ML103470479

Issuance of Amendments 159 & 141, Regarding Revision to Technical Specification Table 3.3.1-1

ML103470479
Person / Time
Site:	Vogtle
Issue date:	02/07/2011
From:	Patrick Boyle Plant Licensing Branch II
To:	Ajluni M Southern Nuclear Operating Co
Boyle, Patrick, NRR/DORL/LPL2-1/415-3936
References
TAC ME3302, TAC ME3303
Download: ML103470479 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 February 7, 2011 Mr. M. J. Ajluni Nuclear Licensing Director Southern Nuclear Operating Company, Inc.

P.O. Box 1295 Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION TO TECHNICAL SPECIFICATION (TS)

TABLE 3.3.1-1 (TAC NOS. ME3302 AND ME3303)

Dear Mr. Ajluni:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 159 to Facility Operating License NPF-68 and Amendment No. 141 to Facility Operating License NPF-81 for the Vogtle Electric Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the Facility Operating Licenses in response to your application dated February 2, 2010, and supplemented by letter dated October 29, 2010.

Specifically, the amendment adds Surveillance Requirement 3.3.1.15 to TS Table 3.3.1-1, "Reactor Trip System Instrumentation," Function 3, "Power Range Neutron Flux High Positive Rate," which requires verification that the response time is within limits.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

,,1af~/r, -* :J~.yf---.

Patrick G. Boyle, Project Manager Plant licenSing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosures:

1. Amendment No. 159 to NPF-68

2. Amendment No. 141 to NPF-81

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 159 License No. NPF-68

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Facility Operating License No. NPF-68 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated February 2, 2010, as supplemented by letter dated October 29, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 159, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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 J. Kulesa, Chier ant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-68 Date of Issuance: February 7,2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 141 License No. NPF-81

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Facility Operating License No. NPF-81 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated February 2,2010, as supplemented by letter dated October 29, 2010 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 141, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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 vt9~ ~

~ J. Kulesa, Chier Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-81 Date of Issuance: February 7,2011

ATTACHMENT TO LICENSE AMENDMENT NO. 159 FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND TO LICENSE AMENDMENT NO. 141 FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 Replace the following pages of Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are the Licenses and the TSs identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. NPF-68, page 4 License No. NPF-68, page 4 License No. NPF-81, page 3 License No. NPF-81, page 3 TSs TSs 3.3.1-14 3.3.1-14

(1 ) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2) Technical Specifications and Environmental Protection Plan The Technical Soecifications contained in Appendix A, as revised through Amendment No. 159 and the Environmental Protection Plan contained in Appendix B. both of which are attached hereto. are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.

(4) Deleted (5) Deleted (6) Deleted (7) Deleted (8) Deleted (9) Deleted (10) Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5.

Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets

2.

Communications

3. Minimizing fire spread

4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment

6.

Training on Integrated fire response strategy Renewed Operating License No. NPF-68 Amendment No. 159

-3 (2)

Georgia Power Company, Oglethorpe Power Corporatton, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, pursuant to the Act and 10 CFR Part 50, to possess but not operate the facility at the designated location in Burke County, Georgia, in accordance with the procedures and limitations set forth in this license:

(3)

Southern Nuclear, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel. in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)

Southern Nuclear, 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 calibration or associated with radioactive apparatus or components; (6)

Southem Nuclear, pursuant to the Act and 10 CFR Parts 30.40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein.

C. This license shall be deemed to contain and Is subject to the conditions Specified in lhe Commission's regulations set forth in 10 CFR Chapter I 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 Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through

\\

Amendment No. 141 and the Environmental Protection Plan contained in Appendix B, both ot which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical S pecificati ons and the Environmental Protection Plan.

The Surveillance Requirements (SRs) contained in the Appendix A Technical Specifications and listed below are no! required to be performed immediately upon implementation of Amendment No. 74, The SRs listed below shall be Renewed Operatino License NPF*B1 Amendment No. 141

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 9)

Reactor Trip System Instrumentation APPLICABLE MODES OR NOMINAL OTHER TRIP SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE SETPOINT(n)

FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VAlUE

1.

Manual Reactor Trip 1,2 3(a), 4(a), 5(a) 2 2

B C

SR 3.3.1.13 SR 3.3.1.13 NA NA NA NA

2.

Power Range Neutron Flux

a.

High 1.2 4

D SR 3.3.1.1 SR 3.3.1.2 SR3.3.1.7 SR3.3.1.11 SR 3.3.1.15

<;; 111.3% RTP 109% RTP

b.

Low 1(b},2 4

E SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.15

<;;27.3% RTP 25% RTP

3.

Power Range Neutron Flux High Positive Rate 1,2 4

E SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.15

!> 6.3% RTP with time constant

2 sec 5%RTP with time constant

2 sec

4.

Intermediate Range Neutron Flux 1(b),2(c) 2 F,G SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11

<;;41.9% RTP 25% RTP 2(d) 2 H

SR3.3.1.1 SR 33.1.8 SR 3.3.1.11 s;41.9% RTP 25% RTP (a)

With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b)

Below the P-10 (Power Range Neutron Flux) interlocks.

(c)

Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(n)

A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpeint value is conservative wilh respect to Its associated Allowable Value and the channel is readjusted 10 within the established calibration tolerance band of the Nominal Trip Selpeint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpolnt as necessary in response to plant conditions.

Vogtle Units 1 and 2 3.3.1-14 Amendment No. 159 (Unit 1)

Amendment No. 141 (Unit 2)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 159 TO FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT NO. 141 TO FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425

1.0 INTRODUCTION

By letter dated February 2, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100340033), as supplemented by letter dated October 29,2010 (ADAMS Accession No. ML103090183), Southern Nuclear Operating Company, Inc. (the licensee), submitted a license amendment request to change the technical specifications (TSs) for the Vogtle Electric Generating Plant, Units 1 and 2 (Vogtle 1 and 2). The supplement dated October 29, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 4, 2010 (75 FR 23817).

The proposed change adds response time testing in Surveillance Requirement (SR) 3.3.1.15 for TS Table 3.3.1-1, "Reactor Trip System [RTS] Instrumentation," Function 3, "Power Range, Neutron Flux, High Positive Rate [PFRT]." The SR 3.3.1.15 requires verification that the RTS response time is within limits. The licensee is proposing this change based on a Westinghouse Electric Corporation reanalysis of the rod cluster control assembly bank withdrawal at power (RWAP) event.

As stated in the TS Bases, SR 3.3.1.15 verifies that the individual channel/train actuation response times are less than or less than or equal to, the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in Chapter 16 of the Updated Final Safety Analysis Report (UFSAR). The UFSAR Table 16.3-1 lists the associated response time as N/A for item 3, the PFRT. Since the PFRT is not credited in the analysis of record (AOR) in Chapter 15 of the UFSAR, the response time for the PFRT, is not being verified.

In Nuclear Safety Advisory Letter (NSAL) 09-1 dated February 4, 2009, Westinghouse, the manufacturer of the Vogtle 1 and 2 nuclear steam supply system, indicated that the methodology used for the generic and plant-specific reactor coolant system (RCS) overpressure analyses during an RWAP event incorrectly assumed that use of a minimum initial power level in the RWAP

-2 analyses was conservative, resulting in a maximum peak RCS pressure. The AOR assumed an initial power of 10 percent of rated thermal power (RTP), minus calorimetric uncertainty.

Westinghouse sensitivity studies have identified cases from higher initial power levels that are more limiting, resulting in a higher peak RCS pressure. With the generic analysis key parameters and methodology, some cases exceeded the RCS overpressure limit for initial power levels in the range of 60-percent to BO-percent RTP.

In order to address the concerns of RCS over-pressurization discussed in NSAL-09-1, Westinghouse has performed a reanalysis of the RWAP event for Vogtle 1 and 2. The results of the plant-specific reanalysis show that the peak RCS pressure during an RWAP event is within the RCS pressure limit listed in Vogtle 1 and 2 TS 2.1.2, assuming that credit is taken for a PFRT.

The proposed TS changes are to reflect the reanalysis of the RWAP that assumed a PFRT at 9 percent of RTP with a lag time constant of 2.0 seconds and a trip response time of 0.65 seconds.

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulatory requirements in its review of the license amendment request:

Title 10 of the Code ofFederal Regulations (10 CFR), Part 50, "Domestic Licensing of Production and Utilization Facilities," establishes the fundamental regulatory requirements with respect to the domestic licensing of nuclear production and utilization facilities. Specifically, the general design criteria (GDC) in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 provide, in part, the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety.

GDC 13, "Instrumentation and control," requires that instrumentation shall be provided to monitor variables and sy~tems over their anticipated ranges for normal operation, anticipated operational occurrences, and accident conditions as appropriate to ensure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

GDC 15, "Reactor coolant system design," states that the RCS and associated auxiliary control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary (RCPB) are not exceeded during any condition of normal operation, including anticipated operational occurrences.

GDC 20, "Protection system functions," requires the protection system to be designed (1) to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences, and (2) to sense accident conditions and to initiate the operation of systems and components important to safety

- 3 GDC 21, "Protection system reliability and testability," states that the protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.

GDC 22, "Protection system independence," through GDC 25, "Protection system requirements for reactivity control malfunctions," and GDC 29, "Protection against anticipated operational occurrences," require various design attributes for the protection systems, including independence, safe failure modes, separation from control systems, requirements for reactivity control malfunctions, and protection against anticipated operational occurrences.

The NRC's regulatory requirements related to the content of the TSs are set forth in 10 CFR 50.36, "Technical specifications." This regulation requires that the TSs include items in five categories.

These categories include (1) safety limits, limiting safety system settings, and limiting control setting, (2) limiting conditions for operation (LCOs), (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls.

Section 50.36(c)(3) requires that SRs relating to test, calibration, or inspection ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

Additionally, 10 CFR 50.36(d)(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:

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 (RCPB);

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) 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 (SSC) that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; and Criterion 4: An SSC which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

3.0 TECHNICAL EVALUATION

The PFRT ensures that protection is provided against the rapid increases in neutron flux that are characteristic of a rod cluster control assembly drive rod housing rupture and the accompanying ejection of the assembly. The Vogtle 1 and 2 safety analysis did not previously credit the PFRT for protection against anticipated transients or postulated accidents. However, based on a Westinghouse reanalysis of RWAP transient using more conservative analytical assumptions, the licensee has determined that the Vogtle 1 and 2 safety analysis should credit the PFRT for primary protection. Therefore, the licensee has proposed the revision of TS 3.3.1, "Reactor Trip System Instrumentation," to require verification of PFRT response times.

Westinghouse issued NSAL 09-1, "Rod Withdrawal at Power Analysis for Reactor Control System Overpressure," dated February 4, 2009, which discussed the potential for RCS over-pressurization of a control RWAP with the overall results of analyses requiring PFRT credit for this event. Westinghouse had completed specific analyses of the RWAP for Vogtle 1 and 2.

The results of the Vogtle 1 and 2 analyses, using a conservative methodology, demonstrate that the RCS overpressure limit listed in Vogtle 1 and 2 TS 2.1.2 (i.e., 2,735 pounds per square inch gauge (psig)) is not violated, assuming that credit is taken for a PFRT at or below 9 percent of RTP with a lag time constant of 2 seconds and a trip delay time of 0.65 seconds. However, credit for three trip functions (high pressurizer pressure reactor trip, high power range neutron flux (PRNF) reactor trip, and PFRT) is required in order to meet the RCS pressure acceptance criterion (i.e.,

RCS peak pressure not in excess of 2,735 psig). Without credit for the PFRT, the maximum calculated RCS pressure would not have met the analysis acceptance criterion in some of the 680 cases analyzed.

3.1 Reanalysis of a RWAP Event In Reference 1, the licensee indicated that it performed a reanalysis of 680 cases for an RWAP event initiating from various plant conditions and reactivity insertion rates. All cases assumed reactor trips from the signals of (1) the high flux trip at 120-percent RTP with a trip delay time of 0.5 seconds, (2) the high pressurizer pressure trip at 2440 psia with a trip delay time of 2.0 seconds, and (3) the PFRT at 9-percent RTP with a time constant of 2.0 seconds and a response time of 0.65 seconds for overpressure protection. The current Chapter 15.4.2 of the UFSAR documented the analysis of the RWAP event for Vogtle 1 and 2, for which the limiting acceptance criterion is the limit of the departure from nucleate boiling ratios (DNBRs). The reanalysis in support of the addition of SR 3.3.1.15 to the PFRT was to ensure protection from over-pressurization of the RCS.

SR 3.3.1.15 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis. Chapter 16 of the Vogtle 1 and 2 final safety analysis report (FSAR) includes response time testing acceptance criteria. Response time may be verified by actual response time tests in any series of sequential, overlapping, or total channel measurements, or by the summation of allocated sensor, signal processing, and actuation logic response times with actual response time tests on the remainder of the channel.

During the course of the review, the NRC staff requested the licensee to provide the results of the RWAP reanalysis for review.

In response (Reference 2), the licensee indicated that the RWAP reanalysis was performed with the NRC-approved code, LOFTRAN, documented in WCAP-7907-P-A. It applied the code for the RWAP reanalysis in accordance with the NRC's safety evaluation approving the topical report, and made no methodology changes associated with the reanalysis. All cases were analyzed at minimum reactivity feedback conditions, because an RWAP at maximum reactivity feedback is a slower transient that would be less limiting in terms of RCS pressure. Sensitivity studies were performed for reactivity insertion rate from 15 pcm/sec to 110 pcm/sec and power level from 8 to 102 percent of RTP. The results of the reanalysis are shown in Reference 2 as Figure 1, "Peak RCS Pressure versus Reactivity Rate," and Figure 2, "Peak RCS Pressure versus Power Fraction of RTP." The results in both Figures represent the 340 transient calculations initiating from a pressurizer pressure of 2200 psia and the 340 transient calculations for cases with an initial pressure of 2300 psia. Figure 1 shows that the peak RCS pressure is 2706.32 psia, which is based on a case with reactivity insertion rate of 24 pcm/sec and an initial pressurizer pressure of 2200 psia. Figure 2 shows the peak RCS pressure is also 2703.32 psia, which is calculated for

- 5 a case initiating from BO-percent RTP with an initial pressurizer pressure of 2200 psia, The results of both Figures 1 and 2 show that the peak RCS pressure is less than the pressure limit of 2750 psia, meeting the GDC 15 requirements, Since the method used for the RWAP reanalysis is an NRC-approved method, the values used for the key input parameters are conservative, resulting in a higher peak RCS pressure, and the results of a sensitivity study show that the peak RCS pressure is less than 2750 psia (110 percent of the design pressure of 2500 psia), meeting the GDC 15 requirements and, thus assuring integrity of the RCPS, the NRC staff concludes that the RWAP reanalysis is acceptable, 3.2 Analysis of PRNF Signal Processing The TS defines the RTS response time as that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The loss of stationary gripper coil voltage occurs when the operable rod cluster control assemblies drop into the reactor core and shut down or trip the reactor. The licensee stated that both the PRNF-High and PFRT reactor trip signals are sent from the nuclear instrumentation system (NIS) to the protection system via the same components and are processed from the same PRNF signals and identical bi-stables. The additional signal processing that forms the difference between the flux signal and the lagged flux signal does not add significant time delay for the PFRT.

Therefore, the PFRT trip signal sent to the protection system has a response time similar to the PRNF-High reactor trip function, The licensee stated that the PFRT response time is verified per WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests." The approach is based on having performed an initial baseline RTS response time test. The licensee has reviewed preoperational test data and confirmed the response time of the NIS and solid state protection system (SSPS) for the PFRT during preoperational testing, The slowest response time measured for a channel was 0,100 seconds, Per Vogtle 1 and 2 FSAR Table 16.3-3A, the allocation time for processing is 0.220 seconds, Therefore, the licensee meets the allocation times delineated in the Vogtle 1 and 2 FSAR The licensee has confirmed that the PRNF overall function response time is less than 0,65 seconds as specified by the Westinghouse analysis. This confirmation is based upon the response time measured during preoperational testing of the NIS and SSPS processing times, and the most recent surveillance data for the other components that comprise the string.

The licensee modified SR 3,3.1,15 by including a note stating that neutron detectors are excluded from RTS response time testing because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.

The RWAP analysis assumed a rate setpoint of 9-percent RTP safety analysis limit with a lagging time constant of 2,0 seconds and a 0,65-second trip delay, In Reference 2, the licensee explained the current nominal trip setpoint and allowable value settings. The licensee further confirmed that the nominal trip setpoint of 5-percent RTP with a time constant of;::2 seconds and the existing allowable value of s6.3-percent RTP with a time constant of ;::2 seconds would ensure that the 9-percent RTP safety analysis limit would be protected. Therefore, the NRC staff concluded that the proposed trip setpoint would allow adequate PRNF signal processing.

- 6 3.3 Compliance with the 10 CFR 50.36(d)(2)(ii) requirements During the review, the NRC staff reviewed the PFRT function and the associated trip response time against the criteria specified in 10 CFR 50.36(d)(2)(ii) as follows:

Criterion 1: The PFRT and trip response time were not used to detect and indicate a significant abnormal degradation of the RCPB.

Criterion 2: The PFRT and trip response time were not a process variable, design feature, or operating restriction that was an initial condition of a DBA or transient analysis.

Criterion 3: Credit was taken for the PFRT and trip response time in the reanalysis of an RWAP event for protection from over-pressurization. The PFRT and delay time were considered as part of the primary success path related to the integrity of a fission product barrier. Therefore, the PFRT and the delay time were an SSC that was part of the primary success path and which functioned or actuated to mitigate a DBA or transient that either assumed the failure of or presented a challenge to the integrity of a fission product barrier.

Criterion 4: The PFRT and trip response time were not an SSC which operating experience or probabilistic risk assessment had shown to be significant to public health and safety.

Based on the above discussion, the NRC staff found that the PFRT and trip response time satisfied Criterion 3 in 10 CFR 50.36(d){2)(ii). The addition of SR 3.3.1.15 to Table 3.3.1-1, Function 3, is the appropriate requirement to verify that the response times are less than or equal to the maximum values assumed in the accident analysis.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the surveillance requirements. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (75 FR 23817). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.221(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

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

~ 7 ~

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 M. J. Ajluni (Southern Nuclear Operating Company), to NRC, "Vogtle Electric Generating Plant License Amendment Request for Technical Specification Table 3.3.1-1,"

dated February 2, 2010 (ADAMS Accession No. ML100340033).

2.

Letter from M. J. Ajluni (Southern Nuclear Operating Company), to NRC, "Vogtle Electric Generating Plant, Response for Additional Information RAI Regarding License Amendment Request for Technical Specification Table 3.3.1~1," dated October 29,2010 (ADAMS Accession No. ML103090183).

Principal Contributors: P. Chong, NRRlDE/EICB S. Sun, NRRlDSS/SRXB Date: February 7, 2011

February 7,2011 Mr. M. J. Ajluni Nuclear Licensing Director Southern Nuclear Operating Company, Inc.

P.O. Box 1295 Bin - 038 Birmingham, AL 35201-1295

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION TO TECHNICAL SPECIFICATION (TS)

TABLE 3.3.1-1 (TAC NOS. ME3302 AND ME3303)

Dear Mr. Ajluni:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 159 to Facility Operating License NPF-68 and Amendment No. 141 to Facility Operating License NPF-81 for the Vogtle Electric Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the Facility Operating Licenses in response to your application dated February 2,2010, and supplemented by letter dated October 29,2010.

Specifically, the amendment adds Surveillance Requirement 3.3.1.15 to TS Table 3.3.1-1, "Reactor Trip System Instrumentation," Function 3, "Power Range Neutron Flux High Positive Rate," which requires verification that the response time is within limits.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, IRA!

Patrick G. Boyle, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosures:

1. Amendment No. 159 to NPF-68

2. Amendment No. 141 to NPF-81

3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

Public LPL2-1 R/F RidsRgn2MailCenter Resource RidsNrrLASRohrer Resource RidsNrrDeEicb RidsAcrsAcnw_MailCTR Resource RidsNrrDorlLpl2-1 Resource RidsOgcRp Resource RidsNrrDssSrxb RidsNrrDirsltsb Resource RidsNrrPMVogtle Resource RidsNrrDorlDpr Resource PChong, NRR SSun, NRR ADAMS Accession No ML103470479 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/DSS/SRXB NRRIDElEICB NAME PBoyle SRohrer AUlses 1/23/11 and memos dated 11/9 and 11/17/2010 WKemper (RStattel for)

DATE 11 01/10/11 1123/11 1/14111 OFFICE NRR/LPL2-1/BC NRR/LPL2-1/PM NAME LSubin NLO w/comment GKulesa (VSreenivas for)

PBoyle DATE 01/31/11 0214/11 0217111 OFFICIAL RECORD COpy