ML15005A126

Issuance of Amendment Multiple Administrative Issues with the Technical Specification

ML15005A126
Person / Time
Site:	Waterford
Issue date:	02/23/2015
From:	Michael Orenak Plant Licensing Branch IV
To:	Entergy Operations
Wang A
References
TAC MF3230
Download: ML15005A126 (26)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 23, 2015 Vice President, Operations Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: MULTIPLE ADMINISTRATIVE ISSUES WITH THE TECHNICAL SPECIFICATIONS (TAC NO. MF3230)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 242 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated December 9, 2013, as supplemented by letters dated October 1, 2014, and December 17, 2014.

The amendment revises certain TSs to improve clarity, correct administrative and typographical errors, or establish consistency with NUREG-1432, "Standard Technical Specifications -

Combustion Engineering Plants," Revision 4.0, dated April 2012. This amendment also revises TS Table 3.3-1, Action 6.b.1 to include one technical change.

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

Docket No. 50-382

Enclosures:

1. Amendment No. 242 to NPF-38

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

~~~

Michael D. Orenak, Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS. INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 242 License No. NPF-38

1.

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

A.

The application for amendme'lt by Entergy Operations, Inc. (EOI), dated December 9, 2013, as supplemented by letters dated October 1. 2014, and December 17, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.2 of Facility Operating License No. NPF-38 is hereby amended to read as follows:

2.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 242, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

Attachment:

Changes to the Facility Operating License No. NPF-38 and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION Meena K Khanna, Chief Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 23, 2015

ATTACHMENT TO LICENSE AMENDMENT NO. 242 TO FACILITY OPERA TING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Facility Operating License and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3/4 1-24 3/4 1-24 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-15 3/4 3-15 3/4 3-16 3/4 3-16 3/4 3-31 3/4 3-31 3/4 3-45a 3/4 3-45a 3/4 3-60 3/4 3-60 6-1 6-1 6-2 6-2 6-23 6-23 or indirectly any control over (i) the facility, (ii) power or energy produced by the facility, or (iii) the licensees of the facility.

Further, any rights acquired under this authorization may be exercised only in compliance with and subject to the requirements and restrictions of this operating license, the Atomic Energy Act of 1954, as amended, and the NRC's regulations. For purposes of this condition, the limitations of 10 CFR 50.81, as now in effect and as they may be subsequently amended, are fully applicable to the equity investors and any successors in interest to the equity investors, as long as the license for the facility remains in effect.

(b)

Entergy Louisiana, LLC (or its designee) to notify the NRC in writing prior to any change in (i) the terms or conditions of any lease agreements executed as part of the above authorized financial transactions, (ii) any facility operating agreement involving a licensee that is in effect now or will be in effect in the future, or (iii) the existing property insurance coverages for the facility, that would materially alter the representations and conditions, set forth in the staff's Safety Evaluation enclosed to the NRC letter dated September 18, 1989. In addition, Entergy Louisiana, LLC or its designee is required to notify the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1.

Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2.

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 242, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

AMENDMENT NO. 242

REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to greater than or equal to 145 inches.

APPLICABILITY: MODES 1 ** and 2*#**.

ACTION:

With a maximum of one shutdown CEA withdrawn to less than 145 inches withdrawn, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:

a.

Withdraw the CEA to greater than or equal to 145 inches, or

b.

Declare the CEA inoperable and determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to greater than or equal to 145*inches withdrawn:

a.

Within 15 minutes prior to withdrawal of any CEAs in regulating groups or group P during an approach to reactor criticality, and

b.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

• See Special Test Exception 3.10.2.

1. With Keff greater than or equal to 1.0.

• • Except for surveillance testing pursuant to Specification 4.1.3.1.2.

WATERFORD - UNIT 3 3/4 1-24 AMENDMENT NO. 48%, 242

TABLE 3.3-1 (Continued)

ACTION STATEMENTS

2. Pressurizer Pressure -

High

3. Containment Pressure -

(RPS) High

4. Steam Generator Pressure - Low

5. Steam Generator Level

6. Core Protection Calculator

7. Logarithmic Power Pressurizer Pressure - High Local Power Density - High DNBR-Low Containment Pressure - High Containment Pressure - High (ESF)

Steam Generator Pressure - Low Steam Generator ~P 1 and 2 (EFAS 1 and 2)

Steam Generator Level - Low Steam Generator ~P (EFAS)

Local Power Density - High DNBR-Low Logarithmic Power Level - High Local Power Density - High (1>

DNBR - Low <1>

Reactor Coolant Flow - Low (1>

STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied.

ACTION 4 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

• ACTION 5 - With the number of channels OPERABLE one less those required by the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breakers of the inoperable channel are placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may be bypassed for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for surveillance testing per Specification 4.3.1.1.

ACTION 6 -

a.

With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, each CEA is verified to be within 7 inches (indicated position) of all other CEAs in its group. After 7 days, operation may continue provided that Actions 6.b.1, 6.b.2, and 6.b.3 are met.

• Limited plant cooldown or boron dilution is allowed provided the change is accounted for in the calculated SHUTDOWN MARGIN.

(1)

With the operating bypass enabled.

WATERFORD - UNIT 3 3/4 3-6 AMENDMENT NO. 5, 185, 225, 228 242

ACTION 7 -

ACTION 8 -

TABLE 3.3-1 (Continued)

ACTION STATEMENTS

b.

With both CEACs inoperable, operation may continue provided that:

1.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the DNBR margin required by Specification 3.2.4b (COLSS in service) or 3.2.4d (COLSS out of service) is satisfied and the Reactor Power Cutback System is disabled, and

2.

Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a)

All CEA groups are withdrawn to and subsequently maintained at the "Full Out" position, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2 or for control when CEA group 6 may be inserted no further than 127.5 inches withdrawn.

b)

The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to the inoperable status.

c)

The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the "Off' mode except during CEA group 6 motion permitted by a) above, when the CEDMCS may be operated in either the "Manual Group" or "Manual Individual" mode.

3.

At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, all CEAs are verified fully withdrawn except during surveillance testing pursuant to Specification 4.1.3.1.2 or during insertion of CEA group 6 as permitted by 2.a) above, then verify at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> that the inserted CEAs are aligned within 7 inches (indicated position) of all other CEAs in its group.

With three or more auto restarts of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement 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 /> or open the reactor trip breakers within the next hour.

WATERFORD-UNIT 3 3/4 3-7 AMENDMENT NO. 5, 162, 185, 242

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

~

MINIMUM

-I m

TOTAL NO.

CHANNELS CHANNELS APPLICABLE

o,,

FUNCTIONAL UNIT OF CHANNELS To TRIP OPERABLE MODES ACTION 0

o

4.

MAIN STEAM LINE ISOLATION 0

I

a. Manual (Trip 2 sets of c

1 set of 2 sets of 1, 2, 3 16 z

Buttons) 2 per steam 2 per steam 2 per operat-

=i generator generator ing steam

(.,)

generator

b. Steam Generator 4/steam 2/steam 3/steam 1, 2, 3 13*, 14*

Pressure - Low generator generator generator

c. Containment Pressure -

4 2

3 1, 2, 3 13*, 14*

High

d. Automatic Actuation 4

2 3

1, 2, 3 12 Logic

5.

SAFETY INJECTION SYSTEM SUMP RECIRCULATION (RAS)

(.,)

a. Manual RAS (Trip

~

c..>

Buttons) 2 1

2 1,2, 3,4 12 CJ1

b. Refueling Water Storage Pool-Low 4

2 3

1, 2, 3, 4 19,20

c. Automatic Actuation Logic 4

2 3

1, 2, 3, 4 12

6.

LOSS OF POWER (LOV)

a. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) 3/bus 3/bus 3/bus 1, 2, 3 17, 18

)>

b. 480 V Emergency Bus s::

m Undervoltage (Loss z

Cl of Voltage) 3/bus 3/bus 3/bus 1, 2, 3 17, 18 s::

m

c. 4.16 kV Emergency

~

Bus Undervoltage z

(Degraded Voltage) 3/bus 3/bus 3/bus 1, 2, 3 17, 18 9

t

~

I\\.)

~

I\\.)

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT QF CHANNELS TO TRIP OPERABLE MODES ACTION

7. EMERGENCY FEEDWATER (EFAS)

a. Manual (Trip Buttons) 2 sets of 2 1 set of 2 2 sets of 2 1,2,3 15 per steam per steam per steam generator generator generator

b. SG Level (1/2) -

Low and LlP (1/2) - High 4/steam 2/steam 3/steam 1,2,3 19,20 generator generator generator

c. SG Level (1/2) - Low and No S/G Pressure -

Low Trip (1/2) 4/steam 2/steam 3/steam 1,2,3 19,20 generator generator generator

d. Automatic Actuation Logic 4

2 3

1,2,3 12

e. Control Valve Logic 2/steam 1/steam 2/steam 1,2,3 15 (Wide Range SG generator generator generator Level - Low)

WATERFORD - UNIT 3 3/4 3-16 AMENDMENT NO. 4M 242

ACTION 23 -

ACTION 24 -

ACTION 25 -

ACTION 26 -

ACTION 27 -

ACTION 28 -

TABLE 3.3-6 (Continued)

ACTION STATEMENTS DELETED DELETED With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.9.4.

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable Channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

1.

Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and

2.

If the monitor is not restored to OPERABLE status within 7 days after the failure, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days after the failure outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, operation of the plant may continue provided grab samples are taken once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If the monitor is not restored to OPERABLE status within 30 days after the failure, continue sampling and prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 14 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

WATERFORD-UNIT 3 3/4 3-31 AMENDMENT NO. 91, 176,-49-1; 234-, 242

ACTION 29 -

ACTION 30 -

ACTION 31 -

TABLE 3.3-10 ACTION STATEMENTS With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 30 days, or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status.

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-1 O; either restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least 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 />.

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE in Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels to OPERABLE status.

WATERFORD-UNIT 3 3/4 3-45a AMENDMENT NO. 14, 122, 242

INSTRUMENTATION EXPLOSIVE GAS MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.11 The explosive gas monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 3.11.2.5 are not exceeded.

APPLICABILITY: As shown in Table 3.3-13.

ACTION:

a.

With an explosive gas monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, declare the channel inoperable, and take the ACTION shown in Table 3.3-13.

b.

With less than the minimum number of explosive gas monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the following 14 days to explain why this inoperability was not corrected in a timely manner.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.11 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

WATERFORD-UNIT 3 3/4 3-60 AMENDMENT NO. +1;-68, 242

ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager Plant Operations shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

The General Manager Plant Operations or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that may affect nuclear safety.

6.1.2 The Shift Manager, or during his absence from the control room, a designated individual shall be responsible for the control room command function. A management directive to this effect, signed by the Site Vice President - WF3, shall be reissued to all station personnel on an annual basis.

6.2 ORGANIZATION 6.2.1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a.

Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the FSAR.

b.

The General Manager Plant Operations shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c.

The Site Vice President - WF3 shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6.2.2 UNIT STAFF

a.

Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1; WATERFORD-UNIT 3 6-1 AMENDMENT NO. 18, 41, 63146, ~

242

ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

b.

At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the reactor is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room.

c.

A Health Physics Technician* shall be on site when fuel is in the reactor.

d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

e.

The Operations Manager or the Operations Manager - Shift shall hold a senior reactor operator license.

• This requirement tolerates Health Physics Technician unexpected absences for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided management takes immediate action to fill the required Health Physics Technician position.

• WATERFORD - UNIT 3 6-2 AMENDMENT NO. 18, 41, 50, 61, 146, 221, 242

ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued)

b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c.

A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Radiation Protection Manager in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrems* but less than 500 rads** shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Superintendent on duty and/or health physics supervision/designee. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> a dose in excess of 1000 mrems* but less than 500 rads** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee-initiated changes to the PCP:

a Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program Manual. This documentation shall contain:

• Measurement made at 30 centimeters from the radiation source or from any surface that the radiation penetrates.

• • Measurement made at 1 meter from the radiation source or from any surface that the radiation penetrates.

WATERFORD-UNIT 3 6-23 AMENDMENT NO. 68, 1 rn, 146,-t&t, 242

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 242 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC.

WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By application dated December 9, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13345A686), as supplemented by letters dated October 1, 2014 (ADAMS Accession No. ML14274A245), and December 17, 2014 (ADAMS Accession No. ML14351A360), Entergy Operations, Inc. (the licensee), requested changes to the Technical Specifications (TSs) for Waterford Steam Electric Station, Unit 3 (WF3). The supplemental letter dated October 1, 2014, provided additional information that clarified the application. The supplemental letter dated December 17, 2014, withdrew a component of the original request. The supplemental letters did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 5, 2014 (79 FR 45475).

The proposed changes would revise the TSs to improve clarity, correct administrative and typographical errors, or establish consistency with NUREG-1432, "Standard Technical Specifications - Combustion Engineering Plants," Revision 4.0, dated April 2012 (ADAMS Accession Nos. ML12102A165). This amendment also revises TS Table 3.3-1, Action 6.b.1, to include one technical change.

2.0 REGULATORY EVALUATION

The regulatory requirements and guidance, which the NRC staff considered in its review of the application, are as follows:

Title 10 of the Code of Federal Regulations (10 CFR) Part 50 establishes the fundamental regulatory requirements with respect to the domestic licensing of nuclear production and utilization facilities. Specifically, Appendix A to 10 CFR Part 50, "General Design Criteria [GDC]

for Nuclear Power Plants," provides, in part, the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. The following GDCs are applicable to this amendment:

GDC 10, "Reactor design," states that "[t]he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences."

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

GDC 20, "Protective system functions," states that "[t]he protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety."

GDC 21, "Protection system reliability and testability," requires that the system be designed for high functional reliability and in-service testability, with redundancy and independence sufficient to preclude loss of the protection function from a single failure and preservation of minimum redundancy despite removal from service of any component or channel.

GDC 22, "Protection system independence," requires that the system be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions do not result in loss of the protection function.

GDC 23, "Protection system failure MODES," requires that the system be designed to fail to a safe state in the event of conditions such as disconnection, loss of energy, or postulated adverse environments.

GDC 24, "Separation of protection and control systems," requires that interconnection of the protection and control systems be limited to assure safety in case of failure or removal from service of common components.

The regulations in 10 CFR 50.55a(h), "Protection and safety systems," require that the protection systems meet the requirements in the Institute of Electrical and Electronics Engineers (IEEE) Standard 279. Section 4.2 of IEEE Standard 279-1971 discusses the general functional requirement for protection systems to assure they satisfy the single failure criterion.

NUREG-1432, Revision 4 contains the Standard Technical Specifications (STSs) for Combustion Engineering Plants. The abstract of NUREG-1432 states, in part, that:

This NUREG contains the improved Standard Technical Specifications (STS) for Combustion Engineering (CE) plants. The changes reflected in Revision 4 result from the experience gained from plant operation using the improved STS and extensive public technical meetings and discussions among the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees and the Nuclear Steam Supply System (NSSS)

Owners Groups.

The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132), which was subsequently codified by changes to Section 36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) (60 FR 36953). Licensees are encouraged to upgrade their technical specifications consistent with those criteria and conforming, to the practical extent, to Revision 4 to the improved STS. The Commission continues to place the highest priority on requests for complete conversions to the improved STS. Licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency.

Regulatory Guide (RG) 1.8, "Qualification and Training of Personnel for Nuclear Power Plants,"

Revision 3, dated May 2000 (ADAMS Accession No. ML003706932), provides guidance that is acceptable to the NRC staff regarding qualifications and training for nuclear power plant personnel.

3.0 TECHNICAL EVALUATION

In it's application dated December 9, 2013, the licensee stated the purpose for making the TS changes for this amendment as follows:

Improve clarity, Correct administrative and typographical errors, and Establish consistency with NUREG-1432, Revision 4.0.

The NRC staff reviewed each of the proposed TS changes, individually, to confirm achievement of the stated objectives and to ensure that the changes do not impact compliance with the regulations listed in Section 2.0 of this safety evaluation (SE). The following evaluations are for each TS change, in the order presented in the license amendment request.

3.1 TS 3/4.3.1, "Reactor Protective Instrumentation" TS 3.3.1 specifies requirements for the operation of the Control Element Assembly Calculators (CEACs). The CEACs are designed to detect misaligned control element assemblies and adjust reactor trip set points by applying a penalty factor to the core protection calculators if the Control Element Assembly (CEA) could adversely affect core power distribution. While only one CEAC is required to perform the safety function, two are required in the TSs to be operable to allow for a single failure of a CEAC.

Current TS 3.3.1, Table 3.3-1, "Reactor Protective Instrumentation," Action 6.a, states:

a.

With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, each CEA is verified to be within 7 inches (indicated position) of all other CEAs in its group.

Current TS 3.3.1, Table 3.3-1, Action 6.b.1, states:

b.

With both CEACs inoperable, operation may continue provided that:

1.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the DNBR [departure from nucleate boiling ratio]

margin required by Specification 3.2.4b (COLSS [Core Operating Limit Supervisory System] in service) or 3.2.4d (COLSS out of service) is satisfied and the Reactor Power Cutback System is disabled.

Action 6.b.1 includes verification that the DNBR limit is met, disabling the Control Element Drive Mechanism Control System, and disabling the Reactor Power Cutback System.

The licensee's proposed changes to TS 3.3.1, Table 3.3-1, Actions 6.a. and 6.b.1, are to provide clarity and bring them into alignment with NUREG-1432, Limiting Condition for Operation (LCO) 3.3.3A, "Control Element Assembly Calculators (CEACs) (Digital) (Without Setpoint Control Program)."

The first proposed change in TS 3.3.1 would clarify Table 3.3-1, Action 6.a by adding "after 7 days, operation may continue provided that Actions 6.b.1, 6.b.2, and 6.b.3 are met." This addition is consistent with NUREG-1432. It clarifies the action required if the 7 days has elapsed because after 7 days, the licensee will take actions consistent with both CEAC being inoperable. The NRC staff has determined that the added clarification statement reduces the potential for ambiguity due to possible misinterpretation of the TS requirements.

The second proposed change in TS 3.3.1 would increase the Completion Time period in Table 3.3-1, Action 6.b.1 from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This increase is consistent with NUREG-1432, because the 4-hour Completion Time was found to be an adequate amount of time to accomplish the actions above in a controlled manner while minimizing risk. The NRC staff determined that the change being made to the completion time for TS 3.3.1, Table 3.3-1, Action 6.b.1, is consistent with NUREG-1432.

The NRC staff concludes that the proposed changes to TS 3.3.1, Table 3.3-1, Actions 6.a and 6.b.1 are acceptable because they are consistent with NUREG-1432 and 10 CFR 50.55a(h),

and continue to comply with GDCs 10, 13, 20, 21, and 22.

3.2 "Shutdown CEA Insertion Limit" Current LCO TS 3.1.3.5 requires that "All shutdown CEAs shall be withdrawn to greater than or equal to 145 inches," in MODES 1 and 2*#**. The symbols mean the following according to the TS:

• See Special Test Exception 3.10.2.

1. With Keff greater than or equal to 1.0.

• • Except for surveillance testing pursuant to Specification 4.1.3.1.2.

LCO TS 3.1.3.1 requires that all CEAs be operable. To verify operability, LCO TS 3.1.3.1 requires performance of Surveillance Requirement (SR) 4.1.3.1.2, "[e]ach CEA not fully inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 92 days." The performance of this SR may result in TS 3.1.3.5 not being met for very short periods of time as each CEA is moved and restored to its original position. The exemption from entering LCO TS 3.1.3.5 actions is currently present for MODE 2 but not for MODE 1.

The revision of LCO TS 3.1.3.5, Applicability to MODES 1 ** and 2*#**, would revise the TS to clarify that entry into Actions for LCO TS 3.1.3.5 is not required in MODE 1 when performing SR 4.1.3.1.2. This revision will extend the existing exemption for the performance of SR 4.1.3.1.2 to include MODE 1.

Momentary insertion of individual CEAs during this SR does not have a significant effect on core power distribution, shutdown margin, ejected CEA worth, or initial reactivity insertion rate during a reactor trip. This change is also consistent with NUREG-1432, LCO 3.1.5, "Shutdown Control Element Assembly (CEA) Insertion Limits (Digital)," which includes a note that this LCO is not applicable while performing NUREG-1432, SR 3.1.4.3.

The NRC staff concludes that the change is acceptable based on the fact that the SR 4.1.3.1.2 testing in MODE 1 has no noticeable effect on the core, the change is consistent with NUREG-1432, Revision 4.0, and complies with GDC 21.

3.3 TS 3/4.3.2, "Engineered Safety Features Actuation System Instrumentation" TS 3.3.2, Table 3.3-3, Functional Units:

5.b. Refueling Water Storage Pool - Low; 7.b. SG [Steam Generator] Level (1/2) - Low and 6. [Delta]P (1/2) - High; and, 7.c. SG Level (1/2) - Low and No SIG Pressure - Low Trip (1/2) specifies entry into Actions 19a*, 19b, 20 (where* indicates that Specification 3.0.4 is not applicable) when a channel for Safety Injection System Sump Recirculation Actuation Signals is inoperable.

Current Action 19 of Table 3.3-3 states that:

With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION and/or operation in the other applicable MODE(S) may continue, provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Current Action 19a specifies, in part,

a.

If the inoperable channel is to remain in the bypassed condition, the desirability of maintaining this channel in the bypassed condition shall be documented....

The current TS 3.3.2, Table 3.3-3, does not explicitly state that Action 19 needs to be entered before Action 19a is initiated.

The licensee is seeking to revise TS 3.3.2 to require entry into Actions 19 and 20 when one or more channels for Function 5.b, 7.b, or 7.c are inoperable to provide explicit recognition that all of Action 19 is required. This change is being made to provide clarity and to correct an error that omitted a required TS action.

The NRC staff has determined that this change correctly clarifies the applicability for Action 19 of TS 3.3.2. The NRC staff concludes that the change is acceptable because it corrects an error in the TS, is consistent with NUREG-1432 and 10 CFR 50.55a(h), and continues to comply with GDCs 13, 20, 21, 23, and 24.

3.4 TS 3/4.3.3, "Monitoring Instrumentation" TS 3.3.3.1, Table 3.3-6, "Radiation Monitoring Instrumentation," Action 27.2, applies if one or more Effluent Accident Monitors are inoperable, and includes a requirement that states, in part:

2.

If the monitor is not restored to OPERABLE status within 7 days after the failure, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event...

The licensee seeks to revise the wording of TS 3.3.3.1, Table 3.3-6, Action 27.2, and change, "following the event" to "after the failure," to provide clarity that both the 7-day and 14-day time constraints apply to the start of the failure of the monitor.

TS 3.3.3.1, Table 3.3-6, Action 28, applies if one or more process radiation monitors are inoperable. TS 3.3.3.1, Table 3.3-6, Action 28, states, in part, in the first paragraph that "operation of the plant may continue for up to 30 days provided grab samples are taken.... "

Also, the TS specifies that if the monitor is not restored to operable status within 30 days after the failure, that sampling will continue and a special report will be submitted within 14 days. The licensee seeks to revise the wording of TS 3.3.3.1, Table 3.3-6, Action 28, and delete "up to 30 days" from the first paragraph to clarify that the second paragraph allows operation to continue beyond 30 days, provided that the sampling is continued and the special report is submitted within the allotted time period. The licensee also seeks to revise the second paragraph to clarify that the 14-day constraint only begins if the monitor is not restored to operable status within 30 days after the failure by stating that the sampling will continue and a special report will be submitted within "the next" 14 days, if the monitor is not restored to operable status within 30 days.

The NRC staff has determined that the start time for both the 7-day and 14-day action times for TS 3.3.3.1, Table 3.3-6, Action 27.2, should be the same and at the time of instrument failure.

For TS 3.3.3.1, Table 3.3-6, Action 28, the NRC staff has also determined that plant operation is allowed beyond 30 days following instrument failure as long as a Special Report has been sent to the Commission as stated in the second paragraph of Action 28.

The NRC staff concludes that the proposed changes to the timing requirements associated with TS 3.3.3.1, Table 3.3-6, Actions 27.2 and 28 are acceptable because they clarify the existing TS, and continue to comply with GDC 13.

3.5 "TS 3.3.3.6, Accident Monitoring Instrumentation" TS 3.3.3.6, has an incorrect spelling of the word "Special" in Table 3.3-10, Action 29. The licensee seeks to revise TS 3.3.3.6, Table 3.3-10, Action 29, to correct this typographical error.

The NRC staff concludes this TS revision is acceptable because it corrects a typographical error which does not affect plant operation.

3.6 "TS 3.3.3.11, Explosive Gas Monitoring Instrumentation" TS 3.3.3.11, Action b, currently includes a conditional requirement to submit a Special Report to the Commission when instrumentation cannot be restored to an operable status within 30 days.

However, the action does not specify any time frame for preparation or submittal of this report, as is the case for other similar required actions in the WF3 TSs. The proposed change establishes a new 14-day window following the 30-day inoperable period to allow for preparation and submittal of the Special Report. This change is being made to provide consistency with similar Special Report actions in the WF3 TSs.

The NRC staff reviewed other actions within the licensee's TS requiring Special Report submittals and found that 14 days is the shortest time frame required for submission. The NRC staff concludes that a 14-day window is a conservative proposal because it is the shortest Special Report reporting time frame within the TSs and also provides a reasonable time frame for preparation and submittal of a Special Report.

The NRC staff concludes that the proposed change to TS 3.3.3.11, Action b, is acceptable because it results in a more conservative TS, and continues to comply with GDC 13.

3.7 TS 3/4.8.2. "D.C. Sources" In the original license amendment request, the license requested that an additional provision be added to requirement c.3 of SR TS 4.8.2.1 to provide assurance that the total battery resistance will not exceed the value required for battery bank operability. The request to modify SR TS 4.8.2.1, "D.C. Sources," was withdrawn by the licensee's supplemental letter dated December 17, 2014.

3.8 TS 6.1, "Responsibility," TS 6.2, "Organization." and TS 6.12. "High Radiation Area" Changes of position titles are being requested for multiple locations in the following Administrative Controls TSs:

In TS 6.1.2, the licensee requested a change of the title "Vice President Operations" to "Site Vice President - WF3" and a change of the title "Shift Superintendent" to "Shift Manager."

In TS 6.2.1.c, the licensee requested a change of the title "Vice President Operations" to "Site Vice President -WF3."

In TS 6.2.2.e, the licensee requested a change of the title "Assistant Operations Manager (Shift)" to "Operations Manager - Shift."

In TS 6.12.1.c, the licensee requested a change of the title "Radiation Protection Superintendent - Nuclear" to "Radiation Protection Manager," to be consistent with RG 1.8, "Personnel Selection and Training," Revision 1.

The NRC staff concludes that the title changes of the revised TSs 6.1.2, 6.2.1.c, 6.2.2.e, and TS 6.12.1.c are acceptable because the changes are administrative in nature, do not affect plant operations, and comply with RG 1.8.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Louisiana State official was notified of the proposed issuance of the amendment on December 29, 2014. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on August 5, 2014 (79 FR 45475). Also, the amendment changes administrative procedures or requirements, the position or title of an officer of the licensee, and makes editorial, corrective, or other minor revisions to the license. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 10 CFR 51.22(c)(10). 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Davida K. Cunanan Richard Stattel Date: February 23, 2015

Vice President, Operations Entergy Operations, Inc.

February 23, 2015 Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3-ISSUANCE OF AMENDMENT RE: MULTIPLE ADMINISTRATIVE ISSUES WITH THE TECHNICAL SPECIFICATIONS (TAC NO. MF3230)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 242 to Facility Operating License No. NPF-38 for the Waterford Steam Electric Station, Unit 3. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated December 9, 2013, as supplemented by letters dated October 1, 2014, and December 17, 2014.

The amendment revises certain TSs to improve clarity, correct administrative and typographical errors, or establish consistency with NUREG-1432, "Standard Technical Specifications -

Combustion Engineering Plants," Revision 4.0, dated April 2012. This amendment also revises TS Table 3.3-1, Action 6.b.1 to include one technical change.

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

Docket No. 50-382

Enclosures:

Sincerely, IRA!

Michael D. Orenak, Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

1. Amendment No. 242 to NPF-38

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

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