W3F1-2011-0080, Technical Specification Bases Update to the NRC for the Period January 25, 2011 Through October 31, 2011

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Technical Specification Bases Update to the NRC for the Period January 25, 2011 Through October 31, 2011
ML113130367
Person / Time
Site: Waterford Entergy icon.png
Issue date: 11/09/2011
From: Steelman W
Entergy Nuclear South, Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2011-0080
Download: ML113130367 (18)
v • d • e


Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504 739 6685 Fax 504 739 6698 wsteelm@entergy.com William J. Steelman Licensing Manager Waterford 3 W3F1-2011-0080 November 9, 2011 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Technical Specification Bases Update to the NRC for the Period January 25, 2011 through October 31, 2011 Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

Dear Sir or Madam:

Pursuant to Waterford Steam Electric Station Unit 3 Technical Specification (TS) 6.16, Entergy Operations, Inc. (EOI) hereby submits an update of all changes made to Waterford 3 Technical Specification Bases since the last submittal per letter W3F1-2011-0010 (ADAMS Accession #ML110270215), dated January 24, 2011. This TS Bases update satisfies the submittal frequency requirement listed in 10CFR50.71(e)(4).

There are no commitments associated with this submittal. Should you have any questions or comments concerning this submittal, please contact William J. Steelman at (504) 739-6685.

Sincerely, WJS/RJP/ssf Attachments:

1. Waterford 3 Technical Specification Bases Change List
2. Waterford 3 Technical Specification Bases Revised Pages

W3F1-2011-0080 Page 2 cc: Mr. Elmo E. Collins, Jr. RidsRgn4MailCenter@nrc.gov Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector Marlone.Davis@nrc.gov Waterford Steam Electric Station Unit 3 Dean.Overland@nrc.gov P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Kaly.Kalyanam@nrc.gov Attn: Mr. N. Kalyanam Mail Stop O-07D1 Washington, DC 20555-0001

Attachment 1 to W3F1-2011-0080 Waterford 3 Technical Specification Bases Change List to W3F1-2011-0080 Page 1 of 2 Waterford 3 Technical Specification (TS) Bases Change List T.S. Implementation Affected TS Topic of Change Bases Date Bases Pages Change No.

68 03/16/11 B 3/4 7-2c Change No. 68 to TS Bases section B 3/4 7-2d 3/4.7.1.2 Emergency Feedwater B 3/4 7-2e System was implemented by B 3/4 7-2f Engineering Change (EC) 27828. The change addresses the License Amendment 230, which added ACTION (g) to Technical Specification 3/4.7.1.2 to indicate that the provisions of Specifications 3.0.4 and 4.0.4 are not applicable to the turbine-driven Emergency Feedwater (EFW) pump for entry into Mode 3 only as allowed by Surveillance Requirements 4.7.1.2(b) and 4.7.1.2(c). Change No. 68 to TS Bases section 3/4.7.1.2 explains that ACTION (g) is intended to clarify verbatim compliance issue associated with entering Mode 3 and not having performed the operability testing for the turbine-driven EFW pump. The exception to TS LCO 3.0.4 and 4.0.4 during plant startup allows the turbine-driven EFW pump to be INOPERABLE only when entering Mode 3 under the conditions and for the period (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 750 psig in both steam generators) as contained in TSSR 4.7.1.2(b) and 4.7.1.2(c),

quarterly Inservice Testing (IST) and 18 month Engineered Safety Features Actuation System Instrumentation -

Functional Unit Emergency Feedwater (EFAS), respectively.

to W3F1-2011-0080 Page 2 of 2 T.S. Implementation Affected TS Topic of Change Bases Date Bases Pages Change No.

69 05/24/11 IX Change No. 69 to TS Bases section XIV 3/4.9.4, "Containment Building B 3/4 9-2 Penetrations" and 3/4.9.9, Containment B 3/4 9-3 Purge Valve Isolation System was implemented by EC 28875. The B 3/4 9-4 change addresses License Amendment B 3/4 9-5 231, which was related to Refueling Outage Penetration Protection. The amendment modified Technical Specification (TS) 3/4.9.4, "Containment Building Penetrations" to allow alternative means of penetration closure during Core Alterations or irradiated fuel movement while in refueling operations.

Additionally TS 3/4.9.9, Containment Purge Valve Isolation System was eliminated and those actions and surveillance requirements being retained were moved into TS 3/4.9.4 and the associated TS Bases. In addition, the TS Index Pages for the changes to TS 3/4.9.4 and deletion of TS 3/4.9.9 are included in the page changes.

Attachment 2 to W3F1-2011-0080 Waterford 3 Technical Specification Bases Revised Pages (There are 12 unnumbered pages following this cover page)

TECHNICAL SPECIFICATION BASES CHANGE NO. 68 REPLACEMENT PAGE(S)

(4 pages)

Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 68 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert B 3/4 7-2c B 3/4 7-2c B 3/4 7-2d B 3/4 7-2d B 3/4 7-2e B 3/4 7-2e B 3/4 7-2f B 3/4 7-2f

PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM (Continued)

Limiting Conditions for Operation (Continued)

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time is reasonable based on the redundant capabilities afforded by the EFW system, the time needed for repairs, and the low probability of a design basis event occurring during this period.

e. By maintaining OPERABLE pumping capacity capable of supplying 100% of the required EFW flow and flow paths capable of delivering 100% of the required EFW flow to the steam generators the EFW system is capable of supporting a unit cooldown but may not be capable of performing its design function of residual heat removal for all events. Due to the seriousness of this condition, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed completion time is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

This ACTION primarily addresses flow path inoperability when the system no longer has the ability to deliver 100% of the required EFW flow to one or both steam generators. For example, with one flow path inoperable and not able to provide 100% flow to its respective steam generator this ACTION would be entered. Similarly, if both flow paths were inoperable and only one of the inoperable flow paths could provide 100% of the required EFW flow to its respective steam generator this ACTION would be entered.

Also, if both flow paths were inoperable and neither could provide 100% of the required EFW flow to its respective steam generator but together both flow paths could provide 100% of the required EFW flow to the steam generators (e.g., 50% to one and 50% to the other (or some combination equaling 100%)) this ACTION would be entered.

f. ACTION (f) indicates that all required MODE changes or power reductions are suspended until the EFW system is capable of delivering 100% of the required EFW flow to the steam generators.

With pumping capacity unable to supply 100% of the required EFW flow and/or two flow paths not capable of delivering 100% of the required EFW flow to the steam generators in MODEs 1, 2, and 3, the unit is in a seriously degraded condition with no safety-related means for conducting a cooldown. In such a condition, the unit should not be perturbed by any action, including a power change that might result in a trip. The seriousness of this condition requires that action be started immediately to restore the ability to deliver at least 100% of the required EFW flow to the steam generators combined as soon as possible. This ACTION is modified to indicate that all MODE changes or power reductions are suspended until the ability to deliver 100% of the required flow to the steam generators combined can be restored because they could force the unit into a less than safe condition.

>(EC-27828, Ch. 68)

g. ACTION (g) is intended to clarify verbatim compliance issue associated with entering Mode 3 and not having performed the operability testing for the turbine-driven EFW pump.

The exception to TS LCO 3.0.4 and 4.0.4 during plant startup allows the turbine-

<(EC-27828, Ch. 68)

WATERFORD - UNIT 3 B 3/4 7-2c CHANGE NO. 7, 68

PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM (Continued)

Limiting Conditions for Operation (Continued)

>(EC-27828, Ch. 68) driven EFW pump to be INOPERABLE only when entering Mode 3 under the conditions and for the period (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 750 psig in both steam generators) as contained in TSSR 4.7.1.2(b) and 4.7.1.2(c), quarterly Inservice Testing (IST) and 18 month Engineered Safety Features Actuation System Instrumentation - Functional Unit Emergency Feedwater (EFAS), respectively. When the plant enters Mode 3 during a plant startup coming out of an outage, there is insufficient steam pressure to complete the dynamic final calibration of the governor valve speed control unit of the turbine-driven EFW pump. In this condition, the turbine-driven EFW pump is available (i.e.,

there is a reasonable expectation that once sufficient steam pressure is available to the turbine-driven EFW pump turbine, it will be able to successfully complete the quarterly IST and 18 month EFAS surveillance requirements to fully demonstrate operability).

Although the turbine-driven EFW pump does not have sufficient steam pressure to complete dynamic final calibration of the governor valve speed control unit, the turbine-driven EFW pump still maintains performance capability (albeit at a potentially reduced flow performance based upon governor valve speed control unit settings) to provide the system safety function of cooling the plant to shutdown cooling entry conditions. This exception does not allow Mode 3 to be entered during a plant startup while performing maintenance activities that cause the turbine driven EFW pump to be unavailable.

The safety function of the EFW System to ensure the Reactor Coolant System can be cooled to shutdown cooling system entry conditions continues to be met under all plant conditions and for the worst case postulated accident from the point in time when the plant enters Mode 3 during the plant startup with the inoperable turbine-driven EFW pump through the point in time when the turbine-driven EFW pump is restored to OPERABLE condition. The delay of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both steam generators have reached sufficient steam pressure on the secondary side is to complete post maintenance activities (i.e., dynamic final calibration of the governor valve speed control unit) and then to complete IST and EFAS testing surveillance requirements.

Prior to entry into Mode 2, surveillance requirement testing of various combinations of EFW pumps and valves will ensure ALL required EFW system flow paths and equipment (including the turbine-driven EFW pump as previously described in second paragraph of this section of the bases) are demonstrated operable before the core is taken critical and significant heat is generated.

<(EC-27828, Ch. 68)

Surveillance Requirements

a. Verifying the correct alignment for manual, power operated, and automatic valves in the EFW water and steam supply flow paths provides assurance that the proper flow paths exist for EFW operation. This Surveillance Requirement (SR) does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.

WATERFORD - UNIT 3 B 3/4 7-2d CHANGE NO. 7, 30, 37, 68

PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM (Continued)

Surveillance Requirements (Continued)

>(DRN 03-1807, Ch. 30)

b. The SR to verify pump OPERABILITY pursuant to the Inservice Testing Program ensures that the requirements of ASME Code Section XI are met and provides reasonable assurance that the pumps are capable of satisfying the design basis accident flow requirements. Because it is undesirable to introduce cold EFW into the steam generators while they are operating, testing is typically performed on recirculation flow. Such in-service tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

<(DRN 03-1807, Ch. 30)

This SR is modified to indicate the SR should be deferred until suitable test conditions have been established. This deferral is required because there is an insufficient steam pressure to perform post maintenance activities which may need to be completed prior to performing the required turbine-driven pump SR. This deferral allows the unit to transition from MODE 4 to MODE 3 prior to the performance of the SR and provides a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period once a steam generator pressure of 750 psig is reached to complete the required post maintenance activities and SR. If this SR is not completed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or fails, then the appropriate ACTION must be entered. The twenty-five percent grace period allowed by TS 4.0.2 can not be applied to the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

>(DRN 05-42, Ch. 37)

c. The SR for actuation testing ensures that EFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates EFAS and/or MSIS signals, by demonstrating that each automatic valve in the flow path actuates to its correct position and that the EFW pumps will start on an actual or simulated actuation signal. This Surveillance covers the automatic flow control valves, automatic isolation valves, and steam admission valves but is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month frequency is acceptable, based on the design reliability and operating experience of the equipment.

<(DRN 05-42, Ch. 37)

This SR is modified to indicate that the SR should be deferred until suitable test conditions have been established. This deferral is required because there is an insufficient steam pressure to perform post maintenance activities which may need to be completed prior to performing the required turbine-driven pump SR. This deferral allows the unit to transition from MODE 4 to MODE 3 prior to the performance of the SR and provides a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period once a steam generator pressure of 750 psig is reached to complete the required post maintenance activities and SR. If this SR is not completed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or fails, then the appropriate ACTION must be entered. The twenty-five percent grace period allowed by TS 4.0.2 can not be applied to the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

WATERFORD - UNIT 3 B 3/4 7-2e CHANGE NO. 7, 68

PLANT SYSTEMS BASES 3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM (Continued)

Surveillance Requirements (Continued)

d. The SR for flow testing ensures that the EFW system is aligned properly by verifying the flow paths from the condensate storage pool (CSP) to each steam generator before entering MODE 2 operation after being in MODE 4, 5, 6, or defueled, for 30 days or longer, or whenever feedwater line cleaning through the emergency feedwater line has been performed. Various combinations of pumps and valves may be used such that all flow paths (and flow legs) are tested at least once during the Surveillance.

OPERABILITY of EFW flow paths must be verified before sufficient core heat is generated that would require the operation of the EFW System during a subsequent shutdown. The frequency is reasonable, based on engineering judgment, and other administrative controls to ensure that flow paths remain OPERABLE. To further ensure EFW system alignment, the OPERABILITY of the flow paths is verified following extended outages to determine that no misalignment of valves has occurred. This SR ensures that the flow paths from the CSP to the steam generators are properly aligned.

3/4.7.1.3 CONDENSATE STORAGE POOL The OPERABILITY of the condensate storage pool (CSP) with the minimum water volume of 173,500 gallons (170,000 gallons for EFW system usage and 3,500 gallons for CCW makeup system usage), plus makeup from one Wet Cooling Tower (WCT) basin, ensures that sufficient water is available to cool the Reactor Coolant System to shutdown cooling entry conditions following any design basis accident. This makeup water includes the capability to maintain HOT STANDBY for at least an additional 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to initiating shutdown cooling.

The combined capacity (CSP and one WCT) provides sufficient cooling for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until shutdown cooling is initiated in the event the ultimate heat sink sustains tornado damage concurrent with the tornado event.

If natural circulation is required, the combined capacity (CSP and one WCT) is sufficient to maintain the plant at HOT STANDBY for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by a cooldown to shutdown cooling entry conditions assuming the availability of only onsite or only offsite power, and the worst single failure (loss of a diesel generator or atmospheric dump valve). This requires approximately 303,000 gallons of EFW and complies with BTP RSB 5-1.

>(DRN 04-1243, Ch. 38)

The CSP contained water volume limit (92% indicated in MODES 1, 2, and 3) includes an allowance for water not usable because of vortexing and instrumentation uncertainties. This provides an assurance that a minimum of 170,000 gallons is available for the EFW system and that 3,500 gallons is available for the CCW makeup system. The CSP contained water volume limit (11% indicated in MODE 4) also includes an allowance for water not usable because of vortexing and instrumentation uncertainties. This provides an assurance that minimum of 3,500 gallons is available in the CSP for the CCW makeup system.

The maximum limit on CSP temperature ensures that the assumptions used in design basis accidents with EFW flow remain valid. The minimum limit on CSP temperature ensures that the assumptions used in the MSLB return to power event remain valid.

<(DRN 04-1243, Ch. 38)

WATERFORD - UNIT 3 B 3/4 7-2f CHANGE NO. 7, 38, 68

TECHNICAL SPECIFICATION BASES CHANGE NO. 69 REPLACEMENT PAGE(S)

(6 pages)

Replace the following pages of the Waterford 3 Technical Specification Bases with the attached pages. The revised pages are identified by Change Number 69 and contain the appropriate EC number and a vertical line indicating the areas of change.

Remove Insert IX IX XIV XIV B 3/4 9-2 B 3/4 9-2 B 3/4 9-3 B 3/4 9-3 B 3/4 9-4 B 3/4 9-4 B 3/4 9-5 B 3/4 9-5

>(DRN 05-747, Ch. 40)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8 ELECTRICAL POWER SYSTEMS (Continued) 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS OPERATING........................................................................ 3/4 8-13 SHUTDOWN........................................................................ 3/4 8-15 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.................... 3/4 8-16 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES................................... 3/4 8-52 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.................................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION.............................................................. 3/4 9-2 3/4.9.3 DECAY TIME.......................................................................... 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ 3/4 9-4 3/4.9.5 COMMUNICATIONS................................................................ 3/4 9-5 3/4.9.6 REFUELING MACHINE........................................................... 3/4 9-6 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING.................... 3/4 9-7 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL........................................................... 3/4 9-8 LOW WATER LEVEL............................................................ 3/4 9-9

>(EC-28875, Am. 69) 3/4.9.9 DELETED................................................................................ 3/4 9-10

<(EC-28875, Am. 69) 3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES............................................................ 3/4 9-11 CEAs..................................................................................... 3/4 9-12 3/4.9.11 WATER LEVEL - SPENT FUEL POOL...................................... 3/4 9-13

>(EC-18742, Ch. 65) 3/4.9.12 SPENT FUEL POOL (SFP) BORON CONCENTRATION............ 3/4 9-13a 3/4.9.13 SPENT FUEL STORAGE............................................................. 3/4 9-13b

<(EC-18742, Ch. 65) 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN................................................................. 3/4 10-1

<(DRN 05-747, Ch. 40)

>(DRN 05-747, Ch. 40)

WATERFORD - UNIT 3 IX AMENDMENT NO. 176

<(DRN 05-747, Ch. 40) CHANGE NO. 40, 65, 69

>(DRN 05-747, Ch. 40)

INDEX BASES SECTION PAGE 3/4.9 REFUELING OPERATIONS (Continued)

>(EC-28875, Am. 69) 3/4.9.9 DELETED 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and SPENT FUEL POOL............................................................ B 3/4 9-4

<(EC-28875, Am. 69) 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN......................................................... B 3/4 10-1 3/4.10.2 MTC, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS....................................... B 3/4 10-1 3/4.10.3 REACTOR COOLANT LOOPS........................................... B 3/4 10-1 3/4.10.4 CENTER CEA MISALIGNMENT........................................... B 3/4 10-1 3/4.10.5 NATURAL CIRCULATION TESTING.................................... B 3/4 10-1 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS............................................................ B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS...................................................... B 3/4 11-3

<(DRN 05-747, Ch. 40)

>(DRN 05-747, Ch. 40)

WATERFORD - UNIT 3 XIV AMENDMENT NO 68,176, 188

<(DRN 05-747, Ch. 40) CHANGE NO. 40, 69

REFUELING OPERATIONS BASES 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS (Continued)

>(DRN 03-178, Ch. 21) closure rather than containment OPERABILITY. Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

During CORE ALTERATIONS or movement of irradiated fuel within the containment, the escape of radioactivity to the environment is minimized when the LCO requirements are met.

>(EC-28875, Ch. 69)

Containment penetrations, the personnel airlock doors, and/or the equipment door may be open under administrative control during CORE ALTERATIONS or movement of irradiated fuel in the containment provided a minimum of one closure method (manual or automatic valve, blind flange, or equivalent) in each penetration, one door in each airlock, and the equipment door are capable of being closed in the event of a fuel handling accident. For closure, the equipment door will be held in place by a minimum of four symmetrically-placed bolts.

Containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated or capable of being isolated on at least one side. Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure ventilation barrier for the other containment penetrations during CORE ALTERATIONS or movement of irradiated fuel.

The containment purge and exhaust isolation system must also be OPERABLE during CORE ALTERATIONS or movement of irradiated fuel when open to the outside atmosphere or must be closed under LCO 3.9.4.c.1. An OPERABLE containment purge and exhaust isolation system consists of a containment purge valve capable of isolating on an actual or simulated actuation signal from containment purge isolation from each of the required radiation monitoring instrumentation channels (Note that Technical Specifications 3/4.3.3, Radiation Monitoring is also applicable). The containment purge lines are automatically closed upon a containment purge isolation signal (CPIS) if the fuel handling accident releases activity above prescribed levels. Closure of at least one of the containment purge isolation valves is sufficient to provide closure of the penetration.

Administrative controls shall ensure that appropriate personnel are aware that when the equipment door, both personnel airlock doors, and/or containment penetrations are open, a specific individual(s) is designated and available to close the equipment door, an airlock door and the penetrations as part of a required evacuation of containment, and any obstruction(s)

(e.g., cables and hoses) that could prevent closure of an airlock door and the equipment door be capable of being quickly removed.

<(DRN 03-178, Ch. 21; EC-28875, Ch. 69) 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

AMENDMENT NO. 144, 148, WATERFORD - UNIT 3 B 3/4 9-2 CHANGE NO. 19, 21, 62, 69

REFUELING OPERATIONS BASES 3/4.9.6 REFUELING MACHINE

>(EC-17724, Ch. 62)

The OPERABILITY requirements for the refueling machine ensure that: (1) the refueling machine will be used for movement of CEAs and fuel assemblies, (2) each hoist has sufficient load capacity to lift a CEA or fuel assembly, and (3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations. The Technical Specification Actions a. and b. statements allow the movement of a fuel assembly or CEA to safe condition using administrative controls in the event of a refueling machine failure.

<(EC-17724, Ch. 62) 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly, CEA, and associated handling tool over other irradiated fuel assemblies in the Fuel Handling Building ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION

>(DRN 03-375, Ch. 19)

The requirement that at least one shutdown cooling train be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. If SDC loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations.

Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that which would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operations.

<(DRN 03-375, Ch. 19)

The requirement to have two shutdown cooling trains OPERABLE when there is less than 23 feet of water above the top of the fuel seated in the reactor pressure vessel ensures that a single failure of the operating shutdown cooling train will not result in a complete loss of decay heat removal capability. When there is no irradiated fuel in the reactor pressure vessel, this is not a consideration and only one shutdown cooling train is required to be OPERABLE. With the reactor vessel head removed and 23 feet of water above the top of the fuel seated in the reactor pressure vessel, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling train, adequate time is provided to initiate emergency procedures to cool the core.

WATERFORD - UNIT 3 B 3/4 9-3 CHANGE NO. 19, 21, 69

REFUELING OPERATIONS BASES

>(DRN 03-233, Ch. 22; EC-28875, Am. 69)

<(DRN 03-233, Ch. 22; EC-28875, Am. 69) 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and SPENT FUEL POOL

>(DRN 05-131, Ch. 39)

The restrictions on minimum water level ensure that sufficient water depth is available such that the iodine released as a result of a rupture of an irradiated fuel assembly is reduced by a factor of at least 200. Gap fractions are assumed in accordance with Regulatory Guide 1.183 guidance. The minimum water depth is consistent with assumptions of the safety analysis.

<(DRN 05-131, Ch. 39)

>(EC-18742, Ch. 65) 3/4.9.12 and 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION and SPENT FUEL STORAGE TS 5.6, FUEL STORAGE, reflects the results of the criticality analysis, crediting soluble boron and allowing more flexibility in storing the more reactive Next Generation Fuel (NGF) assemblies in the spent fuel storage racks. The Waterford 3 SFP criticality analysis used a design acceptance criteria of effective (neutron) multiplication factor (keff) no greater than 0.995, if flooded with unborated water, and keff no greater than 0.945, if flooded with borated water.

This provides an additional 0.005 keff analytical margin to the regulatory requirement. This approach provides sufficient margin to offset minor non-conservatisms to provide reasonable assurance that the regulatory requirements are met. Each storage configuration has a geometric arrangement which must be maintained so that the SFP criticality analysis remains valid.

The spent fuel pool (SFP) criticality analysis credits 524 parts per million (ppm) of soluble boron to maintain keff less than 0.95 in the SFP during normal conditions, and 870 ppm under the worst-case accident conditions. The analysis determined that a misloading event in the spent fuel checkerboard loading pattern would have the largest reactivity increase, requiring 870 ppm of soluble boron to meet the regulation. The boron dilution analysis identified a number of assorted sources for slow addition of unborated water to the SFP that could possibly continue undetected for an extended period of time. The maximum flow from any of these sources was determined to be 2 gpm, and dilution of the SFP from 1900 ppm to 870 ppm soluble boron would take approximately 72 days. Slow dilution by undetected sources is adequately addressed by sampling the SFP on the 7-day frequency of SR 4.9.12. Higher flow-rate dilution scenarios would be identified through various alarms and building walkdowns, and

<(EC-18742, Ch. 65)

WATERFORD - UNIT 3 B 3/4 9-4 CHANGE NO. 19, 21, 22, 39, 65, 69

>(EC-18742, Ch. 65)

REFUELING OPERATIONS BASES 3/4.9.12 and 3/4.9.13 SPENT FUEL POOL BORON CONCENTRATION and SPENT FUEL STORAGE (Continued) could be addressed by sampling the SFP on the 7-day frequency of SR 4.9.12. Higher flow-rate dilution scenarios would be identified through various alarms and building walkdowns, and could be addressed within a sufficient time to preclude dilution of the SFP to 870 ppm soluble room.

Adequate safety is maintained in the case of a high flow-rate dilution of the SFP in accordance with 10 CFR 50.68(b)(4) because keff must remain below 1.0 (subcritical), even if the SFP were flooded with unborated water.

Three qualified storage configurations are allowed for Region 2 Fuel Storage locations, based on burnup versus enrichment restrictions: 1) uniform loading of assemblies, 2) checkerboard loading of high and low reactivity assemblies, and 3) checkerboard loading of fresh assemblies and empty cells. The storage configurations may be interspersed with each other throughout the SFP, provided that the geometric interface requirements are met.

Checkerboard loading is not required for Region I Fuel Storage locations.

<(EC-18742, Ch. 65)

WATERFORD - UNIT 3 B 3/4 9-5 CHANGE NO. 65, 69

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