ML063260317
| ML063260317 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 11/29/2006 |
| From: | Chandu Patel NRC/NRR/ADRO/DORL/LPLII-2 |
| To: | Walt T Carolina Power & Light Co |
| Patel C, NRR/DORL/LPL2-2, 415-3025 | |
| References | |
| TAC MD1358 | |
| Download: ML063260317 (13) | |
Text
November 29, 2006 Mr. Thomas D. Walt, Vice President Carolina Power & Light Company H. B. Robinson Steam Electric Plant Unit No. 2 3581 West Entrance Road Hartsville, South Carolina 29550
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 ISSUANCE OF AMENDMENT REGARDING CORE OPERATING LIMIT REPORT REFERENCES (TAC NO. MD1358)
Dear Mr. Walt:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 211 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). This amendment is in response to your application dated April 11, 2006, as supplemented by letter dated November 9, 2006.
The amendment modifies Technical Specification (TS) 5.6.5, Core Operating Limits Report (COLR), to add a U.S. Nuclear Regulatory Commission approved topical report to the listing of analytical methods in TS 5.6.5.b. This change will allow for the use of the S-RELAP5 thermal-hydraulic analysis code for the non-loss-of-coolant accident analyses at HBRSEP2.
A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Chandu P. Patel, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261
Enclosures:
- 1. Amendment No. 211 to DPR-23
- 2. Safety Evaluation cc w/encls: See next page
Mr. T. D. Walt H. B. Robinson Steam Electric Plant, Carolina Power & Light Company Unit No. 2 cc:
David T. Conley Associate General Counsel II - Legal Department Progress Energy Service Company, LLC Post Office Box 1551 Raleigh, North Carolina 27602-1551 Ms. Margaret A. Force Assistant Attorney General State of North Carolina Post Office Box 629 Raleigh, North Carolina 27602 U. S. Nuclear Regulatory Commission Resident Inspectors Office H. B. Robinson Steam Electric Plant 2112 Old Camden Road Hartsville, South Carolina 29550 Mr. Dan Stoddard Plant General Manager H. B. Robinson Steam Electric Plant, Unit No. 2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550 Mr. William G. Noll Director of Site Operations H. B. Robinson Steam Electric Plant, Unit No. 2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550 Public Service Commission State of South Carolina Post Office Drawer 11649 Columbia, South Carolina 29211 J. F. Lucas Manager - Support Services - Nuclear H. B. Robinson Steam Electric Plant, Unit No. 2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550 Mr. C. T. Baucom Supervisor, Licensing/Regulatory Programs H. B. Robinson Steam Electric Plant, Unit No. 2 Carolina Power & Light Company 3581 West Entrance Road Hartsville, South Carolina 29550 Ms. Beverly Hall, Section Chief N.C. Department of Environment and Natural Resources Division of Radiation Protection 3825 Barrett Dr.
Raleigh, North Carolina 27609-7721 Mr. Robert P. Gruber Executive Director Public Staff - NCUC 4326 Mail Service Center Raleigh, North Carolina 27699-4326 Mr. Henry H. Porter, Assistant Director South Carolina Department of Health Bureau of Land & Waste Management 2600 Bull Street Columbia, South Carolina 29201 Mr. Chris L. Burton Manager Performance Evaluation and Regulatory Affairs PEB 7 Progress Energy Post Office Box 1551 Raleigh, North Carolina 27602-1551 Mr. John H. ONeill, Jr.
Shaw, Pittman, Potts, & Trowbridge 2300 N Street NW.
Washington, DC 20037-1128
Package:ML063420085 Amendment: ML063260317 TechPage:ML063410470 NRR-058 OFFICE LPL2-2/PM LPL2-2/LA SPWB/BC OGC LPL2-2/(A)BC NAME CPatel CSola JNakoski as signed on JMartin MC for DPickett DATE 11/28/06 11/28/06 11/10/06 11/2806 11/29/06
CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 211 Renewed License No. DPR-23 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Carolina Power & Light Company (the licensee), dated April 11, 2006, as supplemented by letter dated November 9, 2006, 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.
Accordingly, the license is hereby amended and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is revised to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 211, are hereby incorporated in the license. Carolina Power &
Light Company shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/ Margaret Chernoff for Douglas V. Pickett, Acting Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. DPR-23 and the Technical Specifications Date of Issuance: November 29, 2006
ATTACHMENT TO LICENSE AMENDMENT NO. 211 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace page 3 of Renewed Facility Operating License No. DPR-23 with the attached page 3.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.
Remove Pages Insert Pages 5.0-30 5.0-30
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 211 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261
1.0 INTRODUCTION
By letter dated April 11, 2006 (Reference 1), as supplemented by letter dated November 9, 2006 (Reference 2), the Carolina Power & Light Company (CP&L, licensee) submitted a request for changes to the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2), Technical Specifications (TSs). The requested changes would add topical reports EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors,"
(Reference 3), and EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," as references to TS 5.6.5, Core Operating Limits Report (COLR), to allow the licensee to update the methodologies that are used for safety analyses for HBRSEP2. By letter dated November 9, 2006, the licensee withdrew its proposal to add the topical report EMF-2328(P)(A), to the COLR methodologies list.
The proposed change, as revised, would permit the use of the S-RELAP5 thermal-hydraulic analysis code for the non-loss-of-coolant accident (non-LOCA) analyses in the HBRSEP2 Chapter 15 licensing basis of the Updated Final Safety Analysis Report (UFSAR).
The supplemental letter dated November 9, 2006, provided additional information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial application.
2.0 REGULATORY EVALUATION
Topical Report EMF-2310(P)(A) (Reference 3) pertains to non-LOCA accident and transient analyses that are part of the HBRSEP2 licensing basis. The regulatory bases for these analyses are found in the General Design Criteria (GDC) (Reference 8). The GDCs that pertain to each of the analyses are listed in the Standard Review Plan (SRP) (Reference 4).
Evaluation models of LOCA events are defined in Section 50.46 of Title 10 of the Code of Federal Regulations (10 CFR 50.46). This definition, which can also be applied to non-LOCA analyses, states that:
An evaluation model is the calculational framework for evaluating the behavior of the reactor system during a postulated loss-of-coolant accident (LOCA). It includes one or more computer programs and all other information necessary for application of the calculational framework to a specific LOCA, such as mathematical models used, assumptions included in the programs, procedure for treating the program input and output information, specification of those portions of analysis not included in computer programs, values of parameters, and all other information necessary to specify the calculational procedure.
Section II of Appendix K to 10 CFR Part 50, also written for LOCA analyses, and considered to be applicable to non-LOCA analyses, contains the documentation requirements for evaluation models. It states, as follows:
- 1. a.
A description of each evaluation model shall be furnished. The description shall be sufficiently complete to permit technical review of the analytical approach including the equations used, their approximations in difference form, the assumptions made, and the values of all parameters or the procedure for their selection, as for example, in accordance with a specified physical law or empirical correlation.
b.
A complete listing of each computer program, in the same form as used in the evaluation model, must be furnished to the Nuclear Regulatory Commission upon request.
2.
For each computer program, solution convergence shall be demonstrated by studies of system modeling or noding and calculational time steps.
- 3.
Appropriate sensitivity studies shall be performed for each evaluation model, to evaluate the effect on the calculated results of variations in noding, phenomena assumed in the calculation to predominate, including pump operation or locking, and values of parameters over their applicable ranges. For items to which results are shown to be sensitive, the choices made shall be justified.
4.
To the extent practicable, predictions of the evaluation model, or portions thereof, shall be compared with applicable experimental information.
5.
General Standards for Acceptability Elements of evaluation models reviewed will include technical adequacy of the calculational methods, including: For models covered by § 50.46(a)(1)(ii), compliance with required features of section I of this appendix K; and, for models covered by § 50.46(a)(1)(i), assurance of a high level of probability that the performance criteria of § 50.46(b) would not be exceeded.
Section III of Appendix B to 10 CFR Part 50, governs references to design control measures in the COLR. It states, Design control measures shall be applied to items such as the following:
reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials; accessibility for inservice inspection, maintenance, and repair; and delineation of acceptance criteria for inspections and tests.
3.0 TECHNICAL EVALUATION
The methodology applies the NRC-accepted S-RELAP5 thermal-hydraulic analysis computer code to the analysis of UFSAR Chapter 15 non-LOCA transients. S-RELAP5 is an updated version of the NRC-accepted ANF-RELAP code.
By letters dated May 11, 2001 (Reference 5) and May 19, 2004 (Reference 6), the staff accepted the use of this methodology for analysis of certain non-LOCA events in pressurized-water reactors of Combustion Engineering (2x4 plants) and Westinghouse (3 and 4-loop plants) design. The staffs acceptance also noted that a generic description of S-RELAP5 cannot provide a detailed justification for all plant applications. Therefore, each applicant must provide justification for its specific application of the S-RELAP5 code, which is expected to include, as a minimum, the nodalization, defense of the chosen parameters, any needed sensitivity studies, justification of the conservative nature of the input parameters, and the calculated results. Accordingly, CP&L has provided this information for a loss of forced reactor coolant flow analysis for HBRSEP2 (Reference 7).
The loss of forced reactor coolant flow event is among the non-LOCA events [5] that may be analyzed using this methodology (Reference 3). The specific events for which this methodology has been accepted are:
Increase in Heat Removal by the Secondary System Increase in Feedwater Flow Increase in Steam Flow Inadvertent Opening of Steam Generator Relief/Safety Valve Steam System Piping Failures Inside and Outside Containment Decrease in Heat Removal by Secondary System Loss of Outside External Load Turbine Trip Loss of Condenser Vacuum Closure of Main Steam Isolation Valve Steam Pressure Regulator Failure Loss of Non-Emergency AC Power to the Station Auxiliaries Loss of Normal Feedwater Flow Feedwater System Piping Breaks Inside and Outside Containment Decrease in Reactor Coolant Flow Rate Loss of Forced Reactor Coolant Flow Flow Controller Malfunctions Reactor Coolant Pump Rotor Seizure Reactor Coolant Pump Shaft Break Reactivity and Power Distribution Anomalies Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power Rod Cluster Control Assembly Misoperation Dropped Rod/Bank Single Rod Withdrawal Statically Misaligned Rod Cluster Control Assembly Startup of an Inactive Loop at an Incorrect Temperature Chemical and Volume Control System (CVCS) Malfunction that Results in a Decrease of Boron Concentration Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position Spectrum of Rod Ejection Accidents Increase In Reactor Coolant Inventory Inadvertent Operation of the Emergency Core Cooling System that Increases Reactor Coolant Inventory CVCS Malfunction that Increases Reactor Coolant Inventory Decreases in Reactor Coolant Inventory Inadvertent Opening of a Pressurizer Pressure Relief Valve Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment (for primary and secondary mass and energy release calculations)
Radiological Consequences of Steam Generator Tube Rupture (for primary and secondary mass and energy release calculations)
CP&Ls analysis of a loss of forced reactor coolant flow event for HBRSEP2 (Reference 7),
includes more information than is normally provided in the licensing basis (e.g., in Safety Analysis Reports). The discussion of input assumptions, for example, notes the conservative direction of errors or uncertainties in the input values. There are also nodalization diagrams of the S-RELAP5 models used for the analysis. The nodalization is comparable in detail to nodalizations that are used in other codes that have been approved by the staff for analysis of licensing basis non-LOCA events. The reactor coolant system (RCS) is modeled by multi-node representations for the reactor vessel, which is comprised of an active core region, inlet and outlet plena, a downcomer and baffel region, and an upper head. Three reactor coolant loops are modeled, connected to three steam generators and a pressurizer. The steam generator models contain inlet and outlet plena and multi-node U-tubes for the primary side, and there are multi-node downcomers, U-tube boiling regions, separators and steam domes in the secondary side. Steam lines, steam safety valves and steam line isolation valves are also represented.
The results of the loss of forced coolant flow analyses are typical for a three-loop plant of Westinghouse design. The analysis is performed with assumptions and models that are designed to minimize the departure from nucleate boiling ratio (DNBR). For example, operation of the pressurizer power-operated relief valves (PORVs) was modeled, since that would tend to keep the core pressure low, which would reduce the core thermal margin, which is represented by the DNBR. The results indicate that the PORVs open and relieve steam (i.e., the pressurizer does not become water-solid), and the transient DNBR does not fall below the safety limit DNBR. Thus, it can be concluded that there would be no fuel clad damage, and the event would not develop into a more severe event. Another version of this analysis, in which the PORVs are not assumed to operate, can be performed to show that there would be no overpressurization of the RCS. In this instance, the latter analysis is not necessary, because (1) the loss of forced reactor coolant flow analysis has been submitted only as a demonstration of the applicability of the non-LOCA analysis methodology to HBRSEP2, and not necessarily to establish a new analysis of record, (2) other events of this class (e.g., the loss of load or turbine trip) are known to produce much higher, yet acceptable, RCS pressures in PWRs of Westinghouse design, and (3) the PORVs, alone, successfully limit the RCS pressure to acceptable levels. Therefore, if the PORVs were unavailable, the safety valves, which are about twice as large as the PORVs would be more than adequate to limit the RCS pressure to acceptable levels.
The results of the licensees loss of forced coolant flow analysis indicate that (1) the applicable acceptance criteria used in the licensing basis of HBRSEP2 are satisfied, and (2) the transient trends and values of key parameters (e.g., RCS flow rate, core power level, and RCS temperatures) are consistent with those produced by other NRC-approved codes that are used to analyze licensing basis UFSAR Chapter 15 events.
Based on the staffs review of the licensees loss of forced coolant flow analysis, which was submitted in accordance with the requirement expressed in the staffs acceptance of this non-LOCA analysis methodology (Reference 5), the staff agrees that said methodology may be added to the list of methodologies in the COLR.
As a result of the staffs review of CP&Ls plant-specific analysis, which was submitted in accordance with the limitation contained in the staffs acceptance of the non-LOCA analysis methodology (Reference 3), the staff concludes that a reference to EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," may be added to the COLR methodologies list of TS 5.6.5 for HBRSEP2.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of South Carolina official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes recordkeeping, reporting, or administrative procedures or requirements. The NRC 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 (71 FR 51224). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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: (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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
(1)
Letter from J. F. Lucas, Progress Energy Corporation (PEC), to NRC, Request for Technical Specifications Change to Core Operating Limits Report References, April 11, 2006. (ML061080522 [Agencywide Documents Access and Management System Accession Number])
(2)
Letter from J. F. Lucas (PEC), to NRC, Request for Partial Withdrawal of Technical Specifications Change to Core Operating Limits Report References, November 9, 2006. (ML063180233)
(3)
EMF-2310(P)(A), Rev. 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP, May 2004 (ML041810034)
(4)
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Revision 2, April 1996 (ML033580677)
(5)
Letter from NRC to Framatome ANP, Acceptance for Referencing of Licensing Topical Report EMF-2310(P), Revision 0, "SRP Chapter 15 Non-LOCA Methodology For Pressurized Water Reactors" (TAC No. MA7192), May 11, 2001 (ML011310533)
(6)
Final Safety Evaluation for Topical Report EMF-2310(P), Revision 1, SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, (TAC No. MC0329),
May 19, 2004 (ML041400499)
(7)
ANP-2512 (P), Loss of Forced Reactor Coolant Flow Analysis for Robinson, AREVA NP, March 2006, (Proprietary)
(8)
Title 10 of the Code of Federal Regulations, Appendix A, Part 50, General Design Criteria for Nuclear Power Plants.
(9)
NUREG-1431, Standard Technical Specifications, Westinghouse Plants, NRC, June 2004 Principal Contributor: S. Miranda Date: November 29, 2006