ML110730783
| ML110730783 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/31/2011 |
| From: | Tam P Plant Licensing Branch III |
| To: | Weber L Indiana Michigan Power Co |
| Tam P | |
| References | |
| TAC ME1017 | |
| Download: ML110730783 (16) | |
Text
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- Mr: Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 SUB~IECT: DONALD C. COOK NUCLEAR PLANT, UNIT 2 (CNP-2) - ISSUANCE OF AMENDMENT TO ADOPT A NEW LARGE-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS (TAC NO. ME1017)
Dear Mr. Weber:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 297 to Renewed Facility Operating License No. DPR-74 for CNP-2. The amendment is in response to your application dated March 19, 2009, as supplemented by letters dated November 20, 2009, February 24, March 11, and March 25, 2011. Specifically it approves a new analysis of a large break loss-of-coolant accident performed using a plant-specific adaptation of the NRC-approved methodology set forth in Westinghouse Topical Report WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)." It also revises the Technical Specifications at Sections 3.4.1, "[Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits," 3.5.2, "[Emergency Core Cooling System] - Operating," and 5.6.5, "Core Operating Limits Report (COLR)."
A copy of the associated safety evaluation is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
~~kManager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-316
Enclosures:
- 1. Amendment No. 297 to DPR-74
- 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. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 297 License No. DPR-74
- 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 March 19, 2009, as supplemented by letters dated November 20, 2009, February 24, March 11, and March 25, 2011, 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; S. 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. 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-74 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix S, as revised through Amendment No. 297, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
-2
- 3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
As stated Date of Issuance: March 31, 2011
ATTACHMENT TO LICENSE AMENDMENT NO. 297 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Replace the following page of Renewed Facility Operating License No. DPR-74 with the revised page attached. 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 3.5.2-1 3.5.2-1 5.6-3 5.6-3 5.6-4 5.6-4
-3 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 Parts 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, Sections 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 3468 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to the renewed operating license.
The preoperational tests, startup and other items identified in Attachment 1 to this renewed operating license shall be completed. Attachment 1 is an integral part of this renewed operating license.
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 297, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Additional Conditions (a) Deleted by Amendment No. 76 (b) Deleted by Amendment No.2 (c) Leak Testing of Emergency Core Cooling System Valves Indiana Michigan Power Company shall prior to completion of the first inservice testing interval leak test each of the two valves in series in the Renewed License No. DPR-74 Amendment No. 1 through 297
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 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;
- 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.
APPLICABILITY: MODE 1.
NOT E-------------------------------------------
Pressurizer pressure limit does not apply during:
- a. THERMAL POWER ramp> 5% RTP per minute; or
- b. THERMAL POWER step> 10% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> parameters not within parameter(s) to within limit.
limits.
B. Required Action and 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 /> associated Completion Time not met.
Cook Nuclear Plant Unit 2 3.4.1-1 Amendment No. ~, 297
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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to the limit specified in the COLR.
SR 3.4.1.2 Verify RCS average temperature is less than or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> equal to the limit specified in the COLR.
SR 3.4.1.3 Verify RCS total flow rate is;:: 354,000 gpm and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> greater than or equal to the limit specified in the COLR.
SR 3.4.1.4 -------------------------------NOT E----------------------------
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 24 months rate is;:: 354,000 gpm and greater than or equal to the limit speci"fied in the COLR.
Cook Nuclear Plant Unit 2 3.4.1-2 Amendment No. ~, 297
ECCS - Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.
B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available.
Cook Nuclear Plant Unit 2 Amendment 1\10. ~, 297
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 (FdZ))";
- 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 l::.T and Overpower l::. 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 (approved by Amendment 297, dated March 31, 2011) of WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," Revision 0 (Westinghouse Proprietary), approved by letter from H. N. Berkow, NRC, to J. A.
Gresham, Westinghouse Electric Company, dated November 5, 2004;
- 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.T and Thermal Overtemperature t.T Trip Functions," (Westinghouse Proprietary); and Cook Nuclear Plant Unit 2 5.6-3 Amendment No. 299, ~, 297
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 7. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary).
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:
- a. The scope of inspections performed on each SG,
- b. Active degradation mechanisms found,
- c. Nondestructive examination techniques utilized for each degradation mechanism,
- d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e. Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f. Total number and percentage of tubes plugged to date, and
- g. The results of condition monitoring, including the results of tube pulls and in situ testing.
Cook Nuclear Plant Unit 2 5.6-4 Amendment No. ~, 2W, ~, ~, 297
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 297 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT. UNIT 2 (CNP-2)
DOCKET NO. 50-316
1.0 INTRODUCTION
By application dated March 19, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090930453), as supplemented by letters dated November 20,2009 (ADAMS Accession No. ML093360524), February 24,2011 (ADAMS Accession No. ML110680217), March 11, 2011 (ADAMS Accession No. ML110810103), and March 25, 2011, Indiana Michigan Power Company (I&M, the licensee) submitted a License Amendment Request (LAR) for CNP-2. The licensee requested approval of a new analysis of a large-break loss-of coolant accident (LBLOCA), which was performed using a plant-specific adaptation of the NRC-approved methodology set forth in Westinghouse Topical Report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)." The licensee also requested approval of changes to Technical Specifications Sections 3.4.1, "[Reactor Coolant System] RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," 3.5.2, "[Emergency Core Cooling System] ECCS - Operating," and 5.6.5, "Core Operating Limits Report (COLR)."
The supplemental letters cited in the above paragraph contain clarifying information, do not change the scope of the LAR, and do not change the Nuclear Regulatory Commission (NRC) staff's initial proposed finding of no significant hazards consideration published in the Federal Register on Aug ust 11, 2009 (74 FR 40238).
2.0 REGULATORY EVALUATION
The NRC staff considered the following regulatory requirements and guidance in its review of the proposed LAR.
Section 50.36, of Title 10 of the Code of Federal Regulations (10 CFR) "Technical specifications," requires that technical specifications of a nuclear plant include the following items: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, (5) administrative controls, (6) decommissioning, (7) initial notification, and (8) written reports.
Enclosure
Section 50.46 of 10 CFR, "Acceptance criteria for emergency core cooling systems for light water nuclear power reactors," requires each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an ECCS that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in 10 CFR 50.46(b), including peak cladding temperature, maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long-term cooling.
3.0 TECHNICAL EVALUATION
The NRC staffs technical review of the proposed LAR includes: (1) the proposed modification of TS 3.4.1 "RCS Pressure, Temperature, and Flow Departure from Nuclear Boiling (DNB)
Limits," TS 3.5.2, "ECCS - Operating," and TS 5.6.5, "Core Operating Limits Report (COLR),"
and (2) adopting a new analysis of an LBLOCA for CNP-2.
3.1 TS 3.4.1- RCS Pressure. Temperature, and Flow Departure from Nuclear Boiling Limits The licensee proposed to revise the minimum RCS total flow rate specified in TS Limiting Condition for Operation (LCO) 3.4.1.c and TS Surveillance Requirements 3.4.1.3 and 3.4.1.4 from 366,400 gallons per minute (gpm) to 354,000 gpm.
The NRC staff reviewed the proposed change including: (1) the description of the proposed term of thermal design flow value in relation to RCS total flow from the current value of 366,400 gpm, and (2) the real minimum RCS flow used in the LBLOCA analysis. In the November 20, 2010, response to the NRC staff request for additional information the licensee stated that the total flow rate of 366,400 gpm is the minimum measured flow, which includes allowances for flow measurement uncertainty and is equivalent to the thermal design flow of 354,000 gpm plus an allowance for measurement uncertainty. As revised, LCO 3.4.1.c and SR 3.4.1.4 will continue to say that RCS total flow rate is "greater than or equal to the limit specified in the COLR" (i.e., greater than or equal to 366,400 gpm = 354,000 plus measurement uncertainty).
The NRC staff concludes that this proposed change is acceptable because it is a clarification of the existing definition of the total flow rate, and does not change the actual numerical value of the required RCS flow rate.
3.2 TS 3.5.2 - "ECCS - Operating" The proposed change is to delete: (1) Condition D, "One or more Safety Injection (SI) System cross tie valves closed," and (2) the reference to Condition D in Conditions A and C from the TS 3.5.2 Actions.
The licensee's application states that the current TS 3.5.2 Actions include a Condition D that allows the unit to be in Mode 1, 2, or 3 for an unlimited amount of time if an SI system cross-tie valve is closed, provided that thermal power is reduced to less than or equal to a specified value.
It further states that this allowance is justified by the current LBLOCA and small-break loss-of-coolant accident (SBLOCA) analyses. In its November 20,2009, letter the licensee stated that it has not identified any occurrence in CNP-2's operation history in which Condition D was used to allow operation at reduced power with an SI system cross-tie valve closed. Thus, the licensee did not analyze an operational scenario using Condition D in the proposed new
- 3 LBLOCA analysis. Therefore, the licensee proposed that Condition D be deleted from the TS 3.5.2 Actions, and reference to Condition D be deleted from Condition A and Condition C.
The NRC staff reviewed the proposed change, and agrees that the new LBLOCA analysis proposed by this amendment does not address a condition in which an SI cross-tie valve is closed. Therefore, the allowance provided by TS 3.5.2 Condition D is no long a valid condition and should be deleted. The licensee's proposed change is acceptable.
3.3 TS 5.6.5 - Core Operating Limits Report (COLR)
The licensee proposed to revise TS 5.6.5.b.4, replacing the LBLOCA methodology specified as topical report WCAP-10266-P-A, "The 1981 Version of Westinghouse Evaluation Mode Using BASH Code," with a new LBLOCA methodology described as plant-specific adaptation of WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)."
WCAP-16009-P-A describes a realistic (Le., best-estimate) ECCS evaluation model for demonstratil1g plant compliance with 10 CFR 50.46 for postulated plant-specific LBLOCA transients. WCAP-16009-P-A uses a statistical approach in developing the peak cladding temperature (PCT), local maximum oxidation (LMO), and core-wide oxidation (CWO) results at the 95 th percentile. The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimSlte of the 95th percentile of the PCT, LMO, and CWO parameters with 95 percent confidence level. These parameters are needed to satisfy 10 CFR 50.46 criteria.
The licensee applied a plant-specific adaptation of the ASTRUM methodology for the new CNP 2 analysis to better model the downcomer region by increasing the number of the circumferential noding stacks from four to twelve. The licensee stated that this finer nodalization has been assessed against experimental data and provided good overall agreement with the experimental data (letter, J. N. Jensen of the licensee to NRC, December 27, 2007; ADAMS Accession No. ML080090267). The validation package (described in paragraph below) provides information to support applicability of the plant-specific adaptation of WCAP-16009-P-A to the CNP-2 LBLOCA analysis. The detailed radial noding of the vessel wall remains unchanged from the approved ASTRUM LBLOCA evaluation model and does not change the historically approved method for addressing downcomer boiling during reflood.
The licensee presented results of the plant-specific ASTRUM analysis in Table 2 of Enclosure 2 of the March 19, 2009 application. Among other results, Table 2 shows that the calculated peak cladding temperature (PCT) to be 2107 degrees Fahrenheit, compared to the regulatory limit of 2200 degrees Fahrenheit in 10 CFR 50.46.
On August 11,2010, the NRC staff performed an audit of the validation package to support the application of a 12-downcomer-channel stack model to CNP-2, since the this validation package (WCOBRAITRAC Validation with Revised Downcomer Noding for D.C. Cook Units 1 and 2, dated November 2007) was not previously reviewed by the NRC staff for the deviation from a generic approval of the methodology (NRC letter, H. N. Berkow to J. A. Gresham, approval of WCAP-16009-P-A, Revision 0, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," November 4,2004;
-4 ADAMS Accession No. ML043100073). Among the documents the NRC staff audited was a letter (NF-AE-10-92 W/AEP2-1601) from T. J. Kitchen of Westinghouse to Andy Hawk of the licensee, "Amendment 8 to Fuel Fabrication Contract 1500011, LBLOCA ASTRUM Reanalysis, Supplemental Information to Support Informal NRC Audit of 12 Downcomer Channel Model Validation and CAW-10-2929," dated August 27,2010. Subsequent to the audit, the NRC staff held a number of conference calls with the licensee to obtain clarification and to resolve open issues relating to the CNP-2 LBLOCA analysis with respect to: (1) 12 downcomer channel stacks versus 4, and the licensee's sensitivity study on the impact of the lateral K-factor in the downcomer on the PCT; and (2) post-LOCA boric acid precipitation. On February 24, 2011, the licensee provided supplemental information to resolve issues raised by the NRC staff during and after the audit. Based on its review of the February 24, 2011, letter the NRC staff concludes that:
(1) The review delineated in this safety evaluation is only applicable to the plant-specific CNP-2 LBLOCA, but not for application to other plants without further NRC staff review.
(2) The issue of boric acid precipitation analysis associated with the CNP-2 power level of 3479.8 MWt will be reviewed separately as a future action as result of the licensee's commitment to re-analyze boric acid precipitation made in its March 11, 2011, letter.
Separation and delayed resolution of this boric acid precipitation issue is acceptable to the staff as the resolution of this issue is not necessary for this licensing action regarding LBLOCA analysis, and the NRC staff has performed a confirmatory analysis that demonstrated that sufficient margin exists. However, a future licensee-specific analysis is necessary to demonstrate plant-specific margin and to support timely operator actions.
(3) The reasoning for not using the Idelchick Handbook recommended expression for pressure loss coefficients is acceptable because its lateral loss in the down comer is higher than that calculated from WCOBRAITRAC for a 4-loop pressurized-water reactor and a lower lateral loss in the downcomer will tend to exaggerate the ECCS bypass and liquid sweepout phenomena during a large-break LOCA. which in-turn results in reduced vessel fluid inventory; (4) The NRC staff had made an observation on the licensee's conclusion regarding the CNP-2 sensitivity study on the impact of the lateral k-factor values on PCT during downcomer boiling following an LBLOCA. The followup of the resolution of this observation will occur separately and will be tracked by TAC ME5865. The NRC staff believes that there is sufficient margin in the PCT limiting case such that this lateral k factor issue does not affect the staff's safety evaluation for the current LBLOCA amendment.
In summary, the NRC staff has reviewed the proposed change to a previously approved methodology for LBLOCA analysis with a minor modification on the circumferential noding stacks in the downcomer region for CNP-2, and found the plant-specific adaptation of the ASTRUM methodology is acceptable for LBLOCA analysis for the CNP-2 application because the results shown in Table 2, Enclosure 2 of the March 19,2009, application, meet acceptable criteria prescribed in 10 CFR 50.46 based on CNP-2 plant parameter assumptions.
The NRC staff also found acceptable the licensee's proposed replacement of LBLOCA methodology specified in TS Section 5.6.5.b.4 as WCAP-10266-P-A, "The 1981 Version of
- 5 Westinghouse Evaluation Mode Using BASH Code" with a plant-specific adaptation of WCAP 16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method {ASTRUM)."
3.4 Conclusion of Technical Evaluation The NRC staff has reviewed the proposed LAR and concluded that the request is acceptable:
(1) to use thermal design flow of 354,000 gpm in LCO 3.4.1.c, and SR 3.4.1.3 and 3.4.1.4; (2) to delete Condition D; and (3) to use a plant-specific adaptation of WCAP-16009-P-A for LBLOCA analysis.
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 a requirement with respect to the use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no Significant increase in the amounts 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 considerations, and there has been no public comment on such finding August 11,2009 (74 FR 40238). 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 Commission has concluded, based on the considerations discussed above that there is reasonable assurance that: (1) 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Tai L. Huang, NRR Leonard Ward, NRR Date: March 31, 2011
Mr. Lawrence J. Weber March 31, 2011 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 2 (CNP-2) - ISSUANCE OF AMENDMENT TO ADOPT A NEW LARGE-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS (TAC NO. ME1017)
Dear Mr. Weber:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 297 to Renewed Facility Operating License No. DPR-74 for CNP-2. The amendment is in response to your application dated March 19, 2009, as supplemented by letters dated November 20,2009, February 24, March 11, and March 25,2011. Specifically it approves a new analysis of a large break loss-of-coolant accident performed using a plant-specific adaptation of the NRC-approved methodology set forth in Westinghouse Topical Report WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)." It also revises the Technical Specifications at Sections 3.4.1, "[Reactor Coolant System] Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits," 3.5.2, "[Emergency Core Cooling System] - Operating," and 5.6.5, "Core Operating Limits Report (COLR),"
A copy of the associated safety evaluation is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, IRAJ Peter S. Tam, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-316
Enclosures:
- 1. Amendment No. 297 to DPR-74
- 2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION PUBLIC RidsNrrDorlDpr Resource RidsNrrDssSnpb Resource LPL3-1 R/F RidsOgcRp Resource RidsNrrDirsltsb Resource RidsNrrDorlLpl3-1 Resource RidsAcrsAcnw_MailCTR Resource RidsRgn3MailCenter Resource RidsNrrLABTully Resource RidsNrrDssSrxb Resource THuang. NRR RidsNrrPMDCCook Resource LWard. NRR MHamm. NRR ADAMS A ccession No.: ML110730783 *B!ye-mal.,
OFFICE LPL3-1/PM LPL3-1/LA SRXB/BC SNPB/BC ITSB/BC OGC LPL3-1/BC NAME PTam BTully SMiranda AMendiola MHamm for DRoth RPascarelii for AUlses REllioU*
DATE 3/29/11 3/16/11 3/16/11 3/16/11 3/31/11 3/31/11 3/31/11 OFFICIAL RECORDS COPY