ML15320A333

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Issuance of Amendments Regarding Non-Conservative Technical Specifications Allowable Value
ML15320A333
Person / Time
Site: Catawba  Duke energy icon.png
Issue date: 12/18/2015
From: Geoffrey Miller
Plant Licensing Branch II
To: Henderson K
Duke Energy Carolinas
Miller G
References
CAC MF5293, CAC MF5294
Download: ML15320A333 (27)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 18, 2015

  • Mr. Kelvin Henderson Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1AND2, ISSUANCE OF AMENDMENTS REGARDING NON-CONSERVATIVE TECHNICAL*

SPECIFICATION ALLOWABLE VALUE (CAC NOS. MF5293 AND MF5294)

Dear Mr. Henderson:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 277 to Renewed Facility Operating License NPF-35 and Amendment No. 273 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.

November 24, 2014 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML14330A327), as supplemented by letters dated January 15, 2015 (ADAMS Accession No. ML15020A015), July 31, 2015 (ADAMS Accession No. ML15217A008),

August 17, 2015 (ADAMS Accession No. ML15231A012), and October 23, 2015 (ADAMS Accession No. ML15300A533).

The amendments revises the TSs to correct non-conservative setpoints. Specifically, the Allowable Value and Nominal Trip Setpoint for the Auxiliary Feedwater Loss of Offsite Power (Function 6.d) is modified. Additionally, the values in the associated Surveillance Requirement 3.3.5.2 are modified to the same values. As part of the change, the amendment also adds the

. applicable footnotes in accordance with TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS 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.

K. Henderson If you have any questions, please call me at 301-415-5888.

s~J11 G. Edward Miller, 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. 277 to NPF-35
2. Amendment No. 273 to NPF-52
3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION 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. 277 Renewed License No. NPF-35

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

A The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-35 filed by Duke Energy Carolinas, LLC, acting for itself, and North Carolina Electric Membership Corporation (licensees), dated November 24, 2014, as supplemented by letters dated January 15, 2015, July 31, 2015, August 17, 2015, and October 23, 2015, complies with the standards a.nd 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.

Enclosure 1

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 Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 277 , 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. /'

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULA TORY COMMISSION

~--z:~

Michael T. Markley, Chief Plant Licensing Branch li-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: December 1 8, 2O1 5

)

UNITED STATES NUCLEAR REGULATORY COMMISSION 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. 273 Renewed License No. NPF-52

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

A The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-52 filed by Duke Energy Carolinas, LLC, acting for itself, North Carolina Municipal Power Agency No. 1 and Piedmont Municipal Power Agency (licensees), dated November 24, 2014, as supplemented by letters dated January 15, 2015, July 31, 2015, August 17, 2015, and October 23, 2015, 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;

8. 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 The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

Enclosure 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 Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 273 , 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.

3. 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

~-Zfa~

Michael T. Markley, 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: December 1 8, 201 5

ATTACHMENT TO LICENSE AMENDMENT NO. 277 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 273 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 3.3.2-13 3.3.2-13 3.3.2-14 3.3.2-14 3.3.2-15 3.3.2-15 3.3.2-16 3.3.2-16 3.3.2-17 3.3.2-17 3.3.2-18 3.3.5-2 (1 3.3.5-2

(2) , Technical Specifications Jhe Technical Specifications contained in Appendix A, as revised through Amendment No. 277, 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.

(3) 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 90.71 (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.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license. *

(5) 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 approvedJire 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 following 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. 277

(2) Technical Specifications

  • The Technical Specifications contained in Appendix A, as revised through Amendment No. 273, 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.

(3) 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, ahd 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.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) 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 following 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-52 Amendment No. 273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 6)

Engineered Safety Feature Actuation System Instrumentation I

APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Safety lnjection<bl
a. Manual initiation 1,2,3,4 2 B SR 3.3.2.8 NA NA
b. Automatic 1,2,3,4 2 trains c SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays
c. Containment 1,2,3 3 D SR 3.3.2.1  ::; 1.4 psig 1.2 psig Pressure - High SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10
d. Pressurizer 1,2,3(a) 4 D SR 3.3.2.1  ;,: 1839 psig 1845 psig Pressure - Low *SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10
2. Deleted.
3. Containment lsolation<bl
a. Phase A Isolation

\

(1) Manual 1,2,3,4 2 B SR 3.3.2.8 NA NA Initiation (2) Automatic 1,2,3,4 2 trains c SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (3) Safety Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Injection (a) Above the P-11 (Pressurizer Pressure) interlock.

(b} The requirements of this Function are not applicable to Containment Purge Ventilation System and Hydrogen Purge System components,.

since the system containment isolation valves are sealed closed in MODES 1, 2, 3, and 4.

Catawba Units 1 and 2 3.3.2-13 Amendment Nos. 277 , 273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 6)

Engineered Safety Feature Actuation System Instrumentation I

MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION. CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

3. Containment Isolation (continued)
b. Phase B Isolation (1) Manual Initiation 1,2,3,4 1 per train, B SR 3.3.2.8 NA NA 2 trains (2) Automatic 1,2,3,4 2 trains c SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (3) Containment 1,2,3 4 E SR 3.3.2.1 S; 3.2 psig 3.0 psig Pressure SR 3.3.2.5 High High . SR 3.3.2.9 SR 3.3.2.10
4. Steam Line Isolation
a. Manual Initiation (1) System 1.2(bl,3(b) 2 trains F SR 3.3.2.8 NA NA (2) Individual 1,2(bl,3(b) 1 per line G SR 3.3.2.8 NA NA
b. Automatic 1,2(bl,3(b) 2 trains H SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays C. Containment 1,2(bl,3(b) 4 E SR 3.3.2.1  :::; 3.2 3.0 psig Pressure - High SR 3.3.2.5 psig High SR 3.3.2.9 SR 3.3.2.10
d. Steam Line Pressure (1) Low 1,2(bl,3(a)(b) 3 per steam D SR 3.3.2.1  ;::: 744 psig 775 psig line SR 3.32.5 SR 3.3.2.9 SR 3.3.2.10 (continued)

(a)Above the P-11 (Pressurizer Pressure) interlock.

(b)Except when all MSIVs are closed and de-activated.

Catawba Units 1 and 2 3.3:2-14 Amendment Nos. 277, 273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS. VALUE SETPOINT

4. Steam Line Isolation (continued)

(2) Negative 3 (b)(c) 3 per steam D SR 3.3.2.1 s: 122.3(d) 1oo(d) psi Rate High line SR 3.3.2.5 psi SR 3.3.2.9 SR 3.3.2.10

5. Turbine Trip and Feedwater Isolation
a. Turbine Trip (1) Automatic 1,2 2 trains SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (2) SG Water 1,2 4 perSG J SR 3.3.2.1  :;; 85.6% 83.9%

Level- SR 3.3.2.2 (Unit 1) (Unit 1)

High-High SR 3.3.2.4  :::_78.9% 77.1%

(P-14) SR 3.3.2.5 (Unit2) (Unit 2)

SR 3.3.2.6 SR 3.3.2.9 SR 3.3.2.10 (3) Safety Refer to Function 1 (Safety Injection) for all initiation functions and requirements. See Injection Item 5.a.(1) for Applicable MODES.

b. Feedwater Isolation

( 1) Automatic 2 trains H SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Rela s (continued)

(b) Except when all MSIVs are closed and de-activated.

(c) Trip function automatically blocked above P-11 (Pressurizer Pressure) interlock and may be blocked below P-11 when Steam Line Isolation Steam Line Pressure - Low is not blocked.

(d) Time constant utilized in the rate/lag controller is 50 seconds.

(e) Except when all MFIVs, MFCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

\

Catawba Units 1 and 2 3.3.2-15 Amendment Nos. 277 , 273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 4 of 6)

Engineered Safety Feature Actuation System Instrumentation

  • APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT (2) SG Water 4 per SG D SR 3.3.2.1 s;85.6% 83.9%

Level- High SR 3.3.2.2 (Unit 1) (Unit 1)

High (P-14) SR 3.3.2.4  :::_78.9% 77.1%

SR 3.3.2.5 (Unit 2) (Unit 2)

SR 3.3.2.6 SR 3.3.2.9 SR 3.3.2.10 (3) Safety Refer to Function 1 (Safety Injection) for all initiation functions and requirements. See lnjectioR Item 5.b.(1) for Applicable MODES.

(4) Tavg-Low 4 J SR 3.3.2.1 2: 561°F 564°F SR 3.3.2.5 SR 3.3.2.9 coincident with Refer to Function 8.a (Reactor Trip, P-4) for all initiation functions and requirements.

Reactor Trip, P-4 (5) Doghouse 3 per train L SR 3.3.2.8 s 12 inches 11 inches WaterLevel - per SR 3.3.2.9 above 577 ft above 577 High High doghouse SR 3.3.2.12 floor level ft floor level (e) Except when all MFIVs, MFCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

l Catawba Units 1 and 2 3.3.2-16 Amendment Nos. 277,273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 6)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

6. Auxiliary Feedwater
a. Automatic 1,2,3 2 trains H SR 3.3.2.2 NA NA
b. SG Water Level 1,2,3 4 per SG D SR 3.3.2.1 ~9%. 10.7%

- Low Low SR 3.3.2.5 (Unit 1) (Unit 1)

SR 3.3.2.9 ~ 35.1% 36.8%

SR 3.:p.10 (Unit 2) (Unit 2)

c. Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.
d. Loss of Offsite 1,2,3 3 per bus D SR 3.3.2.3(f)(gJ ~ 3396 v 3450V Power SR 3.3.2.9(0(gJ SR 3.3.2.10
e. Trip of all Main 1,2 3 per pump K SR 3.3.2.8 NA NA Feedwater SR 3.3.2.10 Pumps
f. Auxiliary 1,2,3 3 per train M SR 3.3.2.8 A)~ 9.5 psig A) 10.5 Feedwater Pump SR 3.3.2.10 psig Train A and Train B Suction B) ~ 5.2 psig B) 6.2 psig Transfer on (Unit 1) (Unit 1)

Suction ~ 5.0 psig 6.0 psig Pressure - Low (Unit 2) (Unit 2)

(continued)

(f) 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.

(g)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 3.3.2-17 Amendment Nos. 277,273

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 6 of 6)

Engineered Safety Feature Actuation System Instrumentation MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

7. Automatic Switchover to Containment Sump
a. Automatic 1,2,3,4 2 trains c SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays
b. Refueling Water 1,2,3.4 4 N SR 3.3.2.1 2 91 9 inches 95inches Storage Tank SR 3.3.2.7(a)(b)

(RWST) Level - SR 3.3.2.9(a)(b>

Low SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection .

8. ESFAS Interlocks
a. Reactor Trip, P-4 1,2,3 1 per train, F SR 3.3.2.8 NA NA 2 trains
b. Pressurizer 1,2,3 3 0 SR 3.3.2.5 2 1944 and 1955 psig Pressure. P-11 SR 3.3.2.9  :<> 1966 psig C. T avg - Low Low, 1.2,3 1 per loop o* SR 3.3.2.5 2 550°F 553°F P-12 SR 3.3.2.9
9. Containment Pressure Control System
a. Start Permissive 1,2,3,4 4 per train p SR 3.3.2.1 s 1.0 psid 0.9 psid SR 3.3.2.7 SR 3.3.2.9
b. Termination 1,2,3.4 4 per train p SR 3.3.2.1 0.25 psid 0.35 psid SR 3.3.2.7 SR 3.3.2.9
10. Nuclear Service 1,2,3,4 3 per pit Q,R SR 3.3.2.1 2 El. 555.4 ft . El. 557.5 ft Water Suction SR 3.3.2.9 Transfer Low Pit SR 3.3.2.11
  • Level SR 3.3.2.12 (a) 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.

(b) 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 3.3.2-:18 Amendment Nos. 277 273

LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 3. 5. 1 ----------------------------------NOTE--------------------------------

T esti ng shall consist of voltage sensor relay testing excluding actuation of load shedding diesel start, and time delay times.

Perform TADOT. In accordance with the Surveillance Frequency Control Program SR 3.3.5.2 Perform CHANNEL CALIBRATION with NOMINAL TRIP In accordance with SETPOINT and Allowable Value as follows: the Surveillance Frequency Control

a. Loss of voltage Allowable Value~ 3396 V. Program Loss of voltage NOMINAL TRIP SETPOINT =

3450 v. l

b. Degraded voltage Allowable Value~ 3738 V.

Degraded voltage NOMINAL TRIP SETPOINT =

3766 V.

Catawba Units 1 and 2 3.3.5-2 Amendment Nos.

UNITE[) STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555..0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 277 TO RENEWED FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NPF-52 DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413AND 50-414

1.0 INTRODUCTION

By application dated November 24, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14330A327), as supplemented by letters dated January 15, 2015 (ADAMS Accession No. ML15020A015), July 31, 2015 (ADAMS Accession No. ML15217A008), August 17, 2015 (ADAMS Accession No. ML15231A012), and October 23, 2015 (ADAMS Accession No. ML15300A533), 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 supplements dated January 15, 2015, July 31, 2015, August 17, 2015, and October 23, 2015, provided additional informatio,n that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published the Federal Registeron March 31, 2015 (80 FR 17085).

. The proposed amendments would modify the TSs to correct non-conservative setpoints.

Specifically, the Allowable Value and Nominal Trip Setpoint for the Auxiliary Feedwater Loss-of-Offsite Power (Function 6.d) would be modified. Additionally, the values in the ass0ciated Surveillance Requirement 3.3.5.2 would be modified to the same values. As part of the change, the licensee is also proposing to add the applicable footnotes in accordance with TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS Functions."

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production Utilization Facilities," establishes the fundamental regulatory requirements for nuclear power reactor operations. Specifically, Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 establishes the minimum necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of the public.

In 10 CFR 50.36, "Technical Specifications," the Commission established its regulatory requirements related to the contents of the TS. 10 CFR 50.36 states, "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."

Specifically, 10 CFR 50.36(c)(1 )(ii)(A) states, "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."

Additionally, 10 CFR 50.36(c), states, "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 for operation will be met."

Appendix A to 10 CFR Part 50 provides General Design Criteria (GDC) that must be considered when developing principal design criteria for a water-cooled nuclear power plant plant. Section 3.1 of the Catawba Updated Final Safety Analysis Report (UFSAR) discusses conformance with the GDC. The proposed amendment was evaluated against the following GDC, as incorporated into the Catawba licensing basis through the UFSAR:

  • GDC 13, "Instrumentation and Control," requires, in part, that instrumentation be provided to monitor variables and systems and that controls be provided to maintain these variables and systems within prescribed operating ranges.
  • GDC 17, "Electric power systems," requires, in part, that an onsite power system and an offsite electrical power system be provided with sufficient capacity and capability to permit functioning of structures, systems, and components important to safety.
  • GDC 20, "Protection System Functions," requires, in part, that the protection system be designed to initiate the operation of appropriate systems to ensure that specified acceptable fuel design limits are not exceeded.

2.2 Regulatory Guidance Regulatory Guide (RG) 1.105, "Setpoints for Safety-Related Instrumentation," Revision 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 Standard 67.04-1994, "Setpoints for Nuclear Safety-Related Instrumentation," subject to NRC staff clarifications.

In Regulatory Issue Summary (RIS) 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, "Technical Specifications," Regarding limiting Safety System Settings During Periodic Testing and Calibration of Instrument Channels," dated August 24, 2006 (ADAMS Accession No. ML051810077), the NRC addressed requirements on limiting safety system settings that are assessed during the periodic testing and calibration of instrumentation.

The Pressurized-Water Reactor and Boiling-Water Reactor Owner's Groups' TSTF-493, Revision 4, dated January 5, 2010 (ADAMS Accession No. Ml 100060064), and an errata sheet dated April 23, 2010, (ADAMS Accession No. Ml 101160026), provided information to address the NRC staff's concerns stated in RIS 2006-17.

On May 11, 2010, the NRC published a notice in the Federal Register, "Notice of Availability of the Models for Plant-Specific Adoption of Technical Specifications Task Force Traveler TSTF-493, Revision 4, 'Clarify Application of Setpoint Methodology for LSSS Functions"' (75 FR 26294 ADAMS Accession No. ML11231A714), documenting its position on the adoption of TSTF-493, Revision 4.

NUREG-0800, Chapter 8, Branch Technical Position (BTP) 8-6, originally issued as PSB-1, July 1981, "Adequacy of Station Electrical Distribution System Voltages," provides guidance related to the voltage adequacy of the offsite power source.

3.0 TECHNICAL EVALUATION

3.1 Description of the Auxiliary Power System and Undervoltage Protection Each Catawba unit has a 4160 volt (V) AC Essential Auxiliary Power System which supplies power to Class 1E loads required for safe shut down of the unit following a design basis accident. The system is divided into two completely redundant and independent trains, designated A and B, each consisting of one 4160 V switchgear assembly, three 4160/600 V transformers, two 600 V load centers, and associated loads.

Normally, each Class 1E4160 V switchgear is powered from its associated non-Class 1E train of the 6900 VAC Normal Auxiliary Power System. If normal offsite power is lost, each train of the 4160 VAC Essential Auxiliary Power System is also provided with an emergency diesel generator (EDG) to supply the Class 1E loads required to safely shut down the unit following a design basis accident.

Each of the redundant 4160 V essential buses is provided with two levels of undervoltage protection to monitor bus voltage. Each level is provided with a separate set of three undervoltage relays which are utilized in a two-out-of-three logic scheme.-

The*first level of undervoltage protection relays (also known as loss-of-voltage relays (LOV)) are set to drop out if voltage falls below a nominal voltage of 3500 V (approximately 84.1 % of normal bus voltage) and remains there for approximately 1O cycles. The 10 cycle time delay prevents false EOG starting due to power system transients. The voltage setpoint is selected such that relay operation will not be initiated during normal motor starting; however, these relays will detect loss-of-voltage and initiate action in a time consistent with the accident analysis ..

The second level of undervoltage protection relays (also known as degraded voltage relays) are set to drop out below a nominal voltage of 3766 V (approximately 90.5% of normal bus voltage).

This second level employs two time delays: the first (5 seconds) establishes the existence of a sustained degraded voltage condition and provides an annunciator alarm in the control room; the second (1 O minutes) permits corrective operator action prior to separating the Class 1E and offsite power systems. An occurrence of a safety injection signal subsequent to the first time delay would immediately separate the Class 1E and offsite power systems.

3.2 Proposed TS Changes In the LAA, the licensee stated that in 2011, a concern was documented in the Catawba Corrective Action Program regarding potential for tripping of the 4160 V [Bus] loss-of-voltage relays during a loss-of-coolant accident (LOCA) loading of the essential buses. This potential condition exists if the LOCA load sequencer were to actuate with a pre-existing degraded voltage condition on the essential buses. This condition could cause the 4160 V bus voltage to drop below loss.of-voltage relay setpoint as loads are being added, and result in what is known as a "double sequencing event". The analysis indicated that this event was possible due to the large uncertainty in the loss-of-voltage relay setpoint calculation. An evaluation of this issue, in combination with an industry evaluat~on of degraded voltage analyses, led Gatawba to conclude that the existing setting for the Allowable Value of the loss-of-voltage relays is non-conservative.

In the LAR, the licensee proposed the following TS changes:

1. Allowable Value for the loss-of-voltage relays is proposed to be changed from 3242 V to 3396 V in the following TS Sections:
  • TS 3.3.2, Table 3.3.2-1, Function 6.d - Loss of Offsite Power
2. Nominal Trip Setpoint for loss-of-voltage relays is proposed to be changed fro~ 3500 V to 3450 V in the following TS Sections:
  • TS 3.3.2, Table 3.3.2-1, Function 6.d - Loss of Offsite Power
3. Add footnotes (f) and (g) to the SRs 3.3.2.3 and 3.3.2.9 listed in TS 3.3.2, Table 3.3.2-1, Function 6.d - Loss of Offsite Power, in accordance with the TS Task Force Traveler TSTF-493, Revision 4, "Clarify Application of Setpoint Methodology for LSSS Functions."

3.3 Proposed Allowable Voltage Value and Nominal Trip Voltage Setpoint of Loss-of-Voltage Relays Currently, the Allowable Value voltage setting specified in TS 3.3.2 and TS 3.3.5 is based on a nominal 3500 V trip set point of loss-of-voltage relay minus 258 V uncertainty (3500 V - 258 V=

3242V). Based on the revised uncertainty of 56 V, the revised Allowable Value is calculated as 3500 - 56 V = 3444 V. Therefore, the licensee determined the current Allowable Value (3242 V) in TS 3.3.2 and TS 3.3.5 to be non-conservative.

The licensee performed revised Unit 1 and Unit 2 safety-related voltage analysis calculations considering the lowest value of the degraded voltage relays. These calculations indicated that the lowest expected voltage during a LOCA sequence is 3563 V and 3557 V for Unit 1 and Unit 2, respectively. Adding the new uncertainty (56 V) to the existing loss-of-voltage Nominal Trip Setpoint (3500 V) results in a maximum loss-of-voltage trip of 3556 V. The margin between the lowest expected LOCA sequence voltage and the maximum voltage of loss-of-voltage relay is only 7 V (3563 V - 3556 V) on Unit 1, and 1 V (3557 V - 3556 V) on Unit 2. Therefore, the licensee also proposed to revise the Nominal Trip Setpoint from 3500 V to 3450 V to gain an additional margin of 50 V. By lowering the Nominal Trip Setpoint, the uncertainty value is also reduced to 54 V per the licensee's calculation CNC-1381.05 0017, Revision 17, thus the proposed Allowable Value is calculated as 3396 V (3450 V - 54 V). Adding the uncertainty to the new loss-of-voltage Nominal Trip Setpoint results in a maximum loss-of-voltage trip of 3504 V (3450 V + 54 V = 3504 V). This increases the margin between the lowest LOCA sequence voltage and the maximum tripping voltage of the loss-of-voltage relays to 59 V for Unit 1 (3563 V - 3504 V) and to 53 V for Unit 2 (3557 V - 3504 V).

The licensee stated that the loss-of-voltage Nominal Trip Setpoint is selected above the 80%

rated starting voltage capabilities of the 4160 V essential motors. The nominal rating of the 4160 V motors is 4000 V. The revised proposed Nominal Trip Setpoint of 3450 V of loss-of-voltage relays is still above 80% rated voltage of motors (i.e., above 3200 V).

In the LAR, the licensee evaluated the uncertainty in its calculations for the loss-of-voltage settings. As a part of this evaluation, the licensee identified areas that were overly conservative.

The licensee determined that, based on the vendor instruction book related to loss-of-voltage relays, the inaccuracy in the relay operation at its setting is mainly due to the voltage variation and temperature variation. These two terms when combined with the potential transformer voltage, as well as measuring and test equipment inaccuracies, provide the total uncertainty.

As discussed further in Section 3.4.2 of this SE, this change resulted in the uncertainty being reduced from 7.4% to 1.6% (i.e., from 258 V to 56V based on a nominal setting of 3500V).

Additionally, the licensee stated that the loss-of-voltage relays are calibrated by adjusting and measuring the actual dropout voltage value of the relays. The licensee stated that the factory calibration dial marking has no effect on the setting of the relays and the corresponding tolerance in the relay instruction book should not have been used in its previous calculations.

3.4 NRC Staff Evaluation of TS Changes 3.4.1 Allowable Value and Nominal Trip Setpoint The NRC staff asked the licensee to confirm that the lower analytical voltage limit for loss-of-voltage relay is such that none of the safety-related, normally running, motors would stall when subjected to this voltage. In its supplement dated July 31, 2015, the licensee confirmed that current drawn by each motor at the lower analytic voltage limit was calculated and compared to the protective relay setting and it was verified that the motors would not stall or trip. The NRC staff finds the licensee's response reasonable and therefore acceptable.

The NRC staff also asked the licensee to confirm that upper analytical limit for the undervoltage relay is such that the minimum expected voltage during LOCA start of all safety-related loads remains above the voltage at which protective actions would be initiated. The NRC staff asked the licensee to provide a curve showing the minimum voltages at the 4160 V safety-related buses during the starting of LOCA loads after the safety injection signal based on the minimum switchyard voltage, and to super-impose on this curve, the analytical and reset voltage values of loss-of-voltage relay settings to demonstrate that adequate margin exists so that the motors would not trip due to the undervoltage relay during a LOCA load sequencing.

In its supplement dated August 17, 2015, the licensee provided necessary figures for Unit 1 and Unit 2. In the figure for Unit 1, the NRC staff observed that minimum voltage at 4160 V safety-related bus dips to about 87.5% based on minimum voltage. Thus, the NRC staff finds that there is an adequate margin (i.e., more than 3%, which is typically considered acceptable) between the minimum voltage at 4160 V safety-related bus and the loss-of-voltage relay setpoint (82.93 % is the nominal setpoint, which becomes 84.23% after adjustment for maximum drift value). However, the NRC staff did not find similar margin in the figure for Unit 2.

In its supplements dated August 17, 2015 and October 23, 2015, the licensee clarified that in case of Unit 2, lowest voltage occurs when the containment spray pump would start concurrently with the auxiliary feedwater pump. However, automatic start of the containment spray pump motors has been deleted as part of the Emergency Core Cooling System license amendments (Amendments 257/252 for Unit 1 and 2, respectively, which have since been implemented). The containment spray motors no longer start automatically in response to a design basis accident, they are manually started. Considering this, the voltage profile for Unit 2 will, in fact, be similar to Unit 1 with adequate margin between the minimum voltage at the 4160 V safety-related bus and the loss-of-voltage relay setpoint (82.93 % - nominal setpoint; and 84.23% after adjustment for maximum drift value): Thus, the NRG staff finds the licensee's response reasonable and therefore acceptable.

Based on above evaluation, the NRC staff finds that proposed changes to the loss-of-voltage relay settings (Allowable value and nominal trip settings) would continue to allow the loss-of-voltage relay to perform its safety function without any malfunction, and therefore ensure that the safety equipment would remain available to perform its function to meet the requirements of 10 CFR 50.36, and GDC 17 and 20.

3.4.2 Uncertainty The condition that could cause tripping the loss-of-voltage relays involves a LOCA actuation of the DG sequencers with pre-existing 4160 volt essential bus voltage at the degraded voltage lower limit (i.e., the Allowable Value). In this scenario, the licensee's calculations demonstrate that the essential bus voltage could drop to 3563 volts on Unit 1 and to 3557 volts on Unit 2.

The loss-of-voltage Nominal Trip Setpoint was previously at 3500 volts and the uncertainty was documented in calculation CNC-1381.05-00-0017, "Class 1E Diesel Protective Relaying and Sequencer Undervoltage Relay Settings" (ADAMS Accession No. ML15020A015).

The licensee's calculation included the Total Loop or Channel Uncertainty (CU) as well as the computed trip setpoint (SP), deadband and reset values for the undervoltage relay based on the CU uncertainty, and was reviewed NRC staff. The setpoint, or dropout of the loss-of-voltage relay, was calculated by converting the nominal trip setpoint (previously 3500 volts) to the equivalent 120 volt base relay setting. The previous revisions to the calculation on the CU included an overly conservative application of the calibration dial setting. The factory calibration dial markings have no effect on the setting of the relays and, therefore, the corresponding tolerance in the relay instruction book should not have been used in the calculation. Removing this tolerance was a major contributor to reducing the uncertainty for the loss-of-voltage relays.

The licensee recalculated the CU using square-root-of-the-sum of the squares method. All terms included were random independent variables with the exception of relay drift which was treated as a bias.

In the calculation, the licensee stated that the best source data for relay drift is from detailed calibration records and that the loss-of-voltage relays are calibrated by adjusting and measuring the actual dropout voltage of the relays. In a request for additional information (ADAMS Accession No. ML15132A773), the NRC staff noted the issues involved with limited instrumentation drift data as described in RG 1.105, "Setpoints for Safety-Related Instrumentation." In response to the NRC staff's request, the licensee provided the calibration results for the last 6 months of 2014, confirming that the values used in the uncertainty calculation are bounding.

Per the vendor instruction book for the loss-of-voltage relays, using the method of calibration that Catawba uses, the only relay accuracy terms necessary to include are the voltage variation and temperature variation terms. Combining these two terms with the potential transformer and measuring and test equipment terms resulted in a significantly lower uncertainty. The uncertainty was reduced from 7.4% to 1.6% (i.e., from 258 volts to 56 volts). The current Allowable Value stated in TS 3.3.2 and TS 3.3.5 for the loss-of-voltage relays is 3500 volts - 258 volts= 3242 volts. The Allowable Value (based on the new uncertainty calculation) is 3500 volts - 56 volts = 3444 volts. Therefore, the value currently in TS is non-conservative and was declared Operable but Degraded. Revising the Allowable Value for the loss-of-voltage relays is required and is the reason for this LAR. Additionally, the licensee reviewed a similar uncertainty calculation forthe degraded voltage relays (CNC-1381.05-00-0012, "4160 Volt Essential Auxiliary Power System Switchgear Relay Settings") and confirmed that a similar issue did not exist.

The licensee performed revisions to the Unit 1 and Unit 2 safety-related voltage analysis calculations. The calculations were performed at the lowest value for the degraded voltage relays. This indicated that the lowest voltage during a LOCA sequence was 3563 volts and 3557 volts for Unit 1 and Unit 2, respectively. Adding the new uncertainty (56 volts) to the existing loss-of-voltage Nominal Trip Setpoint (3500 volts) results in a maximum loss-of-voltage trip of 3556 volts. Thus, the margin to tripping the loss-of-voltage relays during a LOCA loading of the essential buses was 7 volts on Unit 1 and 1 volt on Unit 2. Although the current settings meet all required operating conditions, the licensee added additional margin to account for any future loading changes. Since the Allowable Value required a change, the licensee decided to also revise the Nominal Trip Setpoint to gain this additional margin. The licensee determined that a reduction of 50 volts in the loss-of-voltage Nominal Trip Setpoint would add appropriate margin and also would continue to meet other required loading conditions. By lowering the Nominal Trip Setpoint, the uncertainty value, per calculation CNC-1381.05-00-0017, Revision 17, (ADAMS Accession No. ML15020A015), was also reduced to 54 volts thus making the Allowable Value 3396 volts. Adding the new uncertainty to the new loss-of-voltage Nominal Trip Setpoint results in a maximum Joss-of-voltage trip of 3504 volts (3450 volts + 54 volts= 3504 volts). This increases the margin to tripping the loss-of-voltage relays during LOCA loading to 59 volts for Unit 1 and to 53 volts for Unit 2.

Results of the calculations are summarized below:

  • Nominal Trip Setpoint (NTSP) = 3450V
  • Channel Uncertainty from Calculation (below) applied to NTSP = 54V
  • Allowable Value (AV) = 3396V From Calculation CNC-1381.05-00-0017, Revision 17, {ADAMS Accession No. ML15020A015):
  • Total Loop or Channel Uncertainty (CU) = 1.54V
  • LOV Relay Setpoint = 98.5V (120V)
  • LOV Relay Deadband =approx. 3V (3% of 98.5V)

The loss-of-voltage Nominal Trip Setpoint was selected to be above the 80% rated starting voltage capabilities of the 4160 volt essential motors. The nominal rating of the 4160 volt motors is 4000 volts. 80% of 4000 volts is 3200 volts. The existing loss-of-voltage Nominal Trip Setpoint is 3500 volts. This was selected for two reasons: 1) the setpoint is > 3200 volts, and 2) the potential transformers feeding the relays have a ratio of 35/1, making the setpoint exactly on the 100 volt fixed tap of the transformer. Although the setpoint will no longer be on the fixed tap, the revised loss-of-voltage Nominal Trip Setpoint of 3450 volts is still in the acceptable range.

For events where loss of power is important, the loss of power is assumed to occur at the initiation of the event. Given that the licensee has properly accounted for uncertainties in its calculation, and that the results provide acceptable margin, the NRC staff concludes the new settings of the Nominal Trip Setpoint and the Allowable Value are acceptable.

3.4.3 Associated Surveillance Requirements In conjunction with the proposed revision to the Allowable Value and nominal trip setpoint, the licensee proposed to revise the corresponding Surveillance Requirement, SR 3.3.5.2.a. The Surveillance Requirement directs the licensee to perform a channel calibration with the nominal trip setpoint and Allowable Value at a frequency in accordance with the Surveillance Frequency Control Program. The licensee has proposed to change the values listed in SR 3.3.5.2.a to

3396V for the Allowable Value and 3450 V for the nominal trip setpoint. Thus, the NRC staff concludes that the licensee has proposed an appropriate change to the Surveillance Requirement such that it will continue to assure the necessary quality of the systems and components.

3.4.4 TSTF 493-A Implementation As part of this LAR, the licensee is incorporating TSTF-493-A, Revision 4, Option A, for this function in the TS. This TS change is made by the addition of individual surveillance note requirements to applicable instrument functions. The licensee included the channel performance Surveillance Notes as specified by TSTF-493-A for the modification to Table 3.3.2-1 for functional Unit 6.d. The second note identifies that the methodologies for calculating the as-found and the as-left tolerances will be in the Updated Final Safety Analysis Report. The NRC staff concludes that the proposed footnotes do not deviate from the approved TSTF and, as such, are acceptable.

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 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 {80 FR 17085). 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) there is reasonable assurance that 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): R. Beacom V. Goel R. Fitzpatrick Date: December 18, 2015

ML15320A333 *via SE input OFFICE NRR/LPL2-1 /PM NRR/LPL2-1 /LA DIRS/ITSB/BC DE/EE EB/BC NAME GE Miller SFigueroa RElliott JZimmerman*

DATE 12/02/15 12/01/15 12/07/15 11/20/15 OFFICE l;)EEICB OGG /NLO NRR/LPL2-1 /BC NRR/LPL2-1 /PM NAME MWaters STurk MMarkley GE Miller DATE 11/19/15 12/17/15 12/18/15 12/18/15

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