ML13114A714
| ML13114A714 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 05/29/2013 |
| From: | Billoch-Colon A Plant Licensing Branch II |
| To: | William Gideon Carolina Power & Light Co |
| Billoch A NRR/DORL/LPL2-1 301-415-3302 | |
| References | |
| TAC ME9516 | |
| Download: ML13114A714 (13) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 29,2013 Mr. William G. Gideon, Vice President H. B. Robinson Steam Electric Plant Carolina Power & Light Company 3581 West Entrance Road Hartsville, SC 29550
SUBJECT:
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2-ISSUANCE OF AN AMENDMENT TO ELIMINATE STEAM GENERATOR WATER LEVEL-LOW COINCIDENT WITH STEAM FLOW/FEEDWATER FLOW MISMATCH FUNCTION FROM TECHNICAL SPECIFICATIONS TABLE 3.1.1-1. REACTOR PROTECTION SYSTEM INSTRUMENTATION (TAC NO. ME9516)
Dear Mr. Gideon:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 234 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP). This amendment changes the HBRSEP Technical Specifications (TSs) in response to your application dated September 6, 2012 (Agencywide Documents Access and Management System Accession No. ML12263A424), as supplemented by letter dated, December 7,2012 (ML12361A015).
The amendment eliminates Function 14, Steam Generator Water Level-Low Coincident with Steam Flow/Feedwater Flow Mismatch, from the HBRSEP TS Table 3.3.1-1, "Reactor Protection System Instrumentation."
A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, Araceli T. Billoch Colon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261
Enclosures:
- 1. Amendment No. 234 to DPR-23
- 2. Safety Evaluation cc w/enclosures: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 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. 234 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 September 6,2012, as supplemented by letter dated December 7, 2012, 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.
Enclosure 1
-2
- 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 3.B. of Renewed Facility Operating License No. DPR-23 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 234 are hereby incorporated in the license.
The licensee 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 prior exiting the scheduled fall 2013 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to Operating License No. DPR-23 and the Technical Specifications Date of Issuance: May 29, 2013
ATTACHMENT TO LICENSE AMENDMENT NO. 234 RENEWED FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Replace the following pages of the Renewed Facility Operating License and Appendix "An Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove License DPR-23 License DPR-23 Page 3 Page 3 TSs TSs 3.3-16 3.3-16
-3 neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. 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; E. 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 operation of the facility.
- 3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR 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:
A. Maximum Power Level The licensee is authorized to operate the facility at a steady state reactor core power level not in excess of 2339 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 234 are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
(1) For Surveillance Requirements (SRs) that are new in Amendment 176 to Final Operating License DPR-23, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 176. For SRs that existed prior to Amendment 176, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 176.
Renewed Facility Operating License No. DPR-23 Amendment No.234
RPS Instrumentation 3.3.1 Table 3.3.1*1 (page 4 of 7)
Reactor Protection System Instrumentation APPLICABLE MODES OR NOMINAl OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE SETPOINT FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE (1)
- 14. Deleted
- 15. Turbine Trip
- b. Turbine Stop 2 P SR3.3.1.15 NA NA Valve Closure
- 16. Safety Injection (SI) 1.2 2 trains Q SR 3.3.1.14 NA NA Input from Engineered Safety Feature Actuation System (ESFAS)
(continued)
(1) A channel is OPERABLE with an actual Trip Setpoint value found outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint.
(1) Above the P-8 (Power Range Neutron Flux) interlock.
HBRSEP Unit NO.2 3.3-16 Amendment No. 234
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 234 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 to the U.S. Nuclear Regulatory Commission (NRC) dated September 6,2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12263A424), as supplemented by letter dated December 7,2012 (ML12361A015), Carolina Power & Light Company (the licensee), doing business as Progress Energy Carolinas, Inc.,
submitted a license amendment request for changes to the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP, the licensee), Technical Specifications (TSs) to eliminate Function 14, Steam Generator Water (SG) Level-Low Coincident with Steam Flow/Feedwater Flow Mismatch, from the HBRSEP TS Table 3.3.1-1, "Reactor Protection System [RPS]
Instrumentation." The request for change is based on the installation of new equipment to eliminate the circumstance that warranted this particular reactor trip function in the HBRSEP TS.
The supplement dated December 7,2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's initial proposed no significant hazards consideration determination as published in the Federal Register on November 27,2012 (77 FR 70840).
2.0 REGULATORY EVALUATION
The purposes of the RPS are to: 1) initiate a reactor trip if safe operating limits are exceeded, and 2) initiate engineered safety features action(s) if an accident occurs. The initiation of a reactor trip by the RPS prevents the core from operating in a condition that could cause damage to the core. The reactor plant operating limits are determined and set by the utility's Final Safety Analysis Report, and are incorporated into the plant TSs.
Section 182a of the Atomic Energy Act requires nuclear power plant operating licenses to include TSs as part of any license. The NRC regulatory requirements related to the content of the TSs are contained in Title 10, Code of Federal Regulations (10 CFR), Part 50, Section 50.36, 'Technical Specifications." The TS requirements in 10 CFR 50.36 include the following categories: 1) safety limits, limiting safety systems settings and control settings; 2) limiting Enclosure 2
- 2 conditions for operation; 3) surveillance requirements; 4) design features; 5) administrative controls; 6) decommissioning; 7) initial notification; and 8) written reports.
Section 50.36(c)(2)(i) of 10 CFR states that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
Additionally, 10 CFR 50.36( c)(2)(ii) sets forth four criteria to be used in determining whether an LCO is required to be included in the TSs. These criteria are as follows:
- Criterion 1: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
- Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- Criterion 4: A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
Section 50.36(c)(3) to 10 CFR states that Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.
Section 50.55a(h)(2) to 10 CFR "Protection systems" states that for nuclear power plants with construction permits issued after January 1, 1971, but before May 13, 1999, protection systems must meet the requirements stated in [Institute of IEEE [Electrical and Electronics Engineers]
Std. 279, "Criteria for Protection Systems for Nuclear Power Generating Stations."
3.0 TECHNICAL EVALUATION
Originally, HBRSEP used one of the three redundant SG water level signals for input to the feedwater control subsystem and all three SG water level signals to initiate an RPS trip if at least two of the three signals reach the trip setpoint value. This original configuration resulted in adverse control and protection system interaction. A single failure could initiate an adverse water level control action that would also require a subsequent protective action to prevent exceeding design safety limits. For example, a postulated failure (high) of one SG water level channel could initiate a one out of three high-level signal to the RPS. Since this signal is also used for feedwater control, the failed channel would send a high-level signal to feedwater control circuits, causing the feedwater regulating valve for the associated SG to shut. This
- 3 would cause the SG water level to decrease. In such a scenario, the IEEE Standard 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations," Clause 4.7.3, imposes the requirement for degradation by a second random failure in addition to the first initiating event. The underlying logic is that the initial protection system failure is the initiating event for the transient. Therefore, the initial failed channel does not constitute the "single failure" that IEEE-279 imposes on the protection system. As such, an additional protection failure must be postulated to occur, and the protection system must continue to be capable of initiating the appropriate protective action.
The limiting postulated single failure (a second failure postulated by IEEE 279, Clause 4.7.3) would be a failure of one of the remaining two SG level channels. This would leave only one operating channel. In this case, the system would be unable to satisfy the 2/3 logic needed to initiate a SG Water Level Low-Low Reactor Trip.
To address this single failure concern, the original design included a function to prevent adverse control and protection system interaction from taking place because of failure of an SG water level signal. The design includes an additional trip function that actuated on low SG water level coincident with a mismatch in feedwater flow and steam flow (e.g., SG Water Level-Low Coincident with Steam Flow/Feedwater Flow Mismatch). This added trip function ensures that the necessary protective action (Le., reactor trip) to meet safety limits would occur. The SG water level channels that input to this trip do not include the control channel. Therefore, the reduction in water level in the affected SG sensed by the remaining operable level channel, combined with the Steam Flow/Feedwater Flow Mismatch Signal, would result in the required reactor trip.
During Refueling Outage 27, the licensee performed an engineering change that altered the number of SG water level signals available for feedwater control. With this change, the three SG water level signals used as input to the RPS for each SG provide inputs to a median signal selector (MSS) module. This module identifies the median signal of the three inputs and sends the selected signal to the feedwater controllers. Each RPS SG water level signal is provided to the MSS module through an isolation device to separate the qualified and nonqualified portions of the control loops. With these changes, failure of one SG level measurement channel will no longer result in a transient induced by the feedwater controllers. Therefore, with this modified design, the SG Water Level-Low. Coincident with Steam Flow/Feedwater Flow Mismatch, is no longer required to prevent an adverse control or protection system interaction.
This modified system design rejects the high and low SG level signals and the MSS modules will prevent the control system from acting on any single failed protection system instrument channel. Since no adverse control system action will now result from a single. failed protection instrument channel. IEEE-279 does not require consideration of a second random protection system failure. Signals resulting from a single failed high or low SG level channel will be rejected for control purposes; therefore, it will not affect the capability of the feedwater control system to control the SG water level.
The MSS modules installed at HBRSEP provide similar protection from adverse control and protection system interaction as the MSS installed in Beaver Valley Power Station, Units 1 and 2, that have demonstrated reliable operation since 2001 and 1987, respectively. The HBRSEP MSS modules are similar in design to the Beaver Valley Power Station, Units 1 and 2,
-4 MSS modules in that they are not part of the protection system and are designed to reduce the frequency of system failures.
A failure of an MSS module at HBRSEP does not compromise the ability of the protection system to perform its safety-related functions. The design provides the capability for complete online testing of MSS modules. The performance of periodic calibrations verifies the functionality and calibration of each MSS module for each SG. Calibration will be scheduled during refueling outages. Online calibration following repairs can be performed only while the feedwater control system is in manual control, and each loop calibration will include a functional performance check to verify that the MSS module has been fully restored before returning to automatic operation of the feedwater control system.
The licensee stated that the MSS modules installed at HBRSEP are equivalent to the Beaver Valley Power Station, Units 1 and 2, MSSs in terms of quality, testability, and the use of highly reliable components. The combination of demonstrated performance, low likelihood of failure, and the ability to detect failures through continued periodic testing provide the necessary degree of confidence for MSS module operational readiness and reliability. In addition to eliminating control and protection system interaction concerns, the reliability of the feedwater control system is enhanced with the use of the MSS modules, since potential plant transients that could result from the failure of a single SG level instrument channel are eliminated.
The elimination of the SG Water Level-Low, Coincident with Steam Flow/Feedwater Flow Mismatch, reactor trip from the HBRSEP TS and plant design is consistent with the HBRSEP Updated Final Safety Analysis Report Chapter 15 safety analyses, that do not credit the reactor trip on SG Water Level-Low, Coincident with Steam Flow/Feedwater Flow Mismatch.
In response to the staff request for additional information, the licensee confirmed by letter dated December 7,2012, that the manufacturer has qualified the isolation devices for 1E safety to non-safety isolation. Specific feedwater control for each loop is supplied from the isolation device to the respective control loop to meet the requirements of IEEE 279-1971, Clause 4.7.2.
By letter dated December 7, 2012, the licensee stated that procedures in place can calibrate all modules used for the MSS, both online and during shutdown. Calibration procedures include injecting analog signals and verification of the correct translation of the control outputs while bypassing a channel. In addition, there are computer points for each signal provided to the MSS that can be used to verify each signal through cross checking. These methods provide sensor checks that meet the criteria of IEEE-279-1971, Clause 4.9.
The NRC staff has approved similar elimination of the reactor trip function based on the coincidence of low SG water level and steam flow and feedwater flow mismatch for the following plants:
- North Anna, Units 1 and 2, November 20,2012 (ADAMS Accession No. ML12249A299),
- Beaver Valley, Unit 1, August 8,2001 (ADAMS Accession No. ML011410204),
- Beaver Valley, Unit 2, February 20,1990 (ADAMS Accession No. ML003772209),
- Ginna, February 28, 1991 (ADAMS Accession No. ML010540440),
- Prairie Island, April 3, 1989 (ADAMS Accession No. ML022210127),
-5
- Salem, Units 1 and 2, August 7,1995 (ADAMS Accession No. ML011730350), and
- Diablo Canyon, July 12, 1990 (ADAMS Accession No. ML022340113).
The NRC staff finds that the removal of the SG Water Level-Low, Coincident with Steam Flow/Feedwater Flow Mismatch, trip from the HBRSEP will provide safety benefits. The most significant will be reduced challenges to the overall plant safety systems by eliminating potential spurious and inadvertent reactor trips. Another benefit provided by this modification is greater operational flexibility due to the fact that the SG Level Low setpoint for the function being eliminated is higher than the Low-Low Level trip setpoint. This will reduce the potential for unnecessary SG level related trips. As such, removal of this trip function will enhance safe operation of the plant by reducing potential challenges to safety systems and unnecessary plant transients. Removal of this trip function will eliminate a system that can cause unnecessary plant transients, and will also reduce the need for associated additional training and operating precautions.
Based on the installation of the MSS and the lack of any accident analyses crediting an SG Water Level-Low Coincident with Steam Flow/Feedwater Flow Mismatch Reactor Trip Function, the NRC staff concludes that the licensee may remove Function 14 from Table 3.3.1-1 of its TS because an equivalent level of safety is maintained by inclusion of the MSS.
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 a requirement with respect to installation or use of a facility component 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, 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 (77 FR 70840). 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 (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the
- 6 amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: S. Mazumbar Date: M:ly 29, 2013
May 29,2013 Mr. William G. Gideon, Vice President H. B. Robinson Steam Electric Plant Carolina Power & Light Company 3581 West Entrance Road Hartsville, SC 29550 SUB..IECT: H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO.2 - ISSUANCE OF AN AMENDMENT TO ELIMINATE STEAM GENERATOR WATER LEVEL-LOW COINCIDENT WITH STEAM FLOW/FEEDWATER FLOW MISMATCH FUNCTION FROM TECHNICAL SPECIFICATIONS TABLE 3.1.1-1. REACTOR PROTECTION SYSTEM INSTRUMENTATION (TAC NO. ME9516)
Dear Mr. Gideon:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 234 to Renewed Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP). This amendment changes the HBRSEP Technical Specifications (TSs) in response to your application dated September 6,2012 (Agencywide Documents Access and Management System Accession No. ML12263A424), as supplemented by letter dated, December 7, 2012 (ML12361A015).
The amendment eliminates Function 14, Steam Generator Water Level-Low Coincident with Steam Flow/Feedwater Flow Mismatch, from the HBRSEP TS Table 3.3.1-1, "Reactor Protection System Instrumentation."
A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, IRA!
Araceli T. Billoch Colon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-261
Enclosures:
- 1. Amendment No. 234 to DPR-23
- 2. Safety Evaluation cc w/enclosures: Distribution via ListServ DISTRIBUTION:
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