ML101950353

Issuance of Amendments Regarding Replacement of Source Range and Intermediate Range Encore Detection Systems with Equivalent Neutron Monitoring Systems Using Fission Chamber Detectors

ML101950353
Person / Time
Site:	Catawba
Issue date:	08/02/2010
From:	Thompson J H Plant Licensing Branch II
To:	Morris J R Duke Energy Carolinas
Thompson, Jon 415-1119
References
TAC ME1747, TAC ME1748
Download: ML101950353 (33)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 August 2, 2010 Mr. J. R. Morris Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745 CATAWBA NUCLEAR STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REPLACEMENT OF SOURCE RANGE AND INTERMEDIATE RANGE EXCORE DETECTION SYSTEMS WITH EQUIVALENT NEUTRON MONITORING SYSTEMS USING FISSION CHAMBER DETECTORS (TAC NOS. ME1747 AND ME1748)

Dear Mr. Morris:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 258 to Renewed Facility Operating License NPF-35 and Amendment No. 253 to Renewed Facility Operating License NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively.

The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated July 1, 2009, as supplemented by letter dated May 20, 2010. The amendments revise TS 3.3.1, "Reactor Trip System (RTS) Instrumentation" and TS 1.1, "Definitions." The proposed amendments support plant modifications which would replace the existing source range and intermediate range excore detector systems with equivalent neutron monitoring systems. The new instrumentation will perform both the source range and the intermediate range monitoring functions.

A copy of the related Safety Evaluation is also enclosed.

A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

J. Morris -2 If you have any questions, please call me at 301-415-1119.

Sincerely, Jon Thompson, Project Plant Licensing Branch Division of Operating Reactor Office of Nuclear Reactor Docket Nos. 50-413 and

Enclosures:

1. Amendment No. 258 to 2. Amendment No. 253 to 3. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 258 Renewed License No. NPF-35 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility)

Renewed Facility Operating License No. NPF-35 filed by the Duke Energy Carolinas, LLC, acting for itself, and North Carolina Electric Membership Corporation (licensees), dated July 1,2009, as supplemented by letter dated May 20,2010, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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.

-2Accordingly, 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 Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 258 , which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-35 and the Technical Specifications Date of Issuance:

August 2, 2010 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO.1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 253 Renewed License No. NPF-52 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility)

Renewed Facility Operating License No. NPF-52 filed by the Duke Energy Carolinas, LLC, acting for itself, North Carolina Municipal Power Agency NO.1 and Piedmont Municipal Power Agency (licensees), dated July 1,2009, as supplemented by letter dated May 20,2010, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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.

-2Accordingly, 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 Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment 1\10. 253 ,which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-52 and the Technical Specifications Date of Issuance:

August 2 I 2010 ATTACHMENT TO LICENSE AMENDMENT NO. 258 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 253 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Licenses Licenses NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 TSs TSs 1.1-4 1.1-4 3.3.1-12 3.3.1-12 3.3.1-14 3.3.1-14 3.3.1-15 3.3.1-15 3.3.1-16 3.3.1-16 3.3.1-17 3.3.1-17 3.3.1-18 3.3.1-18 3.3.1-19 3.3.1-19 3.3.1-20 3.3.1-20 3.3.1-21

-4Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No 258 which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications. Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection The Updated Final Safety Analysis Report supplement as revised on December 16,2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)* Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report. as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. "The parenthetical notation follow ing the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplement wherein this renewed license condition is discussed.

Renewed License No. NPF-35 Amendment No. 258 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No 253 Nhich are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications. Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.

Duke shall complete these activities no later than February 24, 2026, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. Fire Protection Program (Section 9.5.1, SER, SSER #2, SSER #3, SSER #4, SSER #5)* Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended, for the facility and as approved in the SER through Supplement 5, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. *The parenthetical notation follo wing the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or its supplements wherein this renewed license condition is discussed.

Renewed License No. NPF-52 Amendment No. 253 Definitions 1.1 1.1 Definitions (continued)

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay. MODE A MODE shall correspond to anyone inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel. NOMINAL TRIP SETPOINT The NOMINAL TRIP SETPOINT shall be the design value of a setpoint.

The trip setpoint implemented in plant hardware may be less or more conservative than the NOMINAL TRIP SETPOINT by a calibration tolerance.

Unless otherwise specified, if plant conditions warrant, the trip setpoint implemented in plant hardware may be set outside the NOMINAL TRIP SETPOINT calibration tolerance band as long as the trip setpoint is conservative with respect to the NOMINAL TRIP SETPOINT.

OPERABLE-OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are: a. Described in Chapter 14 of the UFSAR; b. Authorized under the provisions of 10 CFR 50.59; or c. Otherwise approved by the Nuclear Regulatory Commission. (continued)

Catawba Units 1 and 2 1.1-4 Amendment Nos. 258, 253 RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3. 1.9 ----------------------------------

NOTE Verification of setpoint is not required.

Perform T ADOT. 92 days SR 3.3.1.1 0 This Surveillance shall include verification that the time constants are adjusted to the prescribed values. Perform CHANNEL CALIBRATION.

18 months SR 3.3. 1.11 ----------------------------------

NOTE

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. Power Range Neutron Flux high voltage detector saturation curve verification is not required to be performed prior to entry into MODE 1 or 2. 3. Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed prior to entry into MODE 1 or 2.* Perform CHANNEL CALIBRATION.

18 months SR 3.3.1.12 Perform CHANNEL CALIBRATION.

18 months SR 3.3.1.13 Perform COT. 18 months (continued) This Note applies to the Westinghouse-supplied compensated ion chamber neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors which do not require detector plateau voltage verification.

Therefore, this Note does not apply to the fission chamber neutron detectors.

Catawba Units 1 and 2 Amendment Nos. 258, 253 RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 8) Reactor Trip System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT 1. Manual Reactor Trip 1,2 2 B SR 3.3.1.14 NA NA 3(a), 4(a), 5(a) 2 C SR 3.3.1.14 NA NA 2. Power Range Neutron Flux a. High 1,2 4 D SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 110.9% RTP 109% RTP b. Low 4 E SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 27.1% RTP 25% RTP 3. Power Range Neutron Flux High Positive Rate 1,2 4 D SR 3.3.1.7 SR 3.3.1.11 RTP with time constant :<:: 2 sec 5%RTP with time constant :<::2 sec (continued) (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.

Catawba Units 1 and 2 3.3.1-14 Amendment Nos. 258,253 ,

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 8) Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT Intermediate Range 2 F,G Neutron Flux 2 H Source Range 2 I) Neutron Flux 2 J,K 6. Overtemperature

/';T 1,2 4 E SR SR SR 3.3.1.11 SR SR SR 3.3.1.11 SR SR SR 3.3.1.11 SR SR 3.3.1.SR 3.3.1.11 SR SR SR SR SR SR

31% s.38%

; 31% s.38% :; 1.4 S. 1.44 E5 :; 1.4 S. 1.44 E5 Refer Note 1 25% RTP 25% RTP 1.0 E5 cps 1.0 E5 cps Refer to Note 1 (Page 3.3.1-19) (co nti n ued) The S. 31% RTP Allowable Value applies to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The S. 38% RTP Allowable Value applies to the replacement fission chamber Intermediate Range neutron detectors. ** The S. 1.4 E5 cps Allowable Value applies to the Westinghouse-supplied boron triflouride (BF3) Source Range neutron detectors.

The BF:, neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The S. 1.44 E5 cps Allowable Value applies to the replacement fission chamber Source Range neutron detectors. With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal. Below the P-10 (Power Range Neutron Flux) interlocks. Above the P-6 (Intermediate Range Neutron Flux) interlocks. Below the P-6 (Intermediate Range Neutron Flux) interlocks. If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NOMINAL TRIP SETPOINT (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance.

The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR. Catawba Units 1 and 2 Amendment Nos. 258, 253 RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 8) Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT Overpower I'!.T 1,2 4 E SR 3.3.1.1 Refer to Refer to SR 3.3.1.3 Note 2 (Page Note 2 SR 3.3.1.6 3.3.1-20) (Page SR 3.3.1.7 3.3.1-20)

SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.17 Pressurizer Pressure a. Low 1(e) 4 L SR 3.3.1.1 1938(f) psig 1945(f) SR 3.3.1.7 psig SR 3.3.1.10 SR 3.3.1.16 High 1,2 4 E SR 3.3.1.1 $ 2399 psig 2385 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 9. Pressurizer Water 1(e) 3 SR 3.3.1.1 $ 93.8% 92% Level-High SR3.3.1.7 SR 3.3.1.10 Reactor Coolant Flow-Low Single Loop 1 (g) 3 per loop M SR 3.3.1.1 89.7% 91% SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 Two Loops 1(h) 3 per loop L SR 3.3.1.1 89.7% 91% SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 (continued) Above the P-7 (Low Power Reactor Trips Block) interlock. Time constants utilized in the lead-lag controller for Pressurizer Pressure -Low are 2 seconds for lead and 1 second for lag. Above the P-8 (Power Range Neutron Flux) interlock. Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) interlock.

Catawba Units 1 and 2 Amendment Nos258, 253 RTS Instrumentation Table 3.3.1-1 (page 4 of 8) Reactor Trip System Instrumentation 3.3.1 FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT 11. 12. Undervoltage RCPs Underfrequency RCPs 1 (e) 1 (e) 1 per bus 1 per bus L L SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.16 <:5016V <: 55.9 Hz 5082 V 56.4 Hz 13. Steam Generator (SG) Water Level Low Low 1,2 4 per SG E SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 <: 9% (Unit 1) <: 35.1% (Unit 2) of narrow range span 10.7% (Unit 1) 36.8% (Unit 2) of narrow range span 14. Turbine Trip a. Stop Valve EH Pressure Low 1 (j) 4 N SR 3.3.1.10 SR 3.3.1.15 <: 500 psig 550 psig b. Turbine Stop Valve Closure 1 (j) 4 0 SR 3.3.1.10 SR 3.3.1.15 <: 1% open NA 15. Safety Injection (SI) Input from Engineered Safety Feature Actuation System (ESFAS) 1,2 2 trains P SR 3.3.1.5 SR 3.3.1.14 NA NA (e) Above the P-7 (Low Power Reactor Trips Block) interlock. (continued) (i) Not used. (j) Above the P-9 (Power Range Neutron Flux) interlock.

Catawba Units 1 and 2 3.3.1-17 Amendment Nos. 25,8, 253 t 3.3.1 RTS Instrumentation Table 3.3.1-1 (page 5 of Reactor Trip System APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILu\NCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT Reactor Trip Intermediate 2(d) 2 R SR 3.3.1.11 6E-11 1E-10 Range Neutron SR 3.3.1.13 amp*** amp*** Flux, P-6 1E-5% RTP RTP Low Power 1 per train S SR 3.3.1.5 NA NA Reactor Trips Block. P-7 Power Range 4 S SR 3.3.1.11 $ 50.2% RTP 48% RTP Neutron Flux, SR 3.3.1.13 P-8 Power Range 4 S SR 3.3.1.11 $70% RTP 69% RTP Neutron Flux, SR 3.3.1.13 P-9 Power Range 4 R SR 3.3.1.11 10% RTP Neutron Flux, SR 3.3.1.13 RTP and $ 12.2% P-10 RTP f. 2 S SR 3.3.1.12 10% RTP $12.2% RTP SR 3.3.1.13 turbine turbine Pressure, impulse impulse pressure pressure equivalent equivalent Reactor Trip 1,2 2 trains Q.U SR 3.3.1.4 NA NA Breakers(k) 3(a), 4(a), 5(a) 2 trains C SR 3.3.1.4 NA NA Reactor Trip Breaker 1,2 1 each per T SR 3.3.1.4 NA NA Undervoltage and RTB Shunt Trip Mechanisms 3(a), 4(a), 5(a) 1 each per C SR 3.3.1.4 NA NA RTB Automatic Trip Logic 1,2 2 trains P,U SR 3.3.1.5 NA NA 3(a). 4(a), 5(a) 2 trains C SR 3.3.1.5 NA NA (continued)

The> 6E-11 amp Allowable Value and the 1 E-1 0 amp NOMINAL TRIP SETPOINT value apply to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The 6.6E-6% RTP Allowable Value and the 1 E-5% RTP NOMINAL TRIP SETPOINT value apply to the replacement fission chamber Intermediate Range neutron detectors. With RTBs closed and Rod Control System capable of rod withdrawal. Below the P-6 (Intermediate Range Neutron Flux) interlocks. Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB. Catawba Units 1 and 2 Amendment Nos.258, 253 RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 8) Reactor Trip System Instrumentation Note 1: Overtemperature The Overtemperature Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.3% (Unit 1) and 4.5% (Unit 2) of RTP. t,T(1+r 1 s) [1 J (1+r 4 s) [T 1 (1 +r 2 s 1+r 3(1 +r s s) (1+r a s) Where: is the measured RCS by loop narrow range RTDs, of. is the indicated at RTP, of. s is the Laplace transform operator, sec-1. T is the measured RCS average temperature, of. T' is the nominal T avg at RTP (allowed by Safety Analysis), 2 the values specified in the COLR. P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, =the value specified in the COLR K 1 = Overtemperature reactor NOMINAL TRIP SETPOINT, as presented in the COLR, K 2 = Overtemperature T reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K 3 = Overtemperature T reactor trip depressurization setpoint penalty coefficient, as presented in the COLR, L1, L2 = Time constants utilized in the lead-lag compensator for as presented in the COLR, L3 = Time constant utilized in the lag compensator for T, as presented in the COLR, L4, LS = Time constants utilized in the lead-lag compensator for T avg , as presented in the COLR, La = Time constant utilized in the measured T avg lag compensator, as presented in the COLR, and

= a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: for qt-qb between the "positive" and "negative" breakpoints as presented in the COLR;

= 0, where Qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and Qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; for each percent that the magnitude of qt-qb is more negative than the "negative" breakpoint presented in the COLR, the Trip Setpoint shall be automatically reduced by the "negative" slope presented in the COLR; and (continued)

Catawba Units 1 and 2 Amendment Nos. 258, 253 RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of Reactor Trip System for each percent i11 that the magnitude of qt -qb is more positive than the f 1 (i1I) "positive" breakpoint presented in the COLR, the i1T Trip Setpoint shall be automatically reduced by the f 1 (i1I) "positive" slope presented in the COLR. Note 2: Overpower i1 T The Overpower i1T Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 2.6% (Unit 1) and 3.1 % (Unit 2) of RTP. i1T is the measured RCS i1T by loop narrow range RTDs, of. i1T o is the indicated i1T at RTP, of. s is the Laplace transform operator, sec-1. T is the measured RCS average temperature, of. T" is the nominal T avg at RTP (calibration temperature for i1T instrumentation), .::: the values specified in the COLR. = Overpower i1 T reactor NOMINAL TRIP SETPOINT as presented in the COLR, K s = the value specified in the COLR for increasing average temperature and the value specified in the COLR for decreasing average temperature, Ka = Overpower i1 T reactor trip heatup setpoint penalty coefficient as presented in the COLR for T >T" and Ka =the value specified in the COLR for T .::: TOO, .1, .2 = Time constants utilized in the lead-lag compensator for i1 T, as presented in the COLR, .3 = Time constant utilized in the lag compensator for i1T, as presented in the COLR, .6 = Time constant utilized in the measured T av9 lag compensator, as presented in the COLR, .7 = Time constant utilized in the rate-lag controller for T avg , as presented in the COLR, and fz(i1I) a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: for %-qb between the "positive" and "negative" f 2 (i1I) breakpoints as presented in the COLR; f 2 (i1I) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued)

Catawba Units 1 and 2 Amendment Nos.258, 253 I RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of Reactor Trip System for each percent ,11 that the magnitude of % -qb is more negative than the f 2 (.1I) "negative" breakpoint presented in the COLR, the ,1 T Trip Setpoint shall be automatically reduced by the f 2 (.1I) "negative" slope presented in the COLR; and for each percent ,11 that the magnitude of qt-qb is more positive than the h(.1I) "positive" breakpoint presented in the COLR, the ,1 T Trip Setpoint shall be automatically reduced by the f 2 (.1I) "positive" slope presented in the COLR. Catawba Units 1 and 2 Amendment Nos. 258, 253 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 258 TO RENEWED FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 253 TO RENEWED FACILITY OPERATING LICENSE NPF-52 DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By application dated July 1, 2009 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML091950355), as supplemented by letter dated May 20,2010 (ADAMS Accession No. ML 101530462), Duke Energy Carolinas, LLC (Duke, the licensee), requested changes to the Technical Specifications (TSs) for the Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2). The supplement dated May 20,2010, 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) staff's original proposed no significant hazards consideration determination as published the Federal Register on March 9,2010 (75 FR 10826). The proposed changes would revise TS 3.3.1, "Reactor Trip System (RTS) Instrumentation" and TS 1.1, "Definitions." The proposed amendments support plant modifications which would replace the existing source range and intermediate range excore detector systems with equivalent neutron monitoring systems. The new instrumentation will perform both the source range and the intermediate range monitoring functions.

2.0 REGULATORY EVALUATION

The NRC staff reviewed the proposed TS changes in the application against the regulatory requirements and guidance listed below to ensure that there is reasonable assurance that the systems and components affected by the proposed TS changes will perform their safety functions.

-2 2.1 Regulatory Requirements The NRC staff considered the following regulatory requirements:

The regulation at Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.2, "Definitions," states that: Safety-related structures, systems, and components means those structures, systems, and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary (2) The capability to shut down the reactor and maintain it in a safe-shutdown condition; or (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guidelin exposures set forth in 10 CFR 50.34(a)(1) or 10 CFR 100.11, as applicable.

In the regulation at 10 CFR 50.36, "Technical Specifications," the Commission established its regulatory requirements related to the contents of the TSs. Specifically, 10 CFR 50.36(a)(1) states, in part, that: Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. Furthermore, 10 CFR 50.36(c)(1

)(ii)(A) states, in part, the following:

Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence.

In addition, 10 CFR 50.36(c)(3) 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 limiting conditions of operation will be met.

-3 The NRC staff reviewed the proposed TS changes against these 10 CFR 50.36 requirements to ensure that there is reasonable assurance that the systems affected by the proposed TS changes will perform their required safety functions.

The regulation at 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," Criterion 10 (GDC 10), "Reactor design," addresses the requirements for reactor protection systems in that it requires that: The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 13, "Instrumentation and control," requires that: Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure 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. The NRC staff reviewed the proposed TS changes and the affected instrument setpoint calculations and plant surveillance procedures to ensure proper operation of the intermediate range and source range neutron flux instrumentation. GDC 20, "Protection system functions," requires, in part, that The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences

... The NRC staff evaluated the application to ensure that the proposed TS change will still assure that the fuel design limits and plant safety limits (SLs) specified in TS 2.0 are not exceeded, and that these SLs will not be exceeded under plant transient, Anticipated Operational Occurrences, and accident conditions. Regulatory Guidance The NRC staff considered the following regulatory guidance in their review: Regulatory Guide (RG) 1.105, "Setpoints for Safety-Related Instrumentation," Rev. 3, issued December 1999 (ADAMS Accession No. ML993560062), describes a method that the NRC staff considers acceptable for complying with the agency's regulations for ensuring that setpoints for safety-related instrumentation are initially within and remain within the TS limits. RG 1.105 endorses Part I of Instrument Society of America S67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation," subject to NRC staff clarifications.

The NRC staff used this guide to determine the adequacy of the licensee's setpoint calculation methodologies and the related plant surveillance procedures.

Technical Specifications Task Force Traveler (TSTF)-493, "Clarify Application of Setpoint Methodology for LSSS [limiting safety system settings]

Functions," Rev. 4, dated January 5, 2010 (ADAMS Accession No. ML 100060064), and an errata sheet, "Transmittal of TSTF-493, Rev. 4, Errata," dated April 23, 2010 (ADAMS Accession No. ML 101160026), clarify the application of setpoint methodology.

The NRC staff specifically verified the compliance of the proposed footnotes with the applicable functions identified in Appendix A to TSTF-493, Rev. 4, and the calculations for the revised setpoints for total loop uncertainties (TLUs), nominal trip set point (NTSP), allowable value (AV), as-found tolerance band, and as-left tolerance band. TECHNICAL EVALUATION Description of the Proposed Amendment The proposed license amendment request (LAR) includes the following changes to the TSs: The definition for NTSP in TS Section 1.1 presently states, in part: "If plant conditions warrant, the trip setpoint implemented in plant hardware may be set outside the NOMINAL TRIP SETPOINT calibration tolerance band as long as the trip setpoint is conservative with respect to the NOMINAL TRIP SETPOINT." The licensee proposed to revise this sentence to read as follows: Unless otherwise specified, if plant conditions warrant, the trip setpoint implemented in plant hardware may be set outside the NOMINAL TRIP SETPOINT calibration tolerance band as long as the trip setpoint is conservative with respect to the NOMINAL TRIP SETPOINT. The current Note 2 in Surveillance Requirement (SR) 3.3.1.11 reads as follows: "Power and Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed prior to entry into MODE 1 or 2." The proposed Note 2 reads as follows: Power Range Neutron Flux high voltage detector saturation curve verification is not required to be performed prior to entry into MODE 1 or 2. In addition, a new Note 3 reads as follows: Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed prior to entry into MODE 1 or 2*. The asterisked footnote applicable to Note 3 is proposed to read as follows:

-5This note applies to the Westinghouse Electric Company supplied compensated ion chamber neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors which do not require detector plateau voltage verification.

Therefore, this note does not apply to the fission chamber neutron detectors. Table 3.3.1-1 presently lists an AV entry of u< 31% rated thermal power (RTP)" for Function 4 in two locations.

The licensee proposed to retain these entries, append them with a new asterisked footnote, and add another AV entry adjacent to each existing entry, to read u< 38% RTP." The asterisked footnote will read as follows: The < 31% RTP AV applies to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The < 38% RTP AV applies to the replacement fission chamber Intermediate Range neutron detectors. TS Table 3.3.1-1 presently lists an AV entry of u< 1.4 E5 cps [counts per second]" for Function 5 in two locations.

The licensee proposed to retain these entries, append them with a new double-asterisked footnote, and add another AV entry adjacent to each existing entry, to read u<1.44E5 cps." The double-asterisked footnote will read as follows: The < 1.4 E5 cps AV applies to the Westinghouse-supplied boron trifluoride (BF 3) source range neutron detectors.

The BF 3 neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The < 1.44 E5 cps AV applies to the replacement fission chamber source range neutron detectors. Table 3.3.1-1 presently lists AV and NTSP entries of u> 6E-11 amp" and U1 E-1 0 amp," respectively, for Function 16, Item a. The licensee proposed to retain these entries, append them with a new triple-asterisked footnote, and add AV and NTSP entries adjacent to each existing entry, to read U> 6.6E-6% RTP" and u1 E-5% RTP," respectively.

The triple-asterisked footnote will read as follows: The> 6E-11 amp AV and the 'I E-10 amp NOMINAL TRIP SETPOINT value apply to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The> 6.6E-6% RTP AV and the 15% RTP NOMINAL TRIP SETPOINT value apply to the replacement fission chamber Intermediate Range neutron detectors. The licensee proposed to add two new lettered footnotes, designated (I) and (m), to Table 3.3.1-1. The two new footnotes would apply to the cross-referenced channel

-6 operational test and channel calibration requirements listed in the SRs column of the table for Functions 4 and 5, specifically the entries for SRs 3.3.1.7, 3.3.1.8, and 3.3.1.11 for which AVs and NTSPs are applicable.

The new footnotes read as follows: If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service. The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NTSP at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field setting) to confirm channel performance.

The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR [Updated Final Safety Analysis Report]. Evaluation of Proposed Modifications The RTS consists of all components, from the Field-mounted process instrumentation (e.g., transmitters, resistance temperature detectors, neutron detectors, etc.) to the reactor trip switchgear, whose functioning is required to initiate a reactor trip when required.

The RTS includes portions of the excore nuclear instrumentation system, which consists of three discrete but overlapping ranges: source range, intermediate range, and power range. Because of issues related to reliability and parts obsolescence, the licensee is replacing the existing Westinghouse-supplied source range and intermediate range excore detector systems for both the McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2) and Catawba 1 and 2 with Thermo Scientific-supplied 300i Neutron Flux Monitoring Systems. The existing source range and intermediate range excore detector systems use BF 3 detectors and compensated ion chamber detectors, respectively.

The new instrumentation will use fission chamber detectors to perform both the source range and intermediate range monitoring functions.

The licensee stated that the modification to replace the source range and intermediate range detectors will not affect any function related to postaccident monitoring and, depending upon the exact alarm setting, may provide improved notification for boron dilution mitigation with an earlier actuation of the high flux at shutdown alarm. The licensee also stated that the proposed TS modifications represent the pre-replacement and post-replacement requirements, since the plant modifications to replace the detectors will occur during separate refueling outages for each unit. After full implementation of all the modifications, a future LAR will be submitted to amend the TSs to show only the replacement requirements.

The licensee, also, stated that within 1 year following the implementation of the modification for the final unit, Duke will submit a follow-up administrative LAR to delete the superseded TS requirements.

-7 The licensee stated that the new Thermo Scientific equipment is compatible with the rest of the nuclear instrumentation and reactor protection systems and will perform the same functional requirements of the equipment that it will replace. However, it will differ in the following aspects from the existing equipment:

Source Range Scale: The source range indication scale will change from 100-106 cps (six decades) to 10-1 106 cps (7 decades).

Source Range High Flux at Shutdown Alarm: The alarm setpoint is updated, when requested; is electronically established based on a selectable ratio of 1.5 to 4 times steady state; and is automatically reduced as state count rate decreases.

For the existing source range instrumentation, the adjustment is manually established at approximately 5 times steady state (McGuire 1 and 2) or 3 times steady state (Catawba 1 and 2). Intermediate Range Scale Units: Intermediate Range scale units will change from amperes to percent power. Intermediate Range Scale: The Intermediate Range indication scale will change from 10.11 3 amps (eight decades) to 10-8-200 percent RTP (over 10 decades).

Source Range Deenergization:

With the existing Westinghouse system, the source range indication is disabled by deenergizing high voltage to the source range detectors when the source range trip is blocked upon receipt of the permissive P-6. This is done in order to prevent damage to the BF 3 detectors as a result of operation beyond their design limits. The removal of high voltage from the Thermo Scientific fission chamber detectors is not required.

They will remain energized through all levels of operation.

Detector Plateau Curve Calibration:

The Thermo Scientific fission chambers do not require detector plateau curves to be obtained as part of the channel calibration.

The fission chambers operate in the ionization chamber region of the detector ionization curve. The pulse output of the detectors is not dependent on the applied voltage over a wide range of voltage. The fission chambers are operated at a fixed high voltage. The power range detectors will remain a Westinghouse installation with vendor-recommended saturation curve testing. Thermo Scientific does not require periodic saturation or plateau curve testing for fission chamber detectors.

-8 The change in units for the intermediate range scale will result in a change in the value of the P-6 setpoint from its present value in amperes to the equivalent value in percent power. The change in the detector output, together with the change in intermediate range units, requires a verification of the coordination between the source range neutron flux trip and the P-6 setpoints.

The source range neutron flux trip setpoint and the P-6 permissive are set relative to the overlap between the source range and intermediate range scales. The P-6 permissive is selected such that its bistable trips after the intermediate range indication comes on scale (so intermediate range operation can be verified) and before the source range indication goes off scale (within the overlap region of the instruments).

The source range neutron flux trip setpoint is also within this overlap region. The source range neutron flux trip setpoint is set between the P-6 permissive and the upper range of the source range scale. The source range trip setpoint must be set sufficiently above the P-6 value in order to allow the operator time to block the source range trip and at the same time must be below the maximum range of the source range indication.

For the source range function, only the relative change from a baseline value is important and not an absolute value of neutron flux. The licensee stated that calculations prepared for this modification verified the correct correlation between the source range neutron flux trip and the P-6 permissive setpoints for the new instrumentation.

The planned plant modification extends the lower end of the intermediate range by two additional decades. The previous Westinghouse intermediate range P-6 setpoint was established at one decade overlap (1 x1 0.10 amps). The proposed P-6 setpoint of "1 E-5% RTP" provides three decades of overlap to ensure adequate margin to the source range trip setpoint to allow the operator time to actuate the source range neutron flux trip block signal and at the same time ensure a conservative signal overlap with the intermediate range indication.

This approach, which provides additional margin, is acceptable to the NRC staff. Also described above, the scope of the proposed changes includes the addition of two lettered footnotes applicable to the affected source range listed in Table 3.3.1-1. The NRC staff finds that these footnotes are consistent with TSTF-493 and, therefore, are acceptable.

The licensee performed setpoint calculations for the modifications, resulting in the need for changes to the associated values listed in TS Table 3.3.1-1, as described in Section 1.0 above. The licensee performed these setpoint calculations in accordance with Duke Energy Engineering Directives Manual (EDM)-102, "Instrument SetpointiUncertainty Calculations," Rev. 3. The licensee stated that the methodology described in EDM-102 is consistent with the intent of Instrument Society of America Standard RP67.04-1994, Part II, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation," and RG 1.105, Rev. 3. Basic Methodology-EDM-102 The licensee stated that the previous setpoint calculation based on EDM-102 applied the percent of span accuracy values for source range and intermediate range channels linearly to the range of 0 to 1x10 6 cps and 0 to 120 percent RTP, respectively.

This method results in overly conservative AVs for the source range and intermediate range channels for a given channel accuracy.

Applying the accuracies this way results in AVs more restrictive than the design capabilities of the instrumentation.

This could require calibration checks more frequently than required by the TSs to ensure compliance.

-9 The new loop uncertainty methodology is primarily based on the squares (SRSS) technique for combination of random-independent uncertainty terms. dependent and bias uncertainty terms are addressed through a combination of the SRSS and algebraic techniques.

The licensee's setpoint calculation included determination of SL, analytical limit (AL), TLUs, NTSP, AV, as-found tolerance band, and as-left tolerance band. In response to an NRC staff request for additional information (RAI) sent by letter dated May 20, 2010, the licensee provided calculations on TLU, AV, NTSP, as-found tolerance band, and as-left tolerance band. The licensee used engineering units in percentage of span in calculating the tolerances and converted them appropriately to account for the logarithmic nature and the seven decades of the scale. Safety Limits and Analytical Limits SLs are the values chosen to reasonably protect the integrity of physical barriers that guard against the uncontrolled release of radioactivity.

Typically, ALs are values used in the safety analyses that were specifically chosen to allow the equipment time to act and prevent exceeding the SLs. The licensee stated that the source range and intermediate range neutron flux trips and the P-6 interlock are not explicitly credited in any design-basis accidents.

Only the power range low setpoint trip of 25-percent RTP is credited for actuating to mitigate the uncontrolled rod cluster control assembly withdrawal from a subcritical or low-power startup accident, as described in Sections 7.2 and 15.4 of the UFSARs for both McGuire 1 and 2 and Catawba 1 and 2. The source range neutron flux trip does provide a diverse trip function in subcritical modes to help ensure that the UFSAR analysis of this event remains bounding, but its function is not explicitly credited.

Since the source range and intermediate range neutron flux trips are not explicitly credited in the accident analyses, no AL has been established for use in the accident analysis.

However, the licensee designated the following AL values for use in the setpoint calculations:

Plant Instrument AL Value McGuire 1 and 2 Source Range 6.0E5 cps Catawba 1 and 2 Source Range 6.0E5 cps McGuire 1 and 2 Intermediate Range 155% RTP Catawba 1 and 2 Intermediate Range 160% RTP An AL is not applicable for the P-6 interlock function.

Total Loop Uncertainty The licensee calculated TLU, converting error in percent span accounting for logarithmic nature and 7 decades of the scale, as follows: TLU =+/ -[RU NT 2+ RU/]1/2 + biases, where: NT =uncertainty associated with the portion of the loop not tested during the channel check, calibration, etc.

-T = uncertainty associated with the portion of the loop tested during the channel check, calibration, etc. RU =total random uncertainty The calculated TLUs are as follows: Plant Instrument TLU McGuire 1 and 2 Source Range 4.66E5 cps Catawba 1 and 2 Source Range 4.66E5 cps McGuire 1 and 2 Intermediate Range 154.08% RTP Catawba 1 and 2 Intermediate Range 157.89% RTP McGuire 1 and 2 P-6 1.62E-6% RTP Catawba 1 and 2 P-6 1.58E-6% RTP Nominal Trip Setpoints The NTSP is the value at which the trip or actuation is intended to occur. The licensee used the following equation to calculate the NTSP: NTSP =AL +/- (TLU + Margin) The calculated NTSPs are as follows: Plant Instrument NTSP McGuire 1 and 2 Source Range 1.0E5 cps* Catawba 1 and 2 Source Range 1.0E5 cps* McGuire 1 and 2 Intermediate Range 25% RTP* Catawba 1 and 2 Intermediate Range 25% RTP* McGuire 1 and 2 P-6 1E-5% RTP Catawba 1 and 2 P-6 1E-5% RTP *This LAR does not propose changes to these values. Allowable Value The licensee stated that for McGuire and Catawba, the AV represents an acceptable benchmark (specified by TS) within which periodic calibrations/checks must fall to ensure operability.

When a channel as-found condition is determined to be less conservative than the AV, the channel must be declared inoperable.

The licensee's AV determination is based on expected uncertainty influences for the portion of the loop tested, including reference accuracy.

calibration uncertainty, representative uncertainty for temperature variations between calibrations, representative drift over surveillance interval, and other factors. The licensee selected AV based on the most conservative AV of either EDM Method 1 or 2:

-11 EDM METHOD 1 AV =SP + RUT-cal. where: SP =nominal setpoint T-cal =representative (minimum) uncertainty term magnitudes associated with the portion of the loop tested and for the desired interval (attributed to the expected variation from as-left conditions)

EDM METHOD 2 AV =AL + RUNT =AL + {[(TLU -Biases)2 -RT T_c il'/2 + Biases} The AV for each setpoint is calculated using the two EDM methods described above. The more conservative calculated value for the two methods is then used as the AV. Although the accuracy of the new instrumentation is better than the existing instrumentation, the net result of applying the rack uncertainties logarithmically is an increase in the source range and intermediate range AVs. The calculated AVs are as follows: Plant Instrument McGuire 1 and 2 Source Range <1.44E5 Catawba 1 and 2 Source Range <1.44E5 McGuire 1 and 2 Intermediate Range Catawba 1 and 2 Intermediate Range <38% McGuire 1 and 2 P-6 >6.6E-6% Catawba 1 and 2 P-6 >6.6E-6% As-Found Tolerance The licensee included reference accuracy, drift, setting tolerance, and measurement and test equipment tolerance in calculating the as-found tolerance.

The calculated as-found tolerances for the source range and intermediate range channels are as follows: Plant Instrument As-Found McGuire 1 and 2 Source Range <2.25% Catawba 1 and 2 Source Range <2.25% McGuire 1 and 2 Intermediate Range <1.82% Catawba 1 and 2 Intermediate Range <1.82% The above as-found tolerances are given in percent span and have been converted to cps or RTP about the previous surveillance as-left value to obtain the as-found tolerance in cps or RTP on the logarithmic scale for these channels.

-12As-Left Tolerance "As left" is the condition in which a channel, or portion of a channel, is left after calibration or final device setpoint verification.

The as-left tolerance is the acceptable setting variation about the setpoint that the technician may leave following calibration.

The size of the setting or as-left tolerance is generally based on the reference accuracy and limitations of the technician in adjusting the module (measurement and test equipment and reading resolution).

The licensee uses previous calibration or surveillance as-left setting value for a channel as the starting point for determining if the next surveillance as-found tolerance is met. The NRC staff finds that the licensee performed the setpoint calculations in conformance with RG 1.105 and TSTF-493 and hence the proposed TS changes in Section 1.0 complies with the requirements of 10 CFR 50.36 specified in Section 2.0, and therefore.

they are acceptable.

3.3 Technical Evaluation Summary The NRC staff finds that the proposed TS changes comply with the regulatory requirements specified in Section 2.0 and are consistent with the acceptable methodology described in TSTF-493.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments.

The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change SRs. 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 10826). Accordingly, the amendments meet 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 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 in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor:

S. Mazumdar, NRR Date: August 2,2010 J. Morris -2 If you have any questions, please call me at Sincerely, IRA! Jon Thompson, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 258 to NPF-35 2. Amendment No. 253 to NPF-52 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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