SBK-L-06244, Submittal of Changes to Seabrook Station Technical Specification Bases

Submittal of Changes to Seabrook Station Technical Specification Bases

ML070080150
Person / Time
Site:	Seabrook
Issue date:	01/03/2007
From:	St.Pierre G Florida Power & Light Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-06244
Download: ML070080150 (5)

Text

0FPL EIe(rg FPL Energy Seabrook Station RO. Box 300 Seabrook, NH 03874 Seabrook Station (603) 773-7000 January 3, 2007 Docket No. 50-443 SBK-L-06244 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001 Seabrook Station Submittal of Changes to the Seabrook Station Technical Specification Bases FPL 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 Specifications Bases as follows:

Remove Insert Page B 3/4 4-17 Page B 3/4 4-17 Page B 3/4 7-2 Page B 3/4 7-2 Page B 3/4 9-4 Page B 3/4 9-4 Should you have any questions concerning this matter, please contact James M. Peschel, Regulatory Programs Manager, at (603) 773-7194.

Very truly yours, FPL Energy Seabrook, LLC Gene St. Pierre Site Vice President cc: S. J. Collins, NRC Region I Administrator E. Miller, NRC Project Manager, Project Directorate 1-2 G.T. Dentel, NRC Senior Resident Inspector an FPL Group company

Enclosure to SBK-L-06244 REACTOR COOLANT SYSTEM BASES 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

This TS requires maintaining the structural integrity of components classified as ASME Class 1, 2, or 3 under ASME Section XI and applies to any systems that contain these ASME Class components. These components were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. As stated in Appendix H of WCAP-14535A (November 1996), Appendix VIll of Section XI of the ASME Boiler and Pressure Vessel Code is not applicable when examining the reactor coolant pump flywheels.

If a flaw or pressure boundary leakage is discovered in an ASME Class 1, 2, or 3 component, the condition should be promptly evaluated in accordance with station procedures for degraded or non-conforming conditions. The actions of this TS are applicable to the condition if the evaluation determines that the flaw or leakage is unacceptable and results in a failure to meet the LCO for structural integrity.

The failure of an ASME Class 1 or 2 component to meet structural integrity requirements requires entry into action a or b, respectively. These actions require restoring structural integrity or isolating the affected component prior to increasing temperature above a specified temperature. If the failure of an ASME Class 1 or 2 component to meet structural integrity requirements is discovered with the component not isolated and above the minimum temperature specified in the associated action, TS 3.0.3 is applicable until the conditions specified in the action are satisfied.

Action c addresses the failure of an ASME Class 3 component to conform to the structural integrity requirements and requires restoring structural integrity of the component or isolating the component from service. While the action specifies no completion time, the actions should be timely and on a schedule commensurate with the safety significance of the affected component.

SEABROOK - UNIT 1 B 3/4 4-17Amendment No. 79, BC 04 03, BC 06-02

PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE (Continued) 3/4.7.1.1 SAFETY VALVES The steam flow rate through each operable MSSV is calculated with consideration of 1) opening setpoint with allowance for as-found setpoint tolerance, 2) accumulation, and 3) inlet pressure. For example, if the maximum number of inoperable MSSVs on any one steam generator is one, then W, is a summation of the calculated flow through each MSSV at the appropriate valve inlet pressure, excluding the MSSV with the highest flow. If the maximum number of inoperable MSSVs per steam generator is three, then Ws is a summation of the calculated flow through each MSSV at the appropriate valve inlet pressure, excluding the three MSIVs with the highest flows. The following plant specific safety valve flow rates were used:

SG Safety Main Steam System Valve Number (Bank No.) Set Pressure (psia) Flow (Ibm/hr per loop) 1 1200 893,160 2 1210 900,607 3 1220 908,055 4 1230 915,502 5 1240 922,950 The Safety Analysis limit values of the power range high neutron flux setpoints calculated from this algorithm are adjusted lower for use in Technical Specification 3.7.1.1. to account for instrument and channel uncertainties.

Surveillance Requirement 4.7.1.1 verifies operability of the safety valves with a verification of each safety valve lift setpoint in accordance with the Inservice Testing (IST) Program. Table 3.7-2 specifies the safety valve lift settings with a

+ 3% tolerance. This lift setting establishes the operability criterion for the safety valves during as-found setpoint verifications. Additionally, this tolerance is used to determine the need for testing of additional safety valves as required by the IST Program. The lift setting in Table 3.7-2 is modified by a footnote that states the setpoint tolerance is + 1% following safety valve testing.

Following an as-found setpoint verification, the lift setting is adjusted to + 1% in accordance with the footnote to allow for drift. However, the safety valves are capable of performing their design function and, therefore, are operable, when found with a setpoint that is within + 3% of the specified lift setting.

SEABROOK - UNIT 1 B 3/4 7-2 Amendment No. 43, BC 06-01

3/4.9 REFUELING OPERATIONS (Continued)

BASES 3/4.9.9 (THIS SPECIFICATION NUMBER IS NOT USED.)

3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis. Suspending fuel movement or crane operation does not preclude moving a component to a safe location.

3/4.9.12 FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM The limitations on the Fuel Storage Building Emergency Air Cleaning System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing. Suspending fuel movement or crane operation does not preclude moving a component to a safe location.

3/4.9.13 SPENT FUEL ASSEMBLY STORAGE Restrictions on placement of fuel assemblies of certain enrichments within the Spent Fuel Pool is dictated by Figure 3.9-1. These restrictions ensure that the Keff of the Spent Fuel Pool will always remain less than 0.95 assuming the pool to be flooded with unborated water. The restrictions delineated in Figure 3.9-1 and the action statement are consistent with the criticality safety analysis performed for the Spent Fuel Pool as documented in the FSAR.

3/4.9.14 NEW FUEL ASSEMBLY STORAGE Restrictions on placement of fuel assemblies of certain enrichments within the New Fuel Storage Vault is dictated by Specification 3/4.9.14. These restrictions ensure that the Keff of the New Fuel Storage Vault will always remain less than 0.95 assuming the area to be flooded with unborated water. In addition, these restrictions ensure that the Keff of the New Fuel Storage Vault will always remain less than 0.98 when aqueous foam moderation is assumed. The restrictions delineated in Specification 3/4.9.14 and the action statement are consistent with the criticality safety analysis performed for the New Fuel Storage Vault as documented in the FSAR.

SEABROOK - UNIT 1 B 3/4 9-4 EAmendment 6, 85, BC 05-07