SBK-L-12052, Submittal of Changes to Technical Specification Bases

Submittal of Changes to Technical Specification Bases

ML12053A309
Person / Time
Site:	Seabrook
Issue date:	02/16/2012
From:	O'Keefe M NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-12052
Download: ML12053A309 (8)

Text

NExTeraM ENERGY SEABROOK February 16, 2012 Docket No. 50-443 SBK-L- 12052 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001 Seabrook Station Submittal of Chanaes to the Seabrook Station Technical Snecification Bases NextEra Energy Seabrook, LLC submits the enclosed changes to the Seabrook Station Technical Specification Bases. The changes were made in accordance with Technical Specification 6.7.6.j.,

"Technical Specification (TS) Bases Control Program." Please update the Technical Specification Bases as follows:

REMOVE INSERT Bases Index Page ii Bases Index Page ii Page B 3/4 4-29 Page B 3/4 4-29 Page B 3/4 4-31 Page B 3/4 4-31 Page B 3/4 6-6 Page B 3/4 6-6 Page B 3/4 8-23 Page B 3/4 8-23 Should you have any questions concerning this submittal, please contact me at (603) 773-7745.

Sincerely, NextEra Energy Seabrook, LLC Michael O'Keefe Licensing Manager NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

U. S. Nuclear Regulatory Commission SBK-L-12052/ Page 2 cc: NRC Region I Administrator NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Senior Resident Inspector

Enclosure to SBK-L-12052 INDEX BASES SECTION PAGE TABLE B 3/4.4-1 (THIS TABLE NUMBER IS NOT USED) ............................................ B 3/4 4-22 3/4.4.10 (THIS SPECIFICATION NUMBER IS NOT USED) ........................................... B 3/4 4-31 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ........... ............................. B 3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS........................................ B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK B 3/4 5-5 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT.................................. B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ................. B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.......................... B 3/4 6-3a 3/4.6.4 COMBUSTIBLE GAS CONTROL B 3/4 6-4 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING ....................... B 3/4 6-5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE........................................ B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION B 3/4 7-9 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM B 3/4 7-9 3/4.7.4 SERVICE WATER SYSTEM / ULTIMATE HEAT SINK B 3/4 7-10 3/4.7.5 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................ B 3/4 7-12 3/4.7.6 CONTROL ROOM SUBSYSTEMS B 3/4 7-12 3/4.7.7 SNUBBERS............................................ B 3/4 7-13 3/4.7.8 SEALED SOURCE CONTAMINATION.......................... B 3/4 7-15 3/4.7.9 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................ B 3/4 7-15 3/4.7.10 (THIS SPECIFICATION NUMBER IS NOT USED) ............................................ B 3/4 7-15 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A. C SOURCES B 3/4 8-1 3/4.8.2 D.C. SOURCES B 3/4 8-19 3/4.8.3 ONSITE POWER DISTRIBUTION B 3/4 8-20 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES B 3/4 8-23 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION...................................... B 3/4 9-2a 3/4.9.3 DECAY TIME B 3/4 9-2c 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS B 3/4 9-2d 3/4.9.5 COMMUNICATIONS B 3/4 9-3 3/4.9.6 REFUELING MACHINE..................................... B 3/4 9-3 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING B 3/4 9-3 SEABROOK - UNIT 1 ii BGR No. 000n2, BC 04 01, 07 04, 11-05

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE/TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION (Continued) provide a gravity feed to the RCS from the Refueling Water Storage Tank in the unlikely event that the CCP and Sl pumps were unavailable after a loss of RHR. Additionally, when steam generator nozzle dams are installed for maintenance purposes and the reactor vessel water level is not in a reduced inventory condition, a vent area equivalent to the opening with a safety valve removed limits RCS pressure during overpressure transients to reduce the possibility of adversely affecting steam generator nozzle dams.

When the reactor vessel head is on and the vessel head closure bolts are fully detensioned, a substantial vent area exists by the gap underneath the reactor vessel head, created by the internal spring forces. A minimum gap as specified in calculation C-S-1-84012 is sufficient to provide for cold overpressure protection, for gravity feed from the RWST, and ensuring nozzle dam integrity. Verification of sufficient gap will be performed prior to crediting the gap as a means for cold overpressure protection.

Cold overpressure protection can also be provided when operating at a reduced inventory condition, i.e., whenever reactor vessel water level is lower than 36 inches below the reactor vessel flange. With RCS water level lower than 36 inches below the RV flange in Mode 5 or Mode 6 with the RV head on and the closure bolts not fully detensioned, a mass addition transient involving simultaneous operation of a CCP and a Sl pump without letdown will not result in a cold overpressurization condition because of the relatively large void volume in the RCS. This void volume consists of the upper plenum of the reactor vessel and the RV head, the pressurizer and steam generator tubes, as a minimum. The relatively large void volume affords ample time for operator action, (e.g., diagnose the water level increase on main control board instrumentation and stopping the pumps) to mitigate the transient. A minimum time of 50 minutes has been determined based on one charging pump operating at 120 gpm without letdown and a Safety Injection pump injecting into the RCS.

The charging pumps and Safety Injection pumps are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the motor circuit breakers out under administrative control. An alternate method of preventing cold overpressurization may be employed. The alternate method uses at least two independent means to prevent cold overpressurization such that a single action will not result in an inadvertent injection into the RCS. This may be accomplished through the pump control switch being placed in Pull-to-Lock position and at least one valve in the discharge flow path closed. The alternate method provides the ability to respond to abnormal situations, expeditiously, from the main control room.

SEABROOK - UNIT 1 B 3/4 4-29 Amendment No. 74, BC 07 01, 07 02,7-03, 11-04

REACTOR COOLANT SYSTEM BASES 3/4.4.10 This specification number is not used SEABROOK - UNIT 1 B 3/4 4-31 ACmedme0 t No. 79, BC 04 03, BC 06 02,T07 11-05

CONTAINMENT SYSTEMS BASES 3/4.6.5.3 CONTAINMENT ENCLOSURE BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment enclosure building will be maintained comparable to the original design standards for the life of the facility.

The function of the containment enclosure building is to collect any fission products that leak from the primary containment structure into the containment enclosure and contiguous areas following an accident. The containment enclosure provides a low leakage rate barrier between the containment and the environment. Structural integrity of the containment enclosure building is necessary to prevent leakage of radioactive materials from the containment enclosure building. A visual inspection of the exposed interior and exterior concrete surfaces of the containment enclosure structure in accordance with the Containment Leakage Rate Testing Program, is sufficient to demonstrate this capability.

SEABROOK - UNIT 1 B 3/4 6-6 BC 04 -12, 10-03

ELECTRICAL POWER SYSTEMS BASES 3/4.8.3 ONSITE POWER DISTRIBUTION (continued)

ACTIONS (continued)

MODES 5 and 6 With less than the minimum required on-site power distribution systems sources, the action statement requires immediately suspending core alterations, positive reactivity changes, or movement of irradiated fuel. With respect to suspending positive reactivity changes, operations that individually add limited, positive reactivity are acceptable when, combined with other actions that add negative reactivity, the overall net reactivity addition is zero or negative. For example, a positive reactivity addition caused by temperature fluctuations from inventory addition or temperature control fluctuations is acceptable if it is combined with a negative reactivity addition such that the overall, net reactivity addition is zero or negative. Refer to TS Bases 3/4.9.1, Boron Concentration, for limits on boron concentration and water temperature for MODE 6 action statements involving suspension of positive reactivity changes.

SURVEILLANCE REQUIREMENTS Operability of the required electrical buses is confirmed by verifying correct breaker alignment and indicated voltage on the buses at least once per seven days. The seven-day frequency is based on the capability of the electrical systems and the indications available in the control room that alert the operator to electrical system malfunctions.

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment electrical penetrations are protected by deenergizing circuits not required during reactor operation. The OPERABILITY of the motor-operated valves thermal overload protection ensures that the thermal overload protection will not prevent safety-related valves from performing their function. The Surveillance Requirements for demonstrating the OPERABILITY of the thermal overload protection are in accordance with Regulatory Guide 1.106, "Thermal Overload Protection for Electric Motors on Motor Operated Valves," Revision 1, March 1977.

SEABROOK - UNIT 1 B 3/4 8-23 BC 04 15, 11 -01