ML18082B377

Attachment - Vogtle Electric Generating Plant Unit 3 Amendment 134(LAR-18-005)

ML18082B377
Person / Time
Site:	Vogtle
Issue date:	07/19/2018
From:	Office of New Reactors
To:	Southern Nuclear Operating Co
kallan p/415-2809
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ML18082B374	List: ML18082B375 ML18082B376 ML18082B377 ML18082B378 ML18082B379 ML18082B380
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ATTACHMENT TO LICENSE AMENDMENT NO. 134 TO FACILITY COMBINED LICENSE NO. NPF-91 DOCKET NO.52-025 Replace the following pages of the Facility Combined License No. NPF-91 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Combined License No. NPF-91 REMOVE INSERT 7 7 Appendix A to Facility Combined License Nos. NPF-91 and NPF-92 REMOVE INSERT 1.1-5 1.1-5 3.1.4-4 3.1.4-4 3.1.7-1 3.1.7-1 3.1.7-2 3.1.7-2 3.1.7-3 3.1.7-3 3.2.3-1 3.2.3-1 3.2.4-3 3.2.4-3 4.0-1 4.0-1

(7) Reporting Requirements (a) Within 30 days of a change to the initial test program described in UFSAR Section 14, Initial Test Program, made in accordance with 10 CFR 50.59 or in accordance with 10 CFR Part 52, Appendix D, Section VIII, Processes for Changes and Departures, SNC shall report the change to the Director of NRO, or the Directors designee, in accordance with 10 CFR 50.59(d).

(b) SNC shall report any violation of a requirement in Section 2.D.(3),

Section 2.D.(4), Section 2.D.(5), and Section 2.D.(6) of this license within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made to the NRC Operations Center in accordance with 10 CFR 50.72, with written follow up in accordance with 10 CFR 50.73.

(8) Incorporation The Technical Specifications, Environmental Protection Plan, and ITAAC in Appendices A, B, and C, respectively of this license, as revised through Amendment No. 134, are hereby incorporated into this license.

(9) Technical Specifications The technical specifications in Appendix A to this license become effective upon a Commission finding that the acceptance criteria in this license (ITAAC) are met in accordance with 10 CFR 52.103(g).

(10) Operational Program Implementation SNC shall implement the programs or portions of programs identified below, on or before the date SNC achieves the following milestones:

(a) Environmental Qualification Program implemented before initial fuel load; (b) Reactor Vessel Material Surveillance Program implemented before initial criticality; (c) Preservice Testing Program implemented before initial fuel load; (d) Containment Leakage Rate Testing Program implemented before initial fuel load; (e) Fire Protection Program

1. The fire protection measures in accordance with Regulatory Guide (RG) 1.189 for designated storage building areas (including adjacent fire areas that could affect the storage area) implemented before initial receipt 7 Amendment No. 134

Technical Specifications Definitions 1.1 1.1 Definitions REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of gripper coils voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single assembly of highest reactivity worth, which is assumed to be fully withdrawn.

However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCAs not capable of being fully inserted, the reactivity worth of these assemblies must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

c. In MODE 2 with keff < 1.0, and MODES 3, 4, and 5, the worth of fully inserted Gray Rod Cluster Assemblies (GRCAs) will be included in the SDM calculation.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

VEGP Units 3 and 4 1.1 - 5 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 -----------------------------------------------------------------------

- NOTE -

Not required to be performed for rods associated with inoperable digital rod position indication or demand position indication.

Verify individual rod positions within alignment limit. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.1.4.2 -----------------------------------------------------------------------

- NOTE -

Not applicable to GRCAs.

Verify rod freedom of movement (trippability) by 92 days moving each rod not fully inserted in the core 10 steps in either direction.

SR 3.1.4.3 -----------------------------------------------------------------------

- NOTE -

Not applicable to GRCAs.

Verify rod drop time of each rod, from the fully Once prior to withdrawn position, is 2.7 seconds from the reactor criticality beginning of decay of gripper coil voltage to dashpot after each removal entry, with: of the reactor head, and after

a. Tavg 500°F, and each earthquake requiring plant

b. All reactor coolant pumps operating. shutdown VEGP Units 3 and 4 3.1.4 - 4 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications Rod Position Indication 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO 3.1.7 The Digital Rod Position Indication (DRPI) for each control rod and the Bank Demand Position Indication for each group shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each inoperable DRPI and each demand position indication.

CONDITION REQUIRED ACTION COMPLETION TIME A. One DRPI per group A.1 Verify the position of the Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable in one or rods with inoperable DRPI more groups. indirectly by using the incore detectors.

OR A.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to 50% RTP.

B. More than one DRPI B.1 Place the control rods Immediately per group inoperable in under manual control.

one or more groups.

AND B.2 Monitor and record Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Reactor Coolant System (RCS) Tavg.

AND B.3 Restore inoperable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DRPI(s) to OPERABLE status such that a maximum of one DRPI per group is inoperable.

VEGP Units 3 and 4 3.1.7 - 1 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications Rod Position Indication 3.1.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more rods with C.1 Verify the position of the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable DRPI have rods with inoperable DRPI been moved in excess indirectly by using the of 24 steps in one incore detectors.

direction since the last determination of the OR rods position.

C.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to 50% RTP.

D. One or more demand D.1.1 Verify by administrative Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> position indication per means all DRPIs for the bank inoperable in one affected banks are or more banks. OPERABLE.

AND D.1.2 Verify the most withdrawn Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> rod and the least withdrawn rod of the affected banks are 12 steps apart.

OR D.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to 50% RTP.

E. Required Action E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and associated Completion Time not met.

VEGP Units 3 and 4 3.1.7 - 2 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications Rod Position Indication 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 -----------------------------------------------------------------------

- NOTE -

Not required to be met for DRPI associated with a rod that does not meet LCO 3.1.4, Rod Group Alignment Limits.

Verify each DRPI agrees within 12 steps of the group Once prior to demand position for the full indicated range of rod criticality after travel. each removal of the reactor head VEGP Units 3 and 4 3.1.7 - 3 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications AFD (CAOC Methodology) 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Constant Axial Offset Control (CAOC) Methodology)

LCO 3.2.3 The AFD:

a. Shall be maintained within the target band specified in the COLR about the target flux difference.

b. May deviate outside the target band with THERMAL POWER

< 90% RTP, but 50% RTP, provided AFD is within the acceptable operation limits specified in the COLR and cumulative penalty deviation time is 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c. May deviate outside the target band with THERMAL POWER

< 50% RTP.

- NOTES -

1. The AFD shall be considered outside the target band when two or more OPERABLE excore channels indicate AFD to be outside the target band.

2. With THERMAL POWER 50% RTP, penalty deviation time shall be accumulated on the basis of a 1 minute penalty deviation for each 1 minute of power operation with AFD outside the target band.

3. With THERMAL POWER < 50% RTP and > 15% RTP, penalty deviation time shall be accumulated on the basis of a 0.5 minute penalty deviation for each 1 minute of power operation with AFD outside the target band.

4. A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of operation may be accumulated with AFD outside the target band without penalty deviation time during surveillance of Power Range Neutron Flux channels in accordance with SR 3.3.1.5, provided AFD is maintained within acceptable operation limits.

APPLICABILITY: MODE 1 with THERMAL POWER > 15% RTP and with the On-Line Power Distribution Monitoring System (OPDMS) not monitoring parameters.

VEGP Units 3 and 4 3.2.3 - 1 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications QPTR 3.2.4 SURVEILLANCE REQUIREMENTS

- NOTE -

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the last verification of OPDMS parameters.

SURVEILLANCE FREQUENCY SR 3.2.4.1 -----------------------------------------------------------------------

- NOTES -

1. With one Power Range Neutron Flux channel inoperable and THERMAL POWER < 75% RTP, the remaining three Power Range Neutron Flux channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

Verify QPTR within limit by calculation. 7 days SR 3.2.4.2 -----------------------------------------------------------------------

- NOTE -

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER 75% RTP.

Verify QPTR is within limit using a minimum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 4 symmetric pairs of fixed incore detectors.

VEGP Units 3 and 4 3.2.4 - 3 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)

Technical Specifications Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site The site for the Vogtle Electric Generating Plant (VEGP), is located in eastern Burke County, Georgia; approximately 26 miles southeast of Augusta, Georgia and 100 miles northwest of Savannah, Georgia; directly across the Savannah River from the US Department of Energys Savannah River Site in Barnwell County, South Carolina.

4.1.1 Site and Exclusion Boundaries The 3,169-acre VEGP site is located on a coastal plain bluff on the southwest side of the Savannah River in eastern Burke County. The site exclusion area boundary (EAB) is bounded by River Road, Hancock Landing Road and 1.7 miles of the Savannah River (River Miles 150.0 to 151.7). The property boundary entirely encompasses the EAB and extends beyond River Road in some areas.

4.1.2 Low Population Zone (LPZ)

The LPZ is defined by the 2-mile-radius circle from the midpoint between the containment buildings of Units 1 and 2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of fuel rods clad with a zirconium based alloy and containing an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.

Limited substitutions of zirconium based alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod and Gray Rod Assemblies The reactor core shall contain 53 Rod Cluster Control Assemblies (RCCAs), each with 24 rodlets/RCCA. The RCCA absorber material shall be silver indium cadmium as approved by the NRC.

Additionally, there are 16 low worth Gray Rod Cluster Assemblies (GRCAs), with 24 rodlets/GRCA, which, in conjunction with the RCCAs, are used to augment mechanical shim (MSHIM) operation.

VEGP Units 3 and 4 4.0 - 1 Amendment No. 134 (Unit 3)

Amendment No. 133 (Unit 4)