ML13043A661

Issuance of Amendment No. 270 to Revise Technical Specification 2.15 to Establish Limiting Condition for Operation Requirements for Reactor Protective System Actuation Circuits

ML13043A661
Person / Time
Site:	Fort Calhoun
Issue date:	02/28/2013
From:	Lynnea Wilkins Plant Licensing Branch IV
To:	Cortopassi L Omaha Public Power District
Wilkins L
References
TAC ME8038
Download: ML13043A661 (38)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 28, 2013 Mr. Louis P. Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station 9610 Power Lane, Mail Stop FC-2-4 Omaha, NE 68008

SUBJECT:

FORT CALHOUN STATION, UNIT NO.1-ISSUANCE OF AMENDMENT TO:

ESTABLISH THE REACTOR PROTECTIVE SYSTEM ACTUATION CIRCUITS LIMITING CONDITIONS FOR OPERATION (TAC NO. ME8038)

Dear Mr. Cortopassi:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 270 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No.1 (FCS). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 10, 2012, as supplemented by letters dated October 1, 2012 and January 22, 2013.

The amendment revises the TSs to establish the limiting condition for operation (LCO) requirements for the reactor protective system actuation circuits in TS 2.15, "Instrumentation and Control Systems," for FCS. Specifically, the amendment renumbers LCO 2.15(1) through 2,15(4) to 2.15.1 (1) through 2.15.1 (4), renumbers LCO 2.15(5) to LCO 2.15.3 with an associated Table 2-6, "Alternate Shutdown and Auxiliary Feedwater Panel Functions," and implements a new LCO 2.15.2 for the reactor protective system logic and trip initiation channels.

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

Si~4

~r Lynnea E. Wilkins, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 270 to DPR-40

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

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. DPR-40

1.

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

A.

The application for amendment by the Omaha Public Power District (the licensee), dated February 10, 2012, as supplemented by letters dated October 1, 2012, and January 22,2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

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

- 2

2.

Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changes as indicated in the attachment to this license amendment. and paragraph 3. B. of Renewed Facility Operating License No. DPR-40 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 270, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance and shall be implemented within 180 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~~~,~"--~ ~

Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. DPR-40 and Technical Specifications Date of Issuance: February 28, 2013

ATIACHMENT TO LICENSE AMENDMENT NO. 270 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Renewed Facility Operating License No. DPR-40 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

License Page REMOVE INSERT -4 Technical Specifications REMOVE INSERT TOC - Page 2 TOC - Page 2 TOC - Page 4 TOC - Page 4 TOC - Page 5 TOC - Page 5 2.15 - Page 1 2.15 - Page 1 2.15 - Page 2 2.15 - Page 2 2.15 - Page 3 2.15 - Page 3 2.15 - Page 4 2.15 - Page 4 2.15 - Page 5 2.15 - Page 5 2.15 - Page 6 2.15 - Page 6 2.15 - Page 7 2.15 - Page 8 2.15 - Page 7 2.15 - Page 9 2.15 - Page 8 2.15 - Page 10 2.15 - Page 9 2.15 - Page 11 2.15 - Page 10

2. 15 - Page 12 2.15 - Page 11 2.15 - Page 13 2.15 - Page 12 2.15 - Page 14 2.15 - Page 13 2.15 - Page 15 2.15 - Page 14 2.15 - Page 16 2.15 - Page 17 2.15 - Page 18

- 3 (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or when associated with radioactive apparatus or components; (5)

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is, subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A.

Maximum Power Level Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not in excess of 1500 megawatts thermal (rate power).

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 270 are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.

C.

Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, II submitted by letter dated May 19, 2006.

OPPD shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The OPPD CSP was approved by License Amendment No. 266.

Renewed Operating License No. DPR-40 Amendment No. 270

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 limiting Safety System Settings, Reactor Protective System 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15.1 Instrumentation and Control Systems 2.15.2 Reactor Protective System (RPS) Logic and Trip Initiation 2.15.3 Alternate Shutdown and Auxiliary Feedwater Panel 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 DELETED 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolar'!t System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,43,46,54,60,84,86, Q3,Q7,104,122,136.Hi2, 160,176,183.214,230. 236,246,248,2ea 270

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS

• TABLES TABLE OF CONTENTS TABLE SECTION 2-1 ESFS Initiation Instrumentation Setting Limits......................................................................... Section 2.14 2-2 Instrument Operating Requirements for RPS.......................................................................Section 2.15.1 2-3 Instrument Operating Requirements for Engineered Safety Features................................. Section 2.15.1 2-4 Instrument Operating Conditions for Isolation Functions......................................................Section 2.15.1 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions.................... Section 2.15.1 2-6 Alternate Shutdown and Auxiliary Feedwater Panel Functions...........................................Section 2.15.3 2-9 RCS Pressure Isolation Valves.................................................................................................. Section 2.1 2*10 Post-Accident Monitoring, Instrumentation Operating Limits................................................... Section 2.21 2-11 RPS Limiting Safety System Settings...................................................................................... Section 2.13 3-1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS.......................................Section 3.1 3-2 Minimum Frequencies for Checks, Calibrations and Testing of...............................................Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks, Calibrations, and Testing................................................... Section 3.1 of Miscellaneous Instrumentation and Controls 3-3a Minimum Frequency for Checks, Calibrations and Functional.................................................. Section 3.1 Testing of Alternate Shutdown Panels (AI-185 and AI-212) ?lnd Emergency Auxiliary Feedwater Panel (AI-179) Instrumentation and Control Circuits 3-4 Minimum Frequencies for Sampling Te8ts.................................................................................Section 3.2 3-5 Minimum Frequencies for Equipment Tests.............................................................................. Section 3.2 3-6 Reactor Coolant Pump Surveillance.......................................................................................... Section 3.3 5.2-1 Minimum Shift Crew Composition.............................................................................................. SectiOn 5.0 TOC - Page 4 Amendment No. 1Hi, 139,1 eg, 246,248, a.&a 270

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL QRDER)

TABLE DESCRIPTION SECTION 2-6 Altemate Shutdown and Auxiliary Feedwater Panel Functions...........................................Section 2.15.3 2-1 ESFS Initiation Instrumentation Setting Limits......................................................................... Section 2.14 2-4 Instrument Operating Conditions for Isolation Functions...........................................,.......... Section 2.15.1 2-2 Instrument Operating Requirements for RPS.......................................................................Section 2.15.1 2-3 Instrument Operating Requirements for Engineered Safety Features................................. Section 2.15.1 2-5 Instrumentation Operating Requirements for Other Safety.................................................Section 2.15.1 Features Functions 3-3a Minimum Frequency for Checks. Calibrations and Functional.................................................. Section 3.1 Testing of Altemate Shutdown Panels (AI-185 and AI-212) and Emergency Auxiliary Feedwater Panel (AI-179) Instrumentation and Control Circuits 3-2 Minimum Frequencies for Checks, Calibrations and Testing of...............................................Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks. Calibrations. and Testing..................................................Section 3.1 of Miscellaneous Instrumentation and Controls 3-1 Minimum Frequencies for Checks. Calibrations,...................................................................... Section 3.1 and Testing of RPS 3-5 Minimum Frequencies for Equipment Tests.............................................................................. Section 3.2 3-4 Minimum Frequencies for Sampling Tests................................................................................. Section 3.2 5.2-1 Minimum Shift Crew Composition.............................................................................................. Section 5.0 2-10 Post-Accident Monitoring Instrumentation Operating Limits................................................... ;Section 2.21 2-9 RCS Pressure Isolation Valves.................................................................................................. Section 2.1 3-6 Reactor Coolant Pump Surveillance.......................................................................................... Section 3.3 2-11 RPS Limiting Safety System Settings...................................................................................... Section 2.13 TOC - Page 5 Amendment No. 116,125,142,16Q, 2e2 270

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15.1 Instrumentation and Control Systems Applicability Applies to plant instrumentation systems.

Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specifications The operability, permissible bypass, and Test Maintenance and Inoperable bypass specifications of the plant instrument and control systems shall be in accordance with Tables 2*2 through 2*5.

(1)

In the event the number of channels of a particular system in service falls one below the total number of installed channels, the inoperable channel shall be placed in either the bypassed or tripped condition within one hour if the channel is equipped with a bypass switch, and eight hours if jumpers or blocks must be installed in the control Circuitry. The inoperable channel may be brPassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of discovering loss of operability; however, i the inoperability is determined to be the result of malfunctioning RTDs or nuclear detectors supplying Signals to the high power level, thermal margin/low pressurizer pressure, and axial power distribution channels, these channels may be bypassed for up to 7 days from time of discovering loss of operability. If the inoperable channel is not restored to OPERABLE status after the allowable time for bypass, it shall be placed in the tripped position or, in the case of malfunctioning RTDs or linear power nuclear detectors, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If active maintenance and/or surveillance testing is being performed to return a channel to active service or to establish operability, the channel may be bypassed during the period of active maintenance and/or surveillance testing. This specification appli~s to the high rate trip-wide range log channel when the plant is at or above 10' % power and is operating below 15% of rated power.

(2)

In the event the number of channels of a particular system in service falls to the limits given in the column entitled "Minimum Operable Channels," one of the inoperable channels must be placed in the tripped position or low level actuation permissive position for the auxiliary feedwater system within one hour, if the channel is equipped with a bypass switch, and within eight hours if jumpers or blocks are required; however, if minimum operable channel conditions for SIRW tank low signal are reached, both inoperable channels must be placed in the bypassed condition within eight hours from time of discovery of loss of operability.

If at least one inoperable channel has not been restored to OPERABLE status after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of discovering loss of operability, the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, operation can continue without containment ventilation isolation signals available if the containment ventilation isolation valves are closed.

2.15 - Page 1 Amendment No. 8,20,S4,ee,88,108,194,208, a49;~

270

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15.1 Instrumentation and Control Systems (Continued)

If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of initiating a hot shutdown procedure at least one inoperable engineered safety features or isolation functions channel has not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This specification aPilied to the high rate trip-wide range log channel when the plant is at or above 10 % power and is operating below 15% of rated power.

(3)

In the event the number of channels on a particular engineered safety features (ESF) or isolation logic subsystem in service falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," sufficient channels shall be restored to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> so as to meet the minimum limits or the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; however, operation can continue without containment ventilation isolation signals available if the ventilation isolation valves are closed. If after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of initiating a hot shutdown procedure sufficient channels have not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(4)

In the event the number of channels of those particular systems in service not described in (3) above falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled npermissible Bypass Conditions," the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If minimum conditions for engineered safety features or isolation functions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovering loss of operability, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the number of OPERABLE high rate trip-wide range log channels falls below that given in the column entitled "Minimum Operable Channels" in Table 2-2 and the reactor is at or above 10.4% power and at or below 15% of rated power, reactor critical operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation may proceed.

If during power operation, the rod block function of the secondary CEA position indication system and rod block circuit are inoperable for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the plant computer POlL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off position except when manual motion of CEA Group 4 is required to control axial power distribution.

2.15 - Page 2 Amendment No. S, 2Q,54,65.S8,125,157,194, 208, 249 270

TECHNICAL SPECIFICATIONS 2.0 2.15 2.15.2 LIMITING CONDITIONS FOR OPERATION Instrumentation and Control Systems Reactor Protective System (RPS) Logic and Trip Initiation Applicability Applies to the operational status of RPS Logic and Trip Initiation channels in MODES 1 and 2; and, When reactor coolant temperature (Tcold) is greater than 210°F or MODE 4 with more than one CEA capable of being withdrawn and RCS boron concentration less than REFUELING BORON CONCENTRATION.

Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specification Six channels of RPS Logic matrices, four channels of RPS Trip Initiation Logic and two channels of RPS Manual Trip shall be OPERABLE.

Required Actions (1)

With one RPS Logic Matrix channel inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(2)

With one RPS Trip Initiation Logic channel inoperable, de-energize the affected clutch power supply within one hour.

(3)

With one RPS Manual Trip channel inoperable, restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3.

(4)

With two RPS Trip Initiation Logic channels affecting the same trip leg inoperable, de-energize the affected clutch power supplies immediately.

(5)

With the required actions of (1), (2), or (4) not met, or with two RPS Manual Trip channels inoperable, or with two or more RPS Logic Matrices inoperable, or with two or more RPS Trip Initiation Logic channels inoperable for reasons other than (4):

a. be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, OR

b. verify reactor coolant boron concentration is at REFUELING BORON CONCENTRATION within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.15 - Page 3 Amendment No.--aoo, 24Q 270

TECHNICAL SPECIFICATIONS 2.0*

LIMITING CONPITIONS FOR OPERATION 2.15 Instrumentation and Control Systems 2.15.3 Alternate Shutdown and Auxiliary Feedwater Panel Applicability Applies to the operational status of Alternate Shutdown and Auxiliary Feedwater Panel Functions in MODES 1 and 2.

Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specification The Alternate Shutdown and Auxiliary Feedwater Panel Functions/Instrumentation or Control Parameters in Table 2-6 shall be OPERABLE.

Required Actions (1)

With the number of OPERABLE channels or control circuits less than the required number of channels, restore the required number of channels to OPERABLE within seven (7) days.

(2)

With the required actions of (1) not met, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.15 - Page 4 Amendment No. 88,129,192,173,194, 208,249 270

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis During plant operation, the complete instrumentation systems will normally be in service.

This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the reactor protective system (RPS) and engineered safety features (ESF) system when one or more of the channels are out of service. Reactor safety is provided by RPS, which automatically initiates appropriate action to prevent exceeding established I

limits. Safety is not compromised, however, by continued operation with certain instrumentation channels out of service since provisions were made for this in the plant design.

The RPS and most engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived circuits in the ESF logic system.

When one of the four channels is taken out of service for maintenance, RPS logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the removed channel. If the bypass is not effected, the out-of-service channel (Power Removed) assumes a tripped condition (except high rate-of-change of power, high power level and high pressurizer pressure},(l) which results in a one-out-of-three channel logic. If in the 2-out-of-4 logic system of the RPS one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is 1-out-of-2. At rated power, the minimum OPERABLE high-power level channel is 3 in order to provide adequate power tilt detection. If only 2 channels are OPERABLE, the reactor power level is reduced to 70%

rated power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts and from exceeding dropped CEA peaking factors.

An RPS Logic matrix channel consists of two matrix power supplies, four matrix relays and their associated contacts as well as all interconnecting wiring. An RPS Trip Initiation Logic channel consists of an M contactor and associated contacts, an interposing relay and all interconnecting wiring. Two RPS Trip Initiation Logic channels associated with the same pair of CEDM clutch power supplies are considered to affect the same trip leg.

Integrated into the trip initiation logic are two RPS Manual Trip channels. Manual Trip #1 operates by directly de-energizing all four M contactors in response to the operation of a manual pushbutton. Manual Trip #2 operates by de-energizing an undervoltage relay which results in the opening of two circuit breakers, CB-AB and CB-CD, which supply power to the CEDM clutch power supplies. Manual Trip channel #1 consists of manual trip pushbutton #1 and interconnecting wiring. Manual Trip channel #2 consists of manual trip pushbutton #2, circuit breakers CB-AB and CB-CD, and associated interconnecting wiring.

With one manual reactor trip channel inoperable, operation may continue until the reactor is shut down for other reasons. No safety analyses assume operation of the Manual trip.

Because of this, the Required Action is to restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3 during the next plant startup.

2.15 - Page 5 Amendment No. 88.125.152.173.194.208.249 270 TSBC

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

The ESF logic system is a Class 1 protection system designed to satisfy the criteria of IEEE 279, August 1968. Two functionally redundant ESF logic subsystems "A" and "B" are '

provided to ensure high reliability and effective in-service testing. These logic subsystems are designed for individual reliability and maximum attainable mutual independence both physicalry and electrically. Either logic subsystem acting alone can automatically actuate engineered safety features and essential supporting systems.

All Engineered Safety Features are initiated by 2-out-of-4 logic matrices except containment high radiation which operates on a 1-out-of-2 basis. The number of installed channels for Containment Radiation High Signal (CRHS) is two. CRHS isolates the containment pressure relief, air sample and purge system valves.

Entry into Technical Specification 2.15.1 (3) is made when conditions have caused one logic I subsystem (DA" or "B") to become inoperable but the redundant logic subsystem remains operable. The loss of a prime initiation relay (which renders all 4 channels of a logic subsystem inoperable) is the condition most likely to cause entry into Technical Specification 2.15.1 (3). In this situation, the remaining ESF logic subsystem still has the I

capability to automatically actuate engineered safety features equipment and essential supporting systems. The 48-hour completion time is commensurate with the importance of avoiding the vulnerability of a single failure in the remaining ESF logic subsystem.

Technical Specification 2.15.1 (3) will not be used upon loss of the common channels that affect both "A" and "B" subsystems prime initiators operability unless the permissible bypass condition is met. Upon exiting TS 2.15.1 (3) following the restoration of a prime initiation relay to OPERABLE status, if any channel(s) remain inoperable, the appropriate Limiting Conditions for Operation (LCO) (TS 2.15.1 (1) or TS 2.15.1 (2) is applicable with the length of inoperability measured from time of discovery of: 1) prime initiation relay inoperable, or 2) channel inoperability, whichever is longer.

The ESF system provides a 2-out-of-4 logic on the signals used to actuate the equipment connected to each of the two emergency diesel generator units.

The rod block system automatically inhibits all CEA motion in the event a LCO on CEA insertion, CEA deviation, CEA overlap or CEA sequencing is approached. The installation of the rod block system ensures that no single failure in the control element drive control system (other than a dropped CEA) can cause the CEAs to move such that the CEA insertion, deviation,'sequencing or overlap limits are exceeded. Accordingly, with the rod block system installed, only the dropped CEA event is considered an Anticipated Operational Occurrence (AOO) and factored into the derivation of the Limiting Safety System Settings (LSSS) and LCO. With the rod block function out-of-service, several additional CEA deviation events must be considered as AOOs. Analysis of these incidents indicates that the single CEA withdrawal incident is the most limiting of these events. An analysis of the at-power single CEA withdrawal incident was performed for Fort Calhoun for various initial Group 4 insertions, and it has been concluded that the LCO and LSSS are valid for a Group 4 insertion of less than or equal to 15%.

2.15 - Page 6 Amendment No. 125,194,208, 24Q TSBC 06-001-0 270 TSBC

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

Operability of the primary CEA position indication system (CEAPIS) channel and the secondary CEAPIS channel is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits of TS 2.10.2. The primary CEAPIS channel utilizes the output of a synchro transmitter geared to the clutch output shaft. CEA position is displayed visually at the main control panel.

The secondary CEAPIS channel utilizes the output of a voltage divider network controlled by a series of reed switches. The reed switches are actuated by a permanent magnet attached to the rack assembly. Position information is supplied to the distributed control system (DCS) flat-panel touch monitors for simultaneous viewing of all CEA group positions.

Limit switches on the regulating CEAs and reed switches on the shutdown CEAs provide an additional means of determining CEA position when the CEAs are fully inserted or fully withdrawn. When the CEAs are fully inserted or fully withdrawn, this indication (displayed on the DCS) can be used in lieu of secondary CEAPIS data to meet the shiftly CHANNEL CHECK of primary CEAPIS. However, as limit switch indication is not fully independent of secondary CEAPIS, primary CEAPIS must be used to verify secondary CEAPIS data.

\\,

In MODES 1 and 2, CEA position indication is required to allow verification that the CEAs are positioned and aligned as assumed in the safety analysis. If one CEA position indication channel is inoperable for one or more CEAs, TS 3.1, Table 3-3, Item 4 (CEA position verification) must be performed within 15 minutes following any CEA motion in that group to ensure that the CEAs are positioned as required.

The operability of the Alternate Shutdown Panel (AI-1S5), including Wide Range Logarithmic Power and Source Range Monitors on AI-212. and Emergency Auxiliary Feedwater Panel (AI-179) instrument and control circuits ensures that sufficient capability is available to permit entry into and maintenance of the Hot Shutdown Mode from locations outside of the Control Room. This capability is required'in the event that Control Room habitability is lost due to fire in the cable spreading room or Control Room.

Variances which may exist at startup between the more accurate!:.T-Power and Nuclear Instrumentation Power (NI-Power) are not significant for enabling of the trip functions. By 15% of rated power as measured by the uncalibrated Nt Power, the Axial Power Distribution (APD) and Loss of Load (LOL) trip functions are enabled while the High Rate of Change of Power trip is bypassed.

The APD trip function acts to limit the axial power shape to the range assumed in the setpoint analysis. Significant margins to local power density limits exist at 15% power, as well as power levels up to at least 30% (where NI calibration occurs).

2.15

• Page 7 Amendment No. 200, 249 270 TSBC-04-001-0 TSBC-OS-003-0 TSBC OS-OOS-O TSBC-11-006-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

The LOL trip function acts as an anticipatory trip for the high pressurizer pressure and high power trips in order to limit the severity of a LOL transient. This trip is not credited in the USAR Chapter 14 Safety Analyses and any variance between flT-Power and NI-Power has no effect on the safety analysis.

The High Rate of Change of Power trip acts to limit power excursions from low power levels and bypassing of this trip at a high power level is conservative. This trip is not credited in the USAR Chapter 14 Safety Analyses for Mode 1 operation. Any variance between flT-Power and NI-Power has no effect on the safety analysis.

Steam generator blowdown isolation ensures that the auxiliary feedwater system performs its design function of maintaining adequate steam generator (SG) water level for decay heat removal once the auxiliary feedwater actuation signal (AFAS) is actuated. The steam generator blowdown isolation function consists of two trains (logic subsystems). Each train closes one SG blowdown isolation valve to each SG. Each SG has redundant (Train A and Train B) blowdown isolation valves. Four clutch power relays initiate closure of the SG blowdown isolation valves with each clutch power relay closing one valve when the reactor trips. Failure of one clutch power relay to initiate SG blowdown isolation or failure of one train will not prevent single valve isolation of SG blowdown flow.

References (1)

USAR, Section 7.2.7.1 2.15 - Page 8 Amendment No. 2:08, 24Q 270 TSBC-08-008-0 TSBC-09-009-0

TECHNICAL SPECIFICATIONS TABLE 2*2 Instrument Operating Reguirements for Reactor Protective System

Test, Maintenance Minimum Minimum Permissible and Functional Operable Degree of Bypass Inoperable No.

Unit Channels Redundancy Condition Bypass Not Used 2(b)(C) 1(c) 2 High Power Level Thermal Power (e)

Input Bypassed below 10.4% of Rated Pqwer(a)(d) 2(b) 3 Thermal Margin/Low Below 10.4% of (e)

Pressurizer Pressure Rated Power(a)(d) 2(b) 4 High Pressurizer 1

None (e)

Pressure 5

Low R.C. Flow 2(b)

Below 10-4% of (e)

Rated Power (a)(d) 6 Low Steam Generator 2/Steam 1/Steam None (e)

Gen(b)

Water Level Gen 7

Low Steam Generator 2/Steam 1/Steam.

Below 600 (e)

Gen(b) pSia(a)(d)

Pressure Gen 2(b) 8 Containment High 1

During Leak Test (e)

Pressure 2(b)(c) 1(c) 9 Axial Power Below 15% of (9)

Distribution Rated Power(9) 10 High Rate Trip-wide 2(b) 1 Below 10.4% and (e)

Range Log Channels above 15% of Rated Power(a)(g) 2(b) 11 Loss of Load Below 15% of (9)

Rated Power(g) 2(b) 12 Steam Generator 1

None (e)

Differential Pressure

a.

Bypass automatically removed.

b.

Specification 2.15.1 (2) is applicable.

2.15 - Page 9 Amendment No. 60,77,88,16a,1Q4, 24Q 270

TECHNICAL SPECIFICATIONS TABLE 2-2 (Continued)

c.

If two channels are inoperable, load shall be reduced to 70% or less of rated power.

d.

For low power physics testing this trip may be bypassed up to 1 O'J% of rated power.

e.

Specification 2.15.1 (1) is applicable.

f.

Deleted.

g.

For each channel, the same bistable automatically activates the Loss of Load and Axial Power Distribution (APD) trips and automatically bypasses the high rate trip at 15% of rated power. Only the APD trip is a Limiting Safety System Setting. Therefore, the bistable is set to actuate within the APD tolerance band.

2.15 - Page 10 Amendment No. 60,77,88, 194, 24Q I 270

TECHNICAL SPECIFICATIONS TABLE 2-3 Instrument Operating Reguirements for Engineered Safety Features

Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No.

Unit Channels Redundancy Condition Bypass 1

Safet~ Injection A

Manual 1

None None N/A B

High Containment Pressure Logic Subsystem A 2(a)(d)(I) 1 During Leak (f)

Logic Subsystem B 2(a)(d)(I) 1 Test C

Pressurizer Low/Low Pressure 2(a)(d)(I)

Logic Subsystem A 1

Reactor Coolant (f) 2(a)(d)(I)

Logic Subsystem B 1

Pressure Less Than 1700 psia(b) 2 Containment Sl2ra~

A Manual(m) 1 None None N/A B

High Containment Pressure Logic Subsystem A 2(a)(C)(d)(I) 1 During Leak (f)

Logic Subsystem B 2(a)(C)(d)(I) 1 Test C

Pressurizer Low/Low Pressure 2(a)(c)(d)(I)

Logic Subsystem A 1

Reactor Coolant (f) 2(a)(c)(d)(I)

Logic Subsystem B 1

Pressure Less Than 1700 psia(b)

D Steam Generator Low Pressure Logic Subsystem A 2/Steam 1/Steam Steam Generator (f)

Gen(a)(C)(d)(I) Gen Pressure Less Than Logic Subsystem B 2/Steam 1/Steam 600 psia(n)

Gen(a)(C)(d}(l) Gen 3

Recirculation A

Manual 1

None None N/A B

SIRW Tank Low Level 2(a)(k)(I)

Logic Subsystem A 1

None U) 2(a)(k)(l)

Logic Subsystem B 1

4 Emergenc~ Off-Site Power Trip 1(e)

A Manual None None N/A B

Emergency Bus Low Voltage (Each Bus) 2(d)

-Loss of Voltage 1

Reactor Coolant (f) 2(a)(d)

-Degraded Voltage 1

Temperature Less Than 300 0 F 2.15 - Page 11 Amendment No. 4~,95,88,Ha,H14,~94,2:4Q,

~270

TECHNICAL SPECIFICATIONS TABLE 2-3 (Continued)

No.

Functional Unit Minimum Operable Channels Minimum Degree of Redundancy Permissible Bypass Condition

Test, Maintenance and Inoperable Bypass 5

Auxiliary Feedwater A

Manual None None N/A B

Auto. Initiation Logic Subsystem A Logic Subsystem B Operating Modes 3,4, and 5

-Steam Generator Low Level

• Steam Generator Low Pressure

• Steam Generator Differential Pressure 2(a)(d)(I) 3(&)(9)(1) 3(a)(9)(I) 1 1

(h)

(i)

(i) a Circuits on ESF Logic Subsystems A and B each have 4 channels.

b Auto removal of bypass above 1700 pSia.

c Coincident containment high pressure, pressurizer low/low pressure, and steam generator low pressure signals are required for initiation of containment spray.

d If minimum OPERABLE channel conditions are reached, one inoperable channel must be placed in the tripped condition or low level actuation position for auxiliary feedwater system within eight hours from the time of discovery of loss of operability. Specification 2.15.1 (2) is applicable.

e Control switch on incoming breaker.

If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from time of discovery of loss of operability. Specification 2.15.1 (1) is applicable.

g Three channels required because bypass or failure results in auxiliary feedwater actuation block in the affected channel.

h Specification 2.15.1 (1) is applicable.

2.15

• Page 12 Amendment No. ge,88,194,249, aea I 270

TECHNICAL SPECIFICATIONS k

m n

TABLE 2-3 (Continued)

If the channel becomes inoperable, that channel must be placed in the bypassed condition within eight hours from time of discovery of loss of operability. If the channel is not returned to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of discovery of loss of operability, one of the eight channels may continue to be placed in the bypassed condition provided the Plant Review Committee has reviewed and documented the judgment concerning prolonged operation in bypass of the inoperable channel. The channel shall be returned to OPERABLE status no later than during the next cold shutdown. If one of the four channels on one steam generator is in prolonged bypass and a channel on the other steam generator becomes inoperable, the second inoperable channel must be placed in bypass within eight hours from time of discovery of loss of operability. If one of the inoperable channels is not returned to OPERABLE status within seven days from the time of discovery of the second loss of operability, the unit must be placed in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If one channel becomes inoperable. that channel must be placed in the bypassed condition within eight hours from time of discovery of loss of operability. If.the channel is not returned to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from time of discovery of loss of operability. one of the eight channels may continue to be placed in the bypassed condition provided the Plant Review Committee has reviewed and documented the judgment concerning prolonged operation in bypass of the inoperable channel. The channel shall be returned to OPERABLE status no later than during the next cold shutdown. If a channel is in prolonged bypass and a channel on the opposite train becomes inoperable. the second inoperable channel must be placed in bypass within eight hours from time of discovery of loss of operability. If one of the inoperable channels is not returned to OPERABLE status within seven days from the time of discovery of the second loss of operability, the unit must be placed in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Specification 2.15.1 (2) is applicable.

Specification 2.15.1 (3) is applicable. If ESF Logic Subsystems A and B are inoperable.

enter Specification 2.0.1.

Steam Generator Low Pressure permissive is required for actuation.

(

Auto removal of bypass prior to exceeding 600 psia.

2,15 - Page 13 Amendment No. 88,173,194,249, 2ee I 270

TECHNICAL SPECIFICATIONS TABLE 2*4 Instrument Operating Conditions for Isolation Functions

Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable Unit Channels Redundancy Condition Bypass Containment Isolallon A

Manual 1,

None None N/A B

Containment High Pressure Logic Subsystem A 2(8)(9)(0)

During Leak (f)

Logic Subsystem B 2(a)(9)(g)

Test C

Pressurizer Low/Low Pressure 2(8)(9)(0)

Logic Subsystem A 1

Reactor Coolant (f) 2(8)(9)(9)

Logic Subsystem B 1

Pressure Less Than 1700 psla1bl 2

Steam Generator Isolation A

Manual 1

None None N/A B

Steam Generator Isolation 1

None None N/A (i) Steam Generator Low Pressure Logic SUbsystem A 2/Steam 1/Steam Steam Generator (1)

Gen(a)(e)(g)

Gen Pressure Less Than 600 psia(C)

Logic Subsystem B 21Steam 1/Steam Gen(a)(e)(g)

Gen (ii) Containment High Pressure Logic Subsystem A 2(8)(e)(0) 1 During Leak (f)

Logic Subsystem B 2(8)(8)(0) 1 Test 3

Ventilation Isolation A

Manual None None N/A B

Containment High Radiation Logic Subsystem A 1(d)(g)

None If Containment (f)

Logic Subsystem B 1(d)(g)

None Relief and Purge Valves are Closed 4

Steam ~enera12r BlowdQwn ISQlgtion 1(h)

A Manual None Operating Modes N/A 3.4,&5 2(h)(i)

B Reactor Trip None Operating Modes

0)

Trains A and B 3, 4, & 5 OR if at least one valve for each steam generator is closed 2.15* Page 14 Amendment No. 88.93.108.162.11:;3.173.184,194,249, I a&e;a83 270

TECHNICAL SPECIFICATIONS TABLE 2*4 (Continued) a Circuits on ESF Logic Subsystems A and B each have 4 channels.

b Auto removal of bypass prior to exceeding 1700 psia.

c Auto removal of bypass prior to exceeding 600 psia.

d A and B trains are both actuated by either the Containment or Auxiliary Building Exhaust Stack initiating channels. The number of installed channels for Containment Radiation High Signal is two for purposes of Specification 2.15'.1 (1 ).

e If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. Specification 2.15.1 (2) is applicable.

If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability. Specification 2.15.1 (1) is applicable.

g Specification 2.15.1 (3) is applicable. If ESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1.

h.

"Minimum Operable Channels" for steam generator blowdown isolation refers to the minimum number of trains (logic subsystems) which are required to be operable to provide manual or automatic SG blowdown isolation.

i.

If both trains become inoperable, power operation may continue provided at least one SG blowdown isolation valve for each steam generator is closed OR be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specifications 2.15.1 (1), (2), (3) and (4) are not applicable; TS LCO 2.0.1 is not applicable.

j.

If one train becomes inoperable, that train may be placed in the bypassed condition. If the train is not returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovery of loss of operability, operation may continue as long as one SG blowdown isolation valve to each steam generator is closed. If the train is not returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovery, with blowdown not isolated to both SGs, be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specifications 2.15.1 (1), (2), (3) and (4) are.!lQl applicable; TS LGO 2.0.1 is not applicable.

/

2.15* Page 15 Amendment No. 88,108,"\\52,173,194,249,a63\\

270

TECHNICAL SPECIFICATIONS TABLE 2-5 Instrumentation Operating Reguirements for Other Safety Feature Functions Minimum Minimum Permissible Functional Operable Degree of Bypass No.

Unit Channels Redundancy Condition 1(a) 1 CEA Position Indication None None Systems 2

Pressurizer Level 1

None Not Applicable NOTES:

(a)

If one channel of CEA position indication is inoperable for one or more CEAs, requirements of speCification 2.15.1 are modified for item 1 to "Perform TS 3.1, Table 3-3. Item 4 within I 15 minutes following any CEA motion in that group.1I SpeCifications 2.15.1 (1), (2), and (3)

. are not applicable.

2.15 - Page 16 Amendment No.-94,66,11 0,249, 266, 267 270 2e8

TECHNICAL SPECIFICATIONS TABLE 2-6 Alternate Shutdown and Auxiliary Feedwater Panel Functions Function/Instrument Required Number or Control Parameter Location of Channels 1.

Reactivity Control

a.

Source Range Power AI-212 1

b.

Reactor Wide Range AI-212 1

Logarithmic Power

2.

Reactor Coolant System Pressure Control

a.

Pressurizer Wide Range AI-179 1

Pressure (0-2500 psia)

3.

Decay Heat Removal via Steam Generators

a.

Reactor Coolant Hot Leg AI-185 1 (Note 1)

Temperature

b.

Reactor Coolant Cold Leg AI-185 1 (Note 1)

Temperature

c.

Steam Generator Pressure AI-179 1 per Steam Generator

d.

Steam Generator Narrow AI-179 1 per Steam Range Level Generator

e.

Steam Generator Wide AI-179 1 per Steam Range Level Generator

4.

Reactor Coolant System Inventory Controls

a.

Pressurizer Level AI-185 1

b.

Volume Control Tank Level AI-185 1

c.

Charging Pump CH-1 Band AI-185 1

its associated controls

d.

Charging Isolation Valve AI-185 1

Control

5.

Transfer Functions

a.

All Transfer Switches/Lockout AI-185 1

Relays

b.

All Transfer Switches/Lockout AI-179 1

Relays Note 1: One reactor coolant hot leg temperature indication and one reactor coolant cold leg temperature indication channel must both be operable on the same steam generator (i.e., RC-2A or RC-2B).

2.15 - Page 17 Amendment No.

270

TECHNICAL SPECIFICATIONS TABLE 2*6 (Continued)

Alternate Shutdown and Auxiliary Feedwater Panel Functions Function/Instrument Required Number or Control Parameter Location of Channels

6.

Auxiliary Feedwater Controls

a.

Steam Generator RC-2A and AI-179 1

2B Auxiliary Feedwater Isolation Inboard and Outboard Valves Control

b.

Steam-Driven Pump FW-10 AI-179 1

Recirculation Valve Control

c.

Steam-Driven Pump FW-10 AI-179 1

Steam Isolation Valve Control

d.

Steam from Steam Generator AI-179 1

RC-2A and RC-2B to FW-10 Steam Isolation Valve Control 2.15 - Page 18 Amendment No.

270

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO.1 DOCKET NO. 50-285

1.0 INTRODUCTION

By application dated February 10, 2012, as supplemented by letters dated October 1, 2012, and January 22,2013 (Agencywide Documents Access and Management System (ADAMS)

Accession Nos. ML12046A838, ML12276A043, and ML13023A046, respectively), Omaha Public Power District (OPPD, the licensee) requested changes to the Technical Specifications (TSs) (Appendix A to Renewed Facility Operating License No. DPR-40) for Fort Calhoun Station, Unit No.1 (FCS).

The amendment would revise the FCS TSs to establish the limiting condition for operation (LCO) requirements for the reactor protective system (RPS) actuation circuits in TS 2.15, "Instrumentation and Control Systems." Specifically, the amendment would renumber LCO 2.15(1) through 2.15(4) to 2.15.1 (1) through 2.15.1(4), renumber LCO 2.15(5) to LCO 2.15.3 with an associated Table 2-6, "Alternate Shutdown and Auxiliary Feedwater Panel Functions," and implement a new LCO 2.15.2 for the RPS logic and trip initiation channels.

The supplemental letters dated October 1, 2012, and January 22, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 7,2012 (77 FR 47128).

2.0 REGULATORY EVALUATION

2.1

System Description

The RPS for FCS consists of instrument channels, trip units, logic circuitry, and other equipment necessary to monitor selected nuclear steam supply system conditions and to effect reliable and rapid reactor shutdown if any condition or combination of conditions deviates from a preselected operating range. Specifically, there are four instrument channels that monitor each plant

- 2 variable and feed trip signals to six logic (ladder) matrices, for example, an input from each combination of channels AB, AC, AD, BC, BD and CD. Each of these ladder matrices produces two-of-two logic trip signals for all pair combinations of the four instrument channels. A trip signal from one of the ladder matrices produces trip signals in all four one-of-six trip paths.

These, in turn, trip all four power trip relays and RPS contactors.

OPPD drawing E-23866-411-003, Figure 7.2-2 in the FCS Updated Safety Analysis Report (USAR), shows the RPS circuitry. By letter dated January 22,2013, the licensee explained that it divided the circuitry in two sections. The first section is the signal processing and logic, and consists of instrument loops that provide four inputs to associated trip units for each of the 12 monitored safety parameters. These parameters are input to the trip unit signals in a 2-out-of-4 combination. The channel trips are then combined in six 2-out-of-4 logic ladder matrices. The second section of the RPS receives the output of the logic ladders and causes all four power trip relays and RPS contactors to trip when any of the logic ladders indicates that a 2-out-of-4 combination of the inputs reached a trip value. The RPS contactors provide the means of powering the control element drive mechanism (CEDM) clutches or removing the clutch power. The contactors consist of an "M" coil, four main contacts (Normally Open), and one auxiliary contact (Normally Closed). When the "M" coil is energized, its associated main contacts are closed to connect the circuit breaker output to the clutch power supplies.

Since the input and logic section processes field signals to determine if an RPS trip should be generated, the four channels are considered initiating channels. The actuation logic consists of M-contactors and associated contacts, an interposing relay and interconnecting wiring. Since the portion of the circuitry that connects the logic ladders to the M-contactors causes the RPS to actuate a reactor trip, it is also considered part of the actuation portion of the system.

In the reactor trip logic, there are two RPS manual trip channels. The manual trip channels are not processed through the logic ladders but act to de-energize the M-contactors directly. Thus, they are considered to be part of the actuation portion of the RPS. Manual reactor trip capability is afforded by two main control panel-mounted pushbuttons. One of these pushbuttons opens the contacts in series with each of the four trip paths, de-energizing all M-contactors. The other pushbutton opens the circuit breakers which provide alternating current (AC) input power to the M-contactor contacts and downstream clutch power supplies. Thus, depressing either pushbutton will cause a reactor trip.

Reactor trip is accomplished by de-energizing the CEDM magnetic clutch and releaSing the control element assemblies (CEAs) to drop into the core.

2.2 General Requirements Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of TS are contained in Title 10 of the Code of Federal Regulations (10 CFR),

Part 50, Section 50.36, "Technical specifications." The TS requirements in 10 CFR 50.36 include the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls.

- 3 The regulations in Criterion 3 of 10 CFR 50.36(c)(2)(ii) states that an LCO must be established for A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failures of or presents a challenge to the integrity of a fission product barrier.

The regulations in 10 CFR 50.36(c)(3), "Surveillance requirements," 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 for operation will be met.

In a memorandum dated September 18,1992 (ADAMS Accession No. ML003763736), the Commission approved the NRC staffs proposal in SECY-92-223, "Resolution of Deviations Identified During the Systematic Evaluation Program," not to apply 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," to plants with construction permits issued prior to May 21, 1971. FCS was licensed for construction prior to May 21, 1971, and at that time committed to the draft General Design Criteria (GDC). The draft GDC, which are similar to Appendix A, "General Design Criteria for Nuclear Power Plants," in 10 CFR Part 50, are contained in Appendix G, "Response to 70 Criteria," of the FCS USAR. The following FCS Design Criterion pertain to the proposed changes:

FCS Design Criterion 14, "Core Protection System," requires that the core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

FCS DeSign Criterion 19, "Protection Systems Reliability," requires that the protection systems are designed for high functional reliability and in-service testability commensurate with the safety functions to be performed.

FCS Design Criterion 25, "Demonstration of Function of Functional Operability of Protection System," requires that means shall be included for testing protection systems while the reactor is in operation to demonstrate that no failure or loss of redundancy has occurred.

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes

The licensee proposed the following changes:

Renumber LCO 2.15 title to 2.15.1, including renumbering the footnotes in Table 2-2 through 2-5 that reference the paragraphs of LCO 2.15;

- 4 Relocate LCO 2.15 paragraph (5) with its list of components into a new LCO 2.15.3 with the listed Alternate Shutdown Panels and the Auxiliary Feedwater Panel instrumentation or control circuits into a new Table 2-6; Add a new LCO 2.15.2 for the RPS logic and trip initiation channels; Relocate the RPS manual trip functional unit from TS Table 2-2 to new LCO 2.15.2; and Update the Table of Contents to reflect the proposed changes.

By letter dated February 10, 2012, the licensee stated that the proposed changes to TS 2.15 will result in the FCS LCO 2.15 being more aligned with NUREG-1432, Revision 3.0, "Standard Technical Specifications, Combustion Engineering Plants," June 2004 (STS) (ADAMS Accession No. ML041830597).

3.2 Background

On June, 14,2010, the RPS M2 contactorfailed to open during periodic surveillance testing.

When this situation presented, Operations could not enter an LCO for this circumstance because none was defined for such condition. Licensee Event Report (LER) 2011-005 was submitted by the licensee on May 9, 2011 (ADAMS Accession No. ML111300096), pursuant to 10 CFR 50.73(a)(2)(i)(B) to explain this event. This report states that Operations declared the RPS M2 contactor inoperable and entered TS LCO action 2.15( 1). Subsequent reviews determined that the station continued to operate in a condition not allowed by the TSs, and that TS 2.0.1 (similar to standard TS 3.0.3.) should have been entered, instead of entering TS 2.15(1).

By letter dated February 10, 2012, OPPD requested a license amendment request (LAR) to the Renewed Facility Operating License No. DPR-40 for FCS. This LAR would address the fact that RPS M2 contactor does not have a specific TS LCO action statement and station action to avoid entering TS 2.0.1 by establishing LCO requirements for the RPS actuation circuits in TS 2.15.

3.3

NRC Staff Evaluation

3.3.1 Renumber LCO 2.15 Title to 2.15.1! Including Renumbering the Footnotes in Table 2-2 through 2-5 that Reference the Paragraphs of LCO 2.15 TS LCO 2.15, "Instrumentation and Control Systems," title is renumbered to TS LCO 2.15.1, "Instrumentation and Control Systems." The footnotes that reference specifications paragraphs 1 through 4 under TS LCO 2.15 are revised to match the revised LCO numbers as follows:

Table 2-2, footnote b states, "Specification 2.15.1 (2) is applicable."

Table 2-2, footnote e states, "Specification 2.15.1 (1) is applicable."

- 5 Table 2-3, footnote d states, "If minimum OPERABLE channel conditions are reached, one inoperable channel must be placed in the tripped condition or low level actuation position for auxiliary feedwater system within eight hours from the time of discovery of loss of operability. Specification 2.15.1 (2) is applicable."

Table 2-3, footnote f states, "If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from time of discovery of loss of operability. Specification 2.15.1 (1) is applicable."

Table 2-3, footnote h states, "Specification 2.15.1 (1) is applicable."

Table 2-3, footnote k states, "Specification 2.15.1 (2) is applicable."

Table 2-3, footnote I states, "Specification 2.15.1 (3) is applicable. If ESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1."

Table 2-4, footnote d states, "A and B trains are both actuated by either the Containment or Auxiliary Building Exhaust Stack initiating channels. The number of installed channels for Containment Radiation High Signal is two for purposes of Specification 2.15.1 (1 )."

Table 2-4, footnote e states, "If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. Specification 2.15.1(2) is applicable."

Table 2-4, footnote f states, "If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability. Specification 2.15.1 (1) is applicable."

Table 2-4, footnote g states, "Specification 2.15.1 (3) is applicable. If ESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1."

Table 2-4, footnote i states, "If both trains become inoperable, power operation may continue provided at least one SG blowdown isolation valve for each steam generator is closed OR be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specifications 2.15.1(1), (2), (3) and (4) are not applicable; TS LCO 2.0.1 is not applicable."

Table 2-4, footnote j states, "If one train becomes inoperable, that train may be placed in the bypassed condition. If the train is not returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovery of loss of operability, operation may continue as long as one SG blowdown isolation valve to each steam generator is closed. If the train is not returned to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of discovery, with blowdown not isolated to both SGs, be in MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specifications 2.15.1 (1), (2), (3) and (4) are not applicable; TS LCO 2.0.1 is not applicable."

-6 Table 2-5, footnote (a) states, "If one channel of CEA position indication is inoperable for one or more CEAs, requirements of specification 2.15.1 are modified for item 1 to "Perform TS 3.1, Table 3-3. Item 4 within 15 minutes following any CEA motion in that group." Specifications 2.15.1 (1), (2), and (3) are not applicable."

These above changes are administrative in nature and are, therefore, acceptable.

3.3.2 Update the Table of Contents to Renect the Proposed Changes The licensee is proposing to revise the TS Table of Contents to reflect the two new proposed LCOs (2.15.2 and 2.15.3) and to show the renumbering of LCO 2.15 to LCO 2.15.1. These changes are administrative in nature and are, therefore, acceptable.

3.3.3 New Proposed TS LCO 2.15 The current FCS TSs do not include an LCO for when the RPS logic unit or trip initiation channels become inoperable. Therefore, if an event similar to that described in LER 2011-005 were to occur, this would require the operators to enter TS 2.0.1. However, the FCS TS includes surveillance requirements (SRs) for the RPS logic. These SRs are defined in TS 3.1, Table 3-1, Item 12, for a quarterly functional test of the RPS logic units, and TS 3.1, Table 3-1, Item 10 for a prior to critical functional test of manual trips. The SRs meet the requirement in FCS Design Criterion GDCs 19 and 25.

In this LAR, FCS is requesting to modify the current TS 2.15 to include a specific LCO action statement and station action to avoid entering TS 2.0.1 when the RPS logic unit or trip initiation channels become inoperable. Also, the addition of LCOs provides additional restrictions on the operation of the plant and provides required actions and time limits if these components are incapable of performing their function. The corrective measures include placing the unit in Hot Shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least subcritical and less than 300 degrees Fahrenheit (OF) within an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and placing the unit in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The RPS meets the FCS Design Criterion 14, and the proposed modification does not change the RPS; therefore, the RPS will continue to meeting this draft GDC. Further, the LAR does not request modifications to the SR for the RPS logic, so the RPS system will continue to meet FCS Design Criterion 19 and 25.

The proposed revision to LCO 2.15 would also result in this LCO being more aligned with NUREG-1432, Revision 3.0, for the RPS. The NRC staff notes that in certain situations the modifications proposed by FCS do not compare uniformly to NUREG-1432 because of the differences in the definition of operating modes. To facilitate comparison during the review of the proposed modifications, by letter dated October 1, 2012, the licensee submitted a diagram comparing operating modes in the STS and those defined in the FCS TS.

- 7 The licensee proposes to revise LCO 2.15 in the following manner:

Renumber LCO 2.15(1) through 2.15(4);

Include a new LCO 2.15.2 forthe RPS logic and trip initiation; and Move the requirements from LCO 2.15(5) to a new LCO 2.15.3, including the list of components into a new Table 2-6.

The following subsections describe the modification proposed for each of these items.

3.3.3.1 Renumbering of LCO 2.15(1) through 2.15(4)

This modification proposes to renumber LCO 2.15(1) through 2.15(4) to become 2.15.1 (1) through 2.15.1 (4), as well as the appropriate renumbering of footnotes in Tables 2-2 through 2-5 that reference these paragraphs.

In addition, FCS proposes to remove item No.1 of TS Table 2-2 and include it as part of the proposed new LCO 2.15.2. In response to the NRC staff's request for additional information (RAI) dated August 31,2012 (ADAMS Accession No. ML12236A243), by letter dated October 1, 2012, the licensee explained that the purpose of this table is to identify operability requirements for initiating channels, and not for actuation trains. As discussed in Section 2.1 of this safety evaluation, manual trip channels are part of the actuation portion of the RPS circuitry as opposed to the initiating channel logic portion of the system. As a consequence, the manual trip channel operability requirements do not belong in Table 2-2. To correct this inconsistency, the licensee relocated this requirement to the proposed TS 2.15.2. Thus, this table will then identify requirements for minimum operable channels or a minimum degree of redundancy that could be generically addressed by the paragraphs of current TS 2,15(1) through 2.15(4).

Based on the information presented, the NRC staff concludes that the modification proposed to TS 2.15(1) through 2.15(4) is acceptable and continues to meet 10 CFR 50.36.

3.3.3.2 New LCO 2.15.2 The licensee has proposed a new FCS TS LCO 2.15.2, "Reactor Protective System (RPS)

Logic and Trip Initiation." The proposed FCS TS LCO 2.15.2 states:

2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems 2.15.2 Reactor Protective System (RPS) Logic and Trip Initiation Applicability Applies to the operational status of RPS Logic and Trip Initiation channels in MODES 1 and 2; and, When reactor coolant temperature (T COld) is greater than 210°F or MODE 4 with more than one CEA rod capable of being withdrawn and

- 8 RCS boron concentration less than REFUELUNG BORON CONCENTRATION.

Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specification Six channels of RPS Logic matrices, four channels of RPS Trip Initiation Logic and two channels of RPS Manual Trip shall be OPERABLE.

Reguired Actions (1)

With one RPS Trip Matrix channel inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(2)

With one RPS Trip Initiation Logic channel inoperable, de-energize the affected clutch power supply within one hour.

(3)

With one RPS Manual Trip channel inoperable, restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3.

(4)

With two RPS Trip Initiation Logic channels affecting the same trip leg inoperable, de-energize the affected clutch power supplies immediately.

(5)

With the required actions of (1), (2), or (4) not met, or with two RPS Manual Trip channels inoperable, or with two or more RPS Logic Matrices inoperable, or with two or more RPS Trip Initiation Logic channels inoperable for reasons other than (4):

a.

be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, OR

b.

verify reactor coolant boron concentration is at REFUELING BORON CONCENTRATION within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The proposed new LCO 2.15.2 would address those instances in which RPS logic unit or trip initiation channel is inoperable. The RPS logic consists of six channels of RPS logic matrices, four channels of RPS trip initiation channels, and two channels of manual trip initiation.

The LAR defines the applicability for this TS LCO to be MODE 1 and 2 and conditions when reactor coolant temperature (Tco1d) is greater than 210°F or MODE 4 with more than one CEA rod

- 9 capable of being withdrawn and RCS boron concentration less than REFUELING BORON CONCENTRATION. By letter dated October 1, 2012, the licensee explained in the response to RAI# 2 and #3 that this last applicability was defined in this manner because the FCS TS does not define a specific mode for RCS temperatures between 210 OF and 515 of, whereas NUREG 1432 defines operating modes that cover this temperature range. In this manner, the applicability defined for the proposed TS 2.15.2 corresponds to the same range of plant operating conditions as defined in NUREG 1432. The NRC staff notes that FCS TS uses Tcold instead of Tavg, which is used in NUREG-1432. The reason for this is that FCS uses Tcold for the definition of MODE 4. Also, Tavg is approximately equal to Tcold when the plant is in MODE 4, Cold Shutdown.

The LAR defines the required actions when the LCO is not met. The LAR provides justification for each of the required actions. The required actions and their justification are similar to the required actions identified in NUREG-1432 TS 3.3.3, RPS logic and trip initiation (analog). By implementing this new LCO, TS 2.15 will include an LCO for failures associated with RPS logic.

For example, if an event similar to that described in LER 2011-005 where to occur, the licensee would enter the proposed TS 2.15.2, required action (2).

One of the required actions included in the proposed TS 2.15.2 is related to inoperability of the RPS manual trip channels. As explained above, this requirement was relocated from the current TS Table 2-2, and included as part of the proposed new LCO 2.15.2, required action (3).

Specifically, item No.1 in the current TS Table 2-2 requires that one manual trip button to be OPERABLE, and no redundancy provided. By letter dated October 1,2012, the licensee explained that the reason for adding this requirement in the proposed TS 2.15.2 is that there is an inconsistency between the number of redundant channels and minimum operable channels defined in Table 2-2 for the manual trip buttons. The current TS Table 2-2 defines that there are two manual trip channels, and only one is required to be operable. According to the current TS, if a manual trip channel is inoperable (the minimum number defined in Table 2-2), the licensee would enter current TS 2.15(2), which requires the inoperable channel to be placed in the trip position. In this case, the trip position would trip the reactor, since manual reactor trip is a 1-out-of-2 system. Tripping the reactor with one inoperable manual trip button should not be necessary since the USAR safety analysis does not assume operation of the manual trip.

Therefore, this condition was modified to require restoration of the inoperable channel to operable prior to entering MODE 2 from MODE 3. In this manner if one of the manual reactor trip channels is inoperable, operation may continue until the reactor is shutdown for other reasons.

Based on the above, the NRC staff concludes that the licensee has defined a new LCO, the plant conditions during which the LCO applies, and the required actions when the LCO is not met. This new LCO meets the requirements of 10 CFR 50.36.

The licensee also stated in its application, "These proposed changes to TS 2.15 will result in the FCS LCO 2.15 being more aligned with NUREG 1432, Standard Technical Specifications, Combustion Engineering Plants for the RPS." The NRC staff agrees with the licensee that FCS TS LCO 2.15.2 is better aligned with NUREG-1432, Revision 3.0.

- 10 3.3.3.3 New LCO 2.15.3 Current FCS TS LCO 2.15 specification paragraph 5 states:

... (5)

In the event that the number of operable channels of the listed Alternate Shutdown Panels or the Auxiliary Feedwater Panel instrumentation or control circuits falls below the required number of channels, either restore the required number of channels to OPERABLE status within seven (7) days, or be in hot shutdown (Mode 3) within the next twelve hours. This specification is applicable in Modes 1 and 2.

The licensee is proposing to move LCO 2.15 specifications paragraph 5 into new proposed FCS TS LCO 2.15.3. FCS TS LCO 2.15 specifications paragraph 5 is applicable to the Alternate Shutdown Panels and the Auxiliary Feedwater Panel instrumentation and control circuits listed within LCO 2.15. The new proposed FCS TS LCO 2.15.3 is applicable to the same Alternate Shutdown Panels and the Auxiliary Feedwater Panel instrumentation and control circuits listed within LCO 2.15; however, the list of equipment is being relocated to new proposed TS Table 2-6. FCS TS LCO 2.15 specifications paragraph 5 is applicable in MODES 1 and 2 and the new proposed FCS TS LCO 2.15.3 is also applicable in MODES 1 and 2. FCS TS LCO 2.15 specifications paragraph 5 requires either restoring the required number of channels to operable status within 7 days, or be in Hot Shutdown (MODE 3) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the number of operable channels of the listed Alternate Shutdown Panels or the Auxiliary Feedwater Panel instrumentation or control circuits falls below the required number of channels.

The new proposed FCS TS LCO 2.15.3 requires restoring the required number of channels to OPERABLE within 7 days, when the number of OPERABLE channels or control circuits is less than the required number of channels. If the required number of channels is not returned to operable status within 7 days, then the plant must be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In the LAR, the licensee stated that no modifications were proposed for the new TS 2.15.3, and only format modifications were made. However, the text proposed in the LAR for TS 2.15.3 is slightly different than the text in the current TS 2.15(5). By letter dated October 1, 2012, the licensee provided additional information and explanation about this variance. In particular, the licensee explained that the modifications made in the text were necessary to clarify the order of the required actions to be performed when the LCO is not met. Further, by letter dated January 22, 2013, the licensee explained that the current TS 2.15(1) through (4) define actions to be taken when the requirements for circuit components listed in Tables 2-2 through 2-5 are not operable. However, the components for the Alternate Shutdown Panel and Auxiliary Feedwater Panel instrumentation or control circuits are not listed in Tables 2-2 through 2-5, instead they are listed as part of TS 2.15(5). Thus, the actions defined in TS 2.15(1) through 2.15(4) are not applicable to the Alternate Shutdown Panel and Auxiliary Feedwater Panel instrumentation or control circuits.

The proposed change moves current TS 2.15(5) requirement and identification of components for the Alternate Shutdown Panel and Auxiliary Feedwater Panel instrumentation or control circuits into the new proposed TS 2.15.3 and its associated Table 2.6. With this modification the licensee has not modified the intent of the current TS 2.15(5).

- 11 Based on the above, the NRC staff concludes that the modification proposed to TS 2.15(5) is acceptable and continues to meet 10 CFR 50.36.

3.4 Conclusion The proposed changes in Sections 3.3.1 and 3.3.2 above are administrative in nature or involve the reorganization or reformatting of requirements without affecting technical content or operational requirements. The NRC staff concludes that the proposed changes are acceptable.

The NRC staff concludes that the licensee established a new LCO for the RPS actuation circuits in TS 2.15, which meets the requirements of 10 CFR 50.36. This new LCO adequately addresses the plant conditions during which the LCO applies and the required actions when the LCO is not met. Furthermore, the NRC staff concludes that the systems will continue to meet the requirements of the FCS Design Criterion14, 19, and 25.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on August 7,2012 (77 FR 47128). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: K. Bucholtz R. Alvarado Date:

February 28, 2013

ML13043A661

• via memo dated

'" OFFICE

• NRRlDORLlLPL4/PM NRRlDORLlLPL4/LA NRRlDSS/STSB/BC NRRlDE/EI CB/BC NAME LWilkins (JSebrosky for)

JBur1<:hardt RElliott*

JThorp' "DATE 2/26/13 2/13/13 10/16/12 1/31/13 OFFICE OGC NLO NRR/DORLlLPL4/BC NRRlDORLlLPL4/PM NAME LSubin MMar1<:ley (BSingal for)

LWilkins (JSebrosky for) iiDATE 2/26/13 2/28/13 2/28/13