ML082670351
| ML082670351 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/17/2008 |
| From: | Beltz T Plant Licensing Branch III |
| To: | Rencheck M Indiana Michigan Power Co |
| beltz T, NRR/DORL/LPL3-1, 301-415-3049 | |
| References | |
| TAC MD7556 | |
| Download: ML082670351 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 17, 2008 Mr. Michael W. Rencheck Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 SUB.JECT:
DONALD C. COOK NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT TO RENEWED FACILITY OPERATING LICENSE REGARDING USE OF THE WESTINGHOUSE ASTRUM LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS METHODOLOGY (TAC NO. MD7556)
Dear Mr. Rencheck:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 306 to Renewed Facility Operating License No. DPR-58 for Unit 1 of the Donald C. Cook Nuclear Plant. The amendment changes the Technical Specifications (TS) in response to your application dated December 27,2007, as supplemented by letter dated July 14, 2008.
The amendment revises TS Section 3.4.1, "RCS [Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," and modifies the analytical method used for determining core operating limits for a large break loss-of-coolant accident (LBLOCA) specified in TS 5.6.5, "Core Operating Limits Report (COLR)." The proposed amendment also requested NRC approval of a new LBLOCA analysis using a plant specific adaptation of Westinghouse topical report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."
A copy of the associated safety evaluation is enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
'ncereIY,
~ It
'1:
/
Terry IS... Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-315
Enclosures:
- 1. Amendment No. 306 to DPR-58
- 2. Safety Evaluation cc w/encls: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 306 License No. DPR-58
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated December 27,2007, as supplemented by letter dated July 14, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-58 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and Appendix B, as revised through Amendment No. 306, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
~~~
Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Operating License and Appendix A Date of Issuance: October 17, 2008
ATTACHMENT TO LICENSE AMENDMENT NO. 306 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Replace the following page of Renewed Facility Operating License No. DPR-58 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
REMOVE INSERT - 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.4.1-1 3.4.1-1 3.4.1-2 3.4.1-2 5.6.3 5.6.3
- 3 and radiation monitoring equipment calibration, and as fission detectors in amounts as required.
(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; and (5)
Pursuant to the Act and 10 CFR 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal in accordance with the conditions specified therein.
(2) Technical Specifications The Technical Specifications contained in Appendix A and Appendix B, as revised through Amendment No. 306, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Less Than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined in Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operating license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than found loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.
(4) Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated December 12,1977, July31, 1979, January 10,1981, February 7, 1983, November 22,1983, December 23,1983, March 16, 1984, August 27,1985 Renewed License No. DPR-58 Amendment 1\\10. 1 through 304, ~, 306
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 APPLICABILITY:
ACTIONS RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:
- a.
Pressurizer pressure is greater than or equal to the limit specified in the COLR;
- b.
RCS average temperature is less than or equal to the limit specified in the COLR; and
- c.
RCS total flow rate is greater than or equal to the limit specified in the COLR. The minimum RCS total flow rate shall be
~ 354,000 gpm.
MODE 1.
N0 TE-------------------------------------------
Pressurizer pressure limit does not apply during:
- a.
THERMAL POWER ramp> 5% RTP per minute; or
- b.
THERMAL POWER step> 10% RTP.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB parameters not within limits.
A.1 Restore RCS DNB parameter(s) to within limit.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. Required Action and associated Completion Time not met.
B.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Cook Nuclear Plant Unit 1 3.4.1-1 Amendment No. ~,306
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to the limit specified in the COLR.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.2 Verify RCS average temperature is less than or equal to the limit specified in the COLR.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.3 Verify RCS total flow rate is ~ 354,000 gpm and greater than or equal to the limit specified in the COLR.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.4
NOTE-----------------------------
Not required to be performed until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after
~ 90% RTP.
Verify by precision heat balance that RCS total flow rate is ~ 354,000 gpm and greater than or equal to 24 months the limit specified in the COLR.
Cook Nuclear Plant Unit 1 3.4.1-2 Amendment No. ~, 306
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 5.
LCO 3.1.6, "Control Bank Insertion Limits";
- 6. LCO 3.2.1, "Heat Flux Hot Channel Factor (Fa(Z))";
- 7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (F~H )";
- 8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";
- 9.
LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Functions 6 and 7 (Overtemperature ~T and Overpower ~T, respectively)
Allowable Value parameter values;
- 10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," (Westinghouse Proprietary);
- 2.
WCAP-8385, "Power Distribution Control and Load Following Procedures - Topical Report," (Westinghouse Proprietary);
- 3.
WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/Fa Surveillance Technical Specification," (Westinghouse Proprietary);
- 4.
Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," (Westinghouse Proprietary);
- 5.
WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," (Westinghouse Proprietary);
- 6.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower""'T and Thermal Overtemperature""'T Trip Functions," (Westinghouse Proprietary); and
- 7.
WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary).
Cook Nuclear Plant Unit 1 5.6-3 Amendment No. ~, ~, 306
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEf\\lDMENT NO. 306 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-315
1.0 INTRODUCTION
By application dated December 27,2007 (Agencywide Document and Access Management System (ADAMS) Accession No. ML080090268), as supplemented by letter dated July 14, 2008 (ADAMS Accession No. ML082040584), the Indiana Michigan Power Company (the licensee) requested a license amendment to revise the Technical Specifications (TS) for the Donald C.
Cook Nuclear Plant, Unit 1 (DCCNP-1). The supplemental letter of July 14, 2008, provided additional information to clarify the application; did not expand the scope of the application as originally noticed; and did not change the Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 29,2008 (73 FR 5223).
The proposed amendment would increase the required minimum Reactor Coolant System (RCS) flow rate specified in TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," and modify the analytical method using for determining core operating limits for a large break loss-of-coolant accident (LBLOCA) specified in TS 5.6.5, "Core Operating Limits Report (COLR)" for DCCf\\lP-1.
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR) Part 50 includes the NRC's requirement that TSs shall be included by applicants for a license authorizing operation of a production or utilization facility. 10 CFR 50.36 (d) requires that TSs include items in five specific categories related to station operation. These categories are (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operations (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TS. This amendment is within categories (2), (3), and (5).
Title 10 of the Code of Federal Regulations (10 CFR), Paragraph 50.46(aX1) identifies the calculation methodology requirements for nuclear power plant LOCA methodologies. 10 CFR 50.46(c) identifies the types of processes which are required to assure that LOCA analyses performed for a given plant actually represent the plant. Section 50.46(a)(3)(i) and (ii) specify criteria to be applied and actions to be taken when significant changes or errors in parts of the plant-specific LOCA methodology are found to have accumulated.
Enclosure
- 2 In accordance with 10 CFR 50.46(a)(3)(ii), the licensee submitted a new DCCNP-1 LBLOCA analysis to the NRC due to an accumulation of changes and errors requiring a scheduled reanalysis. The new LBLOCA analysis uses a higher RCS flow rate value to provide a greater margin to DNB limits and requires a change to the minimum flow rate speciFied in LCO 3.4.1.c.
and surveillance requirements 3.4.1.3 and 3.4.1.4. The new LBLOCA analysis is referenced in the COLR, which is one of the administrative controls to assure operation of the facility in a safe manner and is referenced in the DCCNP-1 TSs, Section 5.6.5.
3.0 TECHNICAL EVALUATION
3.1 Description of Proposed Changes The licensee requested approval to use a plant-specific version of the enhanced adaptation of the WCAP-16009-P-A methodology containing an enhanced treatment of the downcomer. The licensee also requested approval of a DCCNP-1 inclusion of this enhanced version of the WCAP-16009-P-A in the DCCNP-1 TSs.
The licensee proposed to revise TS 5.6.5 by replacing the existing LBLOCA methodology specified in TS 5.6.5.b.4 with the following: "Plant-specific adaptation of WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."
The licensee also proposed to revise TS Limiting Condition for Operation 3.4.1.c and Surveillance Requirements 3.4.1.3 and 3.4.14 to reflect a higher minimum RCS flow rate of 354,000 gallons per minute.
3.2 Evaluation of Proposed Changes The current DCCNP-1 LBLOCA analysis of record was performed in calendar year 2000 using the BASH evaluation model methodology documented in WCAP-10266-P-A, "the 1981 Version of Westinghouse ECCS [emergency core cooling system] Evaluation Model Using the BASH Code." A new LBLOCA analysis for DCCNP-1 was provided to the I\\IRC due to an accumulation of changes and errors requiring a schedule reanalysis in accordance with 10 CFR 50.46(a)(3)(ii).
The ASTRUM methodology received NRC approval for referencing in licensing applications by the safety evaluation in WCAP-16009-P-A.
The licensee provided the following statement to confirm that the reference generically-approved ASTRUM LOCA analysis methodology applies specifically to DCCNP-1: "I&M and Westinghouse have ongoing processes that assure that the ranges and values of the input parameters for the DCCNP-1 LBLOCA analyses bound the ranges and values of the as operated plant parameters."
To address the effects, if any, of the mixed core on peak cladding temperature and oxidation for the pre-resident fuel, the licensee states that both the pre-resident fuel and the fresh fuel are of the same design and therefore there are no mixed core effects to consider.
The results of the DCCNP-1 LBLOCA analyses are provided in Table 1.
- 3 Table 1 D.C. Cook Unit 1: Large Break LOCA Analysis Results Using ASTRUM Parameter Cladding Material Peak Cladding Temperature I Maximum Local Oxidation Percentage Core-Wide H2 Generation (Oxidation)
ASTRUM LBLOCA Analyses Results ZIRLO 2128 of 11.1 %
0.40 %
10 CFR 50.46 Limits (Cylindrical ) Zircaloy or ZIRLO I
I s 2200 °F
- 5 17 %
- 5 1.0 %
The burnup effects on oxidation for the pre-resident fuel were addressed by providing the expected maximum values of LOCA oxidation at beginning of life (-10%) and end of life (- 0%).
The NRC staff agrees with the conclusion that regardless of what time in the "fuel life" the postulated LBLOCA were to occur, the anticipated total LOCA oxidation would be less than 17%.
To verify that the treatment of the vessel wall (radial noding, etc.) during reflood remains as historically approved in addressing the issue of downcomer boiling, the licensee indicated that the number of downcomer nodes in the DCCNP-1 analyses of downcomer boiling had been increased versus the previous generic noding. The NRC staff finds that the DCCNP-1 downcomer noding is consistent with downcomer noding for other similar Westinghouse plants that addressed the issue and, therefore, is acceptable.
Limits on RCS pressure, temperature, and flow rate ensure that the minimum DNB ratio will be met for each of the safety analysis transients analyzed. RCS flow rate normally remains constant during an operational fuel cycle with all reactor coolant pumps running. The minimum RCS flow limit corresponds to that assumed for DNB analyses. Flow rate indications are averaged to determine a value for comparison to the limit. The new LBLOCA analysis uses a higher RCS flow rate value that provides a greater margin to DNB limits and requires a change to the minimum flow rate specified in TS 3.4.1.
3.3 Summary The NRC staff finds the ASTRUM LBLOCA methodology and associated amended TSs are acceptable for licensing application to DCCNP-1. The specific DCCNP-1 LBLOCA analyses performed using ASTRUM and the results of those analyses show compliance with the provisions of 10 CFR 50.46. The NRC staff finds the TS changes to be acceptable, as they accurately reflect the changes found acceptable in this safety evaluation.
- 4
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (73 FR 5223). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that 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 issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Frank Orr, NRR Date: October 17, 2008
Mr. Michael W. Rencheck October 17, 2008 Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT TO RENEWED FACILITY OPERATING LICENSE REGARDING USE OF THE WESTINGHOUSE ASTRUM LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS METHODOLOGY (TAC NO. MD7556)
Dear Mr. Rencheck:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 306 to Renewed Facility Operating License No. DPR-58 for Unit 1 of the Donald C. Cook Nuclear Plant. The amendment changes the Technical Specifications (TS) in response to your application dated December 27, 2007, as supplemented by letter dated July 14, 2008.
The amendment revises TS Section 3.4.1, "RCS [Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (D/\\/B) Limits," and modifies the analytical method used for determining core operating limits for a large break loss-of-coolant accident (LBLOCA) specified in TS 5.6.5, "Core Operating Limits Report (COLR)." The proposed amendment also requested NRC approval of a new LBLOCA analysis using a plant specific adaptation of Westinghouse topical report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."
A copy of the associated safety evaluation is enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Sincerely, lRAI Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-315
Enclosures:
- 1. Amendment No. 306 to DPR-58
- 2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION PUBLIC RidsRgn3MailCenter Resource G. Hill(2)
J. Giessner, Rill RidsAcrsAcnw_MailCTR Resource RidsOgcRp Resource RidsNrrLATHarris Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl3-1 Resource LPL3-1 R/F RidsNrrPMDCCook Resource F. Orr, NRR RidsNrrDirsltsb Resource Package Accession Number: ML082670379 Amendment Accession Number' ML082670351 TS' ML082670393 OFFICE LPL3-1/PM LPL3-1/PM LPL3-1/LA SRXB/BC OGC LPL3-1/BC NAME TBeltz;JJ13 PTam THarris~fw GCranston*
RHolmes NLO LJames~\\
DATE 10/9/08 09/26/08 10/'6/08 07/29/08 10/08/08 101-P8 a
- SE transmitted by memo of 07/29/2008 OFFICIAL RECORD COpy