals cost or price information, production capacities, budget 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;
e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

Page 1

Affidavit The information sought to be withheld is considered to be proprietary for the reasons set 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, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A's fuel design and licensing methodology.

The development of the methods used in these analyses, along with the testing, development and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to GNF-A or its licensor.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit making opportunities. The fuel design and licensing methodology is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC approved methods.

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A or its licensor.

I \JF'EIhcensmg'affida*f\'gnfa._a~fda% a doc Page 2

Affidavit The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed at Wilmington, North Carolina, this 4th day of October ,2002.

L.Watford Nuclear Fuel - Americas, LLC I \NFE\censing'affidavit\'nfa.afridavit doc Page 3

NLS2002129 Page 1 of 6 ATTACHMENT 7 ADDITIONAL INFORMATION REGARDING THE CYCLE SPECIFIC SLMCPR FOR COOPER CYCLE 22 NON-PROPRIETARY VERSION COOPER NUCLEAR STATION NRC DOCKET 50-298, LICENSE DPR-46

Attachment Additional Information Regarding the October 4,2002 Cycle Specific SLMCPR for Cooper Cycle 22 References

[1] Letter, Frank Akstulewicz (NRC) to Glen A. Watford (GE), "Acceptance for Referencing of Licensing Topical Reports NEDC-32601 P, Methodology and Uncertaintiesfor Safety Limit MCPR Evaluations;NEDC-32694P, PowerDistribution 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 M9749 1), 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-FactorCalculationMethodfor GEl1, GEl2 and GEl3 Fuel," (TAC No. M99070 and M95081), January 11, 1999.

[3] GeneralElectricBWR ThermalAnalysis Basis (GETAB): Data,CorrelationandDesign Application,NEDO-1 0958-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 lOxlO Fuel Design Applicability to Improved SLMCPR, Power Distribution and R-Factor Methodologies", FLN-2001-016, September 24, 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 the Applicability of the GEXLI4 Correlation and Associated R-Factor Methodology for Calculating SLMCPR Values in Cores Containing GE14 Fuel", FLN-2001-017, October 1, 2001.

Comparison of Cooper Cycle 22 SLMCPR Value Table I summarizes the relevant input parameters and results of the SLMCPR determination for the Cooper Cycle 22 and 21 cores. The SLMCPR evaluations were performed using NRC approved methods and uncertaintiesll. These evaluations yield different calculated SLMCPR values because 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 Cooper Cycle 22 and Cycle 21 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: (I) 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.

(( ))

The uncontrolled bundle pin-by-pin power distributions were compared between the Cooper Cycle 22 bundles and the Cycle 21 bundles. Pin-by-pin power distributions are characterized in terms of R factors using the NRC approved methodology[2]. For the Cooper Cycle 22 limiting case analyzed at EOR-1K, (( )) the Cooper Cycle 22 bundles are flatter than the bundles used for the Cycle 21 SLMCPR analysis.

Summary

(( )) have been used to compare quantities that impact the calculated SLMCPR value. Based on these comparisons, the conclusion is reached that the Cooper Cycle 22 core/cycle has a flatter core MCPR

((GNF Proprietary Information)) page 1 of 5

((enclosed by double brackets))

Attachment Additional Information Regarding the October 4, 2002 Cycle Specific SLMCPR for Cooper Cycle 22 distribution (( ]J and flatter in-bundle power distributions (( )) than what was used to perform the Cycle 21 SLMCPR evaluation.

The calculated 1.09 Monte Carlo SLMCPR for Cooper Cycle 22 is consistent with what one would expect (( )) the 1.09 SLMCPR value is appropriate.

Based on all of the facts, observations and arguments presented above, it is concluded that the calculated SLMCPR value of 1.09 for the Cooper Cycle 22 core is appropriate. It is reasonable that this value is larger than the 1.08 value calculated for the previous cycle.

For single loop operations (SLO) the calculated safety limit MCPR for the limiting case is 1.11 as determined by specific calculations for Cooper Cycle 22.

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 GEl4 applications satisfy the conditions of the NRC SER111 have been addressed in Reference [4]. Other generically applicable questions related to application of the GEXLI4 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 Cooper Cycle 21 and 22 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.

Prepared by: Verified by:

/. Butrovich E. W. Gibbs echnical Project Manager Technical Project Manager Global Nuclear Fuel - Americas Global Nuclear Fuel - Americas

((GNF Proprietary Information)) page 2 of 5

((enclosed by double brackets))

Attachment Additional Information Regarding the October 4,2002 Cycle Specific SLMCPR for Cooper Cycle 22 Table 1 Comparison of the Cooper Cycle 21 and Cycle 22 SLMCPR QUANTITY, DESCRIPTION Cooper Cycle 21 Cooper Cycle 22 Number of Bundles in Core 548 548 Limiting Cycle Exposure Point EOR-1.2K EOR-1K Cycle Exposure at Limiting Point [MWd/STU] 8000 9975 Reload Fuel Type GE14 GE14 Latest Reload Batch Fraction [%] 21.9 23.4 Latest Reload Average Batch Weight % 3.79 3.95 Enrichment Batch Fraction for GE14 [%] 46.7 70.1 Batch Fraction for GE9B [%] 53.3 29.9 Core Average Weight % Enrichment 3.65 3.75 Core MCPR (for limiting rod pattern) 1.49 1.46

((I))

Power distribution uncertainty GETAB GETAB Non-power distribution uncertainty Revised Revised Calculated Safety Limit MCPR 1.08 1.09

((GNF Proprietary Information)) page 3 of 5

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Attachment Additional Information Regarding the October 4, 2002 Cycle Specific SLMCPR for Cooper Cycle 22 Figure 1 Reference Core Loading Pattern - Cycle 21 F F F F 'F *FI FIF F F F F F F F F F F F F F F F C A C C A C C A CI CIA CI F F F F F F C D D D D C D DIC D D D DC F F F F C F C A C B D B B DI B CIA C F C F F F FICIBICIDLCIBED[CTB BICIDIBICIDICIBTC =FF FIC F C D A D B B C DID C B B D A D C F C F FT CIDICFI DIAID BICICIDIB BIjDICC TB IDIAlDIC j00 F F A ID B C D 1B1 B A ID A B B A D A B B D C B D A F F F CID C BIAICA B C B B B BC B A C A B C D C F F F E C I B D B C D C D C D D C D C D C B D B D C E F F F AIC D C C D A AIC D B B D C A A D C C D C A F F F F C D B B D B A B D B A A B D B A B D B B D C F F F F C D B B D B A B D B A A B D B A B DOB B D C F F F F A C D C C D A A C DB B D CIA A D C C D CIA F F F E C D B DIBIC D C D C DOD C D C DIC B D BID CIE F FIF CF D C B A C A B C B B B B C B A C A B C D C F F F A D B C D B B A D A B B A D A B B D C B D A F F C D C D A D B C C D B B D C C B D A DC CF F C F C D A D B B C D D C B B D A D C F C F F F CIB CID C B DIC B B C D B CID C B C F F F F C F C A C B D B B D B C A C F C F F F F F C D D D D C D D C D D D D C F F F F F F C A C C A C A C C A CIFFF F F F F F F F F F F F F F F FIF F FIF F F 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 FUEL TYPE CYCLE LOADED NUMBER IN CORE A = GE9B-P8DWB350-1OGZ-80U-150-T 19 60 B = GE9B-P8DWB350-10GZ1-80U-150-T 19 100 C = GE14-P1OHNAB385-14GZ-1OOT-148-T 20 136 D = GE14-P10HNAB379-17GZ-100T-150-T-2472 21 120 E = GE9B-P8DWB348-11GZ-80M-150-T 18 4 F = GE9B-P8DWB350-10GZ-80U-150-T 18 128

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Attachment Additional Information Regarding the October 4, 2002 Cycle Specific SLMCPR for Cooper Cycle 22 Figure 2 Reference Core Loading Pattern - Cycle 22 52 B B B B B 8 B B BIB B B B 50 B B A B A A A A B A B B 48 B B A B B D A D C C D A D B B A B B 46 FJBjAjD Dl DIGI DIGIHIHGDIGIDID(DIAIBIF 44 A B D C G G C G C D D C G CIG G C D BIA 42 B BD IAIDI HICIDCIHICICJHICIDJCIHIDIAID B B 40 AID C D D C H C H C H H C H C H C D D C DIA 38 BDGB BIBIDIGIHJC[ H CHID ClH C H H C DH HTDIHICIDTCTCJDICIHID!HICTHIGD[ýBB GCDBH 36 B BID G C H D C C H C D D C H C C D H C G D B BI B A D G C D C H C D C H A A H C D C H C D C G D A B 34 32 B B A D G C H C H C B D HIH D B C H C H C G D A B B 30 B B D G C H C D C H D H C C H D H C D C H C G D B B 28 A A C H D C H C D A H C C C C H A D C H C D H C A A 26 A A C H D C H C D A H C C C C H A D C H C D H C A A 24 B B D G C H C D C H D H C C H D H C D C H C G D B B 22 B B A D G C H C H C B D H H D B C H C H C G D A B B 20 LB A D G C D C IH C D C H A A H IC D C H C D C G D A B 18 B B D G C H D C C HIC D D C H C C D H CIG D B B 16 B B D G H C H D H C D C C D C H D H C H G D B 14 JA D C D D C H C H C H H C H C HIC D D C D A 12 B B D A D H C D C H CIC H C D C H D A D B B 10 A B D C G G C G C D D C G C G G C D B A 8 F B A D D D G DG H H G D G D D D AB F 6 B B A B B D A D C C D A D B B A B B 4 B B A B A A A A B A B B 2 B B B B B3 B B B 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 FUEL TYPE CYCLE LOADED NUMBER IN CORE A = GE9B-P8DWB350-10GZ-80U-150-T 19 60 B = GE9B-P8DWB350-10GZ1-80U-150-T 19 100 C = GE14-P1OHNAB385-14GZ-10OT-148-T 20 136 D = GE14-P10HNAB379-17GZ-100T-150-T-2472 21 120 E

F = GE9B-P8DWB350-10GZ-80U-150-T 18 4 G = GE14-P1ODNAB398-16GZ-100T-150-T-2568 22 40 H = GE14-P1ODNAB393-17GZ-100T-150-T-2610 22 88

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