ML060650507

Changes to Technical Specifications Bases

ML060650507
Person / Time
Site:	Millstone
Issue date:	03/03/2006
From:	Hartz L Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
06-018
Download: ML060650507 (23)

Text

Dominion Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, Virginia 23060 Web Address: www.dom.com March 3, 2006 U.S. Nuclear Regulatory Commission Serial No.06-018 Attention: Document Control Desk MPS LiclMAE RO Washington, DC 20555 Docket Nos. 50-336 50-423 License Nos. DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNITS 2 AND 3 CHANGES TO TECHNICAL SPECIFICATIONS BASES In accordance with the requirements of Technical Specification 6.23.d of Millstone Power Station Unit 2 (MPS2) and Technical Specification 6.18.d of Millstone Power Station Unit 3 (MPS3), Dominion Nuclear Connecticut, Inc. (DNC) is providing the Nuclear Regulatory Commission Staff with changes to MPS2 and MPS3 Technical Specifications Bases Sections. MPS2 changes affect Technical Specifications Bases Section 3/4.7. MPS3 changes affect Technical Specifications Bases Sections 3/4.1, 3/4.3, 314.4, 3/45, 314.6 and 3/4.9. These changes are provided for information only.

The changes to the Bases Sections were made in accordance with the provisions of 10 CFR 50.59. These changes have been reviewed and approved by the Site Operations Review Committee.

Attachments 1 and 2 provide the retyped pages of the Technical Specifications Bases for MPS2 and MPS3 respectively.

There are no regulatory commitments contained within this letter.

If you have any questions or require additional information, please contact Mr. Paul R.

Willoughby at (804) 273-3572.

Very truly yours, Leslie N. Hartz Vice President - Nuclear Engineering

Serial No. 06-01 8 Docket Nos. 50-336150-423 Changes To Technical Specifications Bases Page 2 of 2 Attachments:

1. Re-typed Bases Pages for Millstone Unit 2

2. Re-typed Bases Pages for Millstone Unit 3 Commitments made in this letter: None.

cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-141 5 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station

Serial No.06-018 Docket No. 50-336 ATTACHMENT 1 CHANGES TO TECHNICAL SPECIFICATIONS BASES RETYPED PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No.06-018 Docket No. 50-336 Changes to Technical Specifications Bases Millstone Unit 2 Bases Paaes Section No. Page No.

314.7 Plant Svstems B 314 7-4

LBDCR 05-MP2-002 May 19,2005 PLANT SYSTEMS BASES 314.7.4 SERVICE WATER SYSTEM (Continued) determined to be inoperable should be the loop that results in the most adverse plant configuration with respect to the availability of accident mitigation equipment. Restoration of loop independence within the time constraints of the allowed outage time is required, or a plant shutdown is necessary.

It is acceptable to operate with the SW header supply valves to sodium hypochlorite (2-SW-84A and 2-SW-84B) and the SW header supply valves to the north and south filters(2-SW-298 and 2-SW-299) open.The flow restricting orifices in these lines ensure that safety related loads continue to receive minimum required flow during a LOCA (in which the lines remain intact) or during a seismic event (when the lines break). Therefore, operation with these valves open does not affect OPERABILITY of the SW loops.

Surveillance Requirement 4.7.4.1 .a verifies the correct alignment for manual, power operated, and automatic valves in the Service Water (SW) System flow paths to provide assurance that the proper flow paths exist for SW operation. This surveillance does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve automatically repositions within the proper stroke time. This surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position.

The 3 1 day frequency is appropriate because the valves are operated under procedural control and an improper valve position would only affect a single train. This frequency has been shown to be acceptable through operating experience.

Surveillance Requirements 4.7.4.1 .b and 4.7.4.1 .c demonstrate that each automatic SW valve actuates to therequired position on an actual or simulated actuation signal and that each SW pump starts on receipt of an actual or simulated actuation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The 18 month frequency is based on the need to perform these surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the surveillances were performed with the reactor at power. The 18 month frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic is tested as part of the Engineered Safety Feature Actuation System (ESFAS) testing, and equipment performance is monitored as part of the Inservice Testing Program.

314.7.5 DELETED 314.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the Control Room Emergency Ventilation System ensures that

1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

MILLSTONE - UNIT 2 B 314 7-4 Amendment No. 223,246, W, 273,

Serial No.06-018 Docket No. 50-423 ATTACHMENT 2 CHANGES TO TECHNICAL SPECIFICATIONS BASES RETYPED PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No.06-018 Docket No. 50-423 Changes to Technical Specifications Bases Millstone Unit 3 Bases Paaes Section No. Page No.

314.1 Reactivity Control Systems B 314 1-5 314.3 Instrumentation B 314 3-6 314.4 Reactor Coolant System B 314 4-1c, 4-1dl 4-1 el 4-7 314.5 Emergency Core Cooling System B 314 5-2,5-2a, 5-2b, 5-2c 314.6 Containment Systems B 314 6-2a, 6-3a, 6-3b, 6-3c, 6-3d 314.9 Refueling Operations B 314 9-2a

LBDCR 04-MP3-001 ACTIVITY CONTROJI SYSTEM 2/13/04 BASES MOVABLE CONTROL ASSEMBLIES (Continued) then compare the Demand Position Indication System and the DRPI System at least once each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The Rod Deviation Monitor is generated only Erom the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Technical Specification SR 4.1.3.2.1 determines each digital rod position indicator to be OPERABLE by verifying the Demand Position Indication System and the DRPI System agree within 12 steps at least once each 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except during the time when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the D W I System at least once each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

The Rod Deviation Monitor is generated only from the DRPI panel at MB4. Therefore, when rod position indication as displayed by the plant process computer is the only available indication, then perform SURVEILLANCE REQUIREMENTS every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Additional surveillance is required to ensure the plant process computer indications are in agreement with those displayed on the DRPI. This additional SURVEILLANCE REQUIREMENT is as follows:

Each rod position indication as displayed by the plant process computer shall be determined to be OPERABLE by verifying the rod position indication as displayed on the DRPI display panel agrees with the rod position indication as displayed by the plant process computer at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The rod position indication, as displayed by DRPI display panel (MB4),is a non-QA system, calibrated on a refueling interval, and used to implement T/S 3.1.3.2. Because the plant process computer receives field data from the same source as the DRPI System (MB4), and is also calibrated on a reheling interval, it fully meets all requirements specified in T/S 3.1.3.2 for rod position.

Additionally, the plant process'computerprovides the same type and level of accuracy as the DRPI System (MB4).The plant process computer does not provide any alarm or rod position deviation monitoring as does DRPI display panel (MB4).

MILLSTONE - UNIT 3 Amendment No. 68,

LBDCR 05-MP3-028 November 30,2005 INSTRUMENTATION BASES 314.3.3.6ACCIDENT MONITORING INSTRUMENTATION (Continued)

A channel is OPERABLE if four or more sensors, half or more in the upper head region and half or more in the upper plenum region, are OPERABLE.

In the event more than four sensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during the next refueling outage. This is because the sensors are accessible only after the missile shield and reactor vessel head are removed. It is not feasible to repair a channel except dunng a refueling outage when the missile shield and reactor vessel head are removed to refuel the core. If only one channel is inoperable, it should be restored to OPERABLE status in a refueling outage as soon as reasonably possible. If both channels are inoperable, at least one channel shall be restored to OPERABLE status in the nearest refueling outage.

The Reactor Coolant System Subcooling Margin Monitor, Core Exit Thermocouples, and Reactor Vessel Water Level instruments are processed by two separate trains of ICC (Inadequate Core Cooling) and HJTC (Heated Junction ThennoCouple) processors. The preferred indication for these parameters is the Safety Parameter Display System (SPDS) via the non-qualified PPC (Plant Process Computer) but qualified indication is provided in the instrument rack room. When the PPC data links cease to transmit data, the processors must be reset in order to restore the flow of data to the PPC. During reset, the qualified indcation in the instrument rack room is lost.

These instruments are OPERABLE during this reset since the indication is only briefly interrupted while the processors reset and the indication is promptly restored. The sensors are not removed from service dunng this reset. The train should be considered inoperable only if the qualified indication fails to be restored following reset. Except for the non-qualified PPC display, the instruments operate as required 314.3.3.7 DELETED 314.3.3.8 DELETED 314.3.3.9 DELETED 3/4.3.3.10 DELETED 314.3-4 DELETED MILLSTONE - UNIT 3 Amendment No. 4-88, -143,-U9,

LBDCR NO.05-MP3-026 October 19,2005 3/4.4 REACTOR COOLANT SYSTEM BASES (Continued)

Specification 3.4.1.6 The requirement to maintain the isolated loop stop valves shut with power removed ensures that no reactivity addition to the core could occur due to the startup of an isolated loop.

Verification of the boron concentration in an isolated loop prior to opening the first stop valve provides a reassurance of the adequacy of the boron concentration in the isolated loop.

RCS L o o ~ FilledlNot s Filled:

In MODE 5, any RHR train with only one cold leg injection path is sufficient to provide adequate core cooling and prevent stratification of boron in the Reactor Coolant System.

The definition of OPERABILITY states that the system or subsystem must be capable of performing its specified function(s). The reason for the operation of one reactor coolant pump (RCP) or one RHR pump is to:

Provide sufficient decay heat removal capability Provide adequate flow to ensure mixing to:

Prevent stratification Produce gradual reactivity changes due to boron concentration changes in the RCS The definition of "Reactor coolant loops filled" includes a loop that is filled, swept, and vented, and capable of supporting natural circulation heat transfer. This allows the non-operating RHR loop to be removed from service while filling and unisolating loops as long as steam generators on the OPERABLE reactor coolant loops are available to support decay heat removal.

Any loop being unisolated is not OPERABLE until the loop has been swept and vented. The process of sweep and vent will make the previously OPERABLE loops inoperable and the requirements of LC0 3.4.1.4.2, "Reactor Coolant System, COLD SHUTDOWN - Loops Not Filled," are applicable. When the RCS has been filled, swept and vented using an approved procedure, all unisolated loops may be declared OPERABLE.

The definition of "Reactor coolant loops filled" also includes a loop that has been vacuum filled and capable of supporting natural circulation heat transfer. Any isolated loop that has been vacuum filled is OPERABLE as soon as the loop is unisolated.

One cold leg injection isolation valve on an RHR train Fay be closed without considering the train to be inoperable, as long as the following conditions exist:

CCP temperature is at or below 95°F Initial RHR temperature is below 184°F MILLSTONE - UNIT 3 B 314 4-IC Amendment No. W,

LBDCR 05-MP3-026 October 19,2005 3/4.4 REACTOR COOLANT SYSTEM BASES (Continued)

The single RHR cold leg injection flow path is goJ utilized until a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown CCP flow is at least 6,600 gprn RHR flow is at least 2,000 gprn In the above system lineup, total flow to the core is decreased compared to the flow when two cold legs are in service. This is acceptable due to the substantial margin between the flow required for cooling and the flow available, even through a slightly restricted RHR train.

The review concerning boron stratification with the utilization of the single injection point line, indicates there will not be a significant change in the flow rate or distribution through the core, so there is not an increased concern due to stratification.

Flow velocity, which is high, is not a concern from a flow erosion or pipe loading standpoint. There are no loads imposed on the piping system which would exceed those experienced in a seismic event. The temperature of the fluid is low and is not significant from a flow erosion standpoint.

The boron dilution accident analysis, for Millstone Unit 3 in MODE 5, assumes a full RHR System flow of approximately 4,000 gpm. Westinghouse analysis, Reference (I), for RHR flows down to 1,000 gpm, determined adequate mixing results. As the configuration will result in a RHR flow rate only slightly less then 4,000 gprn there is no concern in regards to a boron dilution accident.

The basis for the requirement of two RCS loops OPERABLE is to provide natural circulation heat sink in the event the operating RHR loop is lost. If the RHR loop were lost, with two loops filled and two loops air bound, natural circulation would be established in the two filled ]

loops.

Natural circulation would not be established in the air bound loops. Since there would be no circulation in the air bound loops, there would be no mechanism for the air in those loops to be carried to the vessel, and subsequently into the filled loops rendering them inoperable for heat I sink requirements.

The LC0 is met as long as at least two reactor coolant loops are OPERABLE and the following conditions are satisfied:

One RHR loop is OPERABLE and in operation, with exceptions as allowed in Technical Specifications; and MILLSTONE - UNIT 3 Amendment No. W ,

LBDCR O4-MP3.-OO1 4

314.4 REACTOR COOLANT SYSTEM BASES (Continued)

Either of the following:

An additional RHR loop OPERABLE, with exceptions as allowed in Technical Specifications; or The secondary side water level of at least two steam generators shall be greater than 17% (These are assumed to be on OPERABLE reactor coolant loops)

When the reactor coolant loops are swept, the mechanism exists for air to be carried into previously OPERABLE loops. All previously OPERABLE loops are declared inoperable and an additional RHR loop is required OPERABLE as specified by LC0 3.4.1.4.2 for Loops not filled.

When the RCS has been filled, swept, and vented using an approved procedure, all unisolated loops may be declared OPERABLE.

ISOLATED LOOP STARTUP The below requirements are for unisolating a loop with all four loops isolated while decay heat is being removed by RHR and to clarify prerequisites to meet TIS requirements for unisolating a loop at any time.

With no RCS loops operating, the two RHR loops referenced in Specification 3.4.1.4.2 are the operating loops. Starting in MODE 4 as referenced in Specification 3.4.1.3, the RHR loops are allowed to be used in place of an operating RCS loop. Specification 3.4.1.4.2 requires two RHR loops OPERABLE and at least one in operation. Ensuring the isolated cold leg temperature is within 20°F of the highest RHR outlet temperature for the operating RHR loops within 30 minutes prior to opening the cold leg stop valve is a conservative approach since the major concern is a positive reactivity addition.

SR 4.4.1.6.1 : When in MODE 5 with all RCS loops isolated, the two RHR loops referenced in LC0 3.4.1.4.2 shall be considered the OPERABLE RCS loops. The isolated loop cold leg temperature shall be determined to be within 20°F of the highest RHR outlet temperature for the operating RHR loops within 30 minutes prior to opening the cold leg stop valve.

Surveillance requirement 4.4.1.6.2 is met when the following actions occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to opening the cold leg or hot leg stop valve:

An RCS boron sample has been taken and analyzed to determine current boron concentration The SHUTDOWN MARGIN has been determined using OP 3209B, "Shutdown Margin" using the current boron concentration determined above 1 MILLSTONE - UNIT 3 B 314 4-le Amendment No. W,

LBDCR NO. 05-MP3-001 July 28,2005 REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

Reducing TW to less than 500°F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

314.4.9 PRESSURElTEMPERATURE LIMITS REACTOR COOLANT SYSTEM (EXCEIYT THE PRESSURIZER)

BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LC0 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

Figures 3.4-2 and 3.4-3 contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational requirements during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. A heatup or cooldown is defined as a temperature increase or decrease of greater than or equal to 1O°F in any one hour period. This definition of heatup and cooldown is based upon the ASME definition of isothermal conditions described in ASME,Section XI, Appendix E.

MILLSTONE - UNIT 3 Amendment No. -153,497,

LBDCR NO. 05-MP3-004 April 2 1,2005 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)

The Charging PumpIReactor Plant Component Cooling Water Pump Ventilation System is required to be available to support charging pump operation. The Charging PurnpReactor Plant Component Cooling Water Pump Ventilation System consists of two redundant trains, each capable of providing 100% of the required flow. Each train has a two position, "Off' and "Auto,"

remote control switch, With the remote control switches for each train in the "Auto" position, the system is capable of automatically transferring operation to the redundant train in the event of a low flow condition in the operating train. The associated fans do not receive any safety related automatic start signals (e.g., Safety Injection Signal).

Placing the remote control switch for a Charging PumpReactor Plant Component Cooling Water Pump Ventilation Train in the "OR' position to start the redundant train or to perform post maintenance testing to verifj. avajlability of the redundant train will not affect the availability of that train, provided appropriate administrative controls have been established to ensure the remote control switch is immediately returned to the "Auto" position after the completion of the specified activities or in response to plant conditions. These administrative controls include the use of an approved procedure and a designated individual at the control switch for the respective Charging PumpfReactor Plant Component Cooling Water Pump Ventilation Train who can rapidly respond to instructions fiom procedures, or control room personnel, based on plant conditions.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from e x c 4 n g mnout conctitions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

Surveillance Requirement 4.5.2.b.1 requires verifying that the ECCS piping is full of water. The ECCS pumps are normally in a standby, nonoperating mode, with the exception of the operating centrifugal charging pump(s). As such, the ECCS flow path piping has the potential to develop voids and pockets of entrained gases. Maintaining the piping fi-om the ECCS pumps to the RCS full of water ensures that the system will perform properly when required to inject into the RCS. This will also prevent water hammer, degraded performance, cavitation, and gas binding of ECCS pumps, and reduce to the greatest extent practical the pumping of non-condensible gases (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an SI signal I

or during shutdown cooling.

This Surveillance Requirement is met by:

VENTING the ECCS pump casings and VENTING or Ultrasonic Test (UT) of the accessible suction and discharge piping high points including the ECCS pump suction crossover piping (i.e., downstream of valves 3RSS*MV8837A/B and I

3RSS*MV8838AIB to safety injection and charging pump suction). VENTING of the 1 MILLSTONE - UNIT 3 B 3/4 5-2 Amendment No. 1-88,+#, 4-57,

LBDCR NO.05-MP3-004 April 2 1, 2005 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) accessible suction and discharge piping high points including the ECCS pump suction crossover piping is required when gas accumulations exceed the gas accumulation limits. NOTE: Certain maintenance (e.g. ECCS pump overhaul) or other evolutions can cause gas or air to enter the ECCS. VENTING of the affected portion of the ECCS is necessary for these evolutions, VENTING of the nonoperating centrifugal charging pumps at the suction line test connection. The nonoperating centrifugal charging pumps do not have casing vent connections and VENTING the suction pipe will assure that the pump casing does not contain voids and pockets of entrained gases.

using an external water level detection method for the water filled portions of the RSS piping upstream of valves 3RSS *MV8837A/B and 3RSS*MV8838A/B. When deemed necessary by an external water level detection method, filling and venting to reestablish the acceptable water levels may be performed after entering LC0 ACTION statement 3.6.2.2 since VENTING without isolation of the affected train would result in a breach of the containment pressure boundary.

The following ECCS subsections are exempt Erom this Surveillance:

the operating centrifugal charging pump(s) and associated piping - as an operating pump is self VENTING and cannot develop voids and pockets of entrained gases.

the RSS pumps, since this equipment is partially dewatered during plant operation.

Each RSS pump is equipped with a pump casing vent line that allows automatic VENTING of the pump casing prior to pump operation following an accident.

the RSS heat exchangers, since this equipment is laid-up diy during plant operation.

Gas is flushed out of the heat exchangers during the initial operation of the RSS pumps following an accident.

the RSS piping that is not maintained filled with water during plant operation. The configuration of this piping is such that it is self VENTING upon initial operation of the RS S pumps.

the ECCS discharge piping within containment. These piping sections are inaccessible during reactor operations due to accessibility (containment entry), safety, and radiological concerns. They are static sections of piping relatively insensitive to gas accuinulations since these lines are stagnant during normal power operation. The ECCS discharge piping inside containment is filled and vented upon system return to sewice.

the Residual Heat Removal (RHR) heat exchangers. These are dual pass, vertical u-tube heat exchangers that do not allow direct measurement of gas voids. System MILLSTONE - UNIT 3 Amendlnent No. W, W, 45.7,

LBDCR NO. 05-MP3-004 April 2 1,2005 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) flush upon heat exchanger return to service and procedural compliance is relied upon to ensure that gas is not present within the heat exchanger u-tubes. I Surveillance Requirement 4.5.2.C.2 requires that the visual inspection of the containment be performed at least once daily if the containment has been entered that day and when the final containment entry is made. This will reduce the number of unnecessary inspections and also reduce personnel exposure.

The Emergency Core Cooling System (ECCS) has several piping cross connection points for use during the post-LOCA recirculation phase of operation. These cross-connection points allow the Recirculation Spray System (RSS) to supply water from the containment sump to the safety injection and charging pumps. The RSS has the capability to supply both Train A and B safety injection pumps and both Train A and B charging pumps. Operator action is required to position valves to establish flow from the containment sump through the RSS subsystems to the safety injection and charging pumps since the valves are not automatically repositioned. The quarterly stroke testing (Technical Specification 4.0.5) of the ECCRSS recirculation flowpath valves discussed below will not result in subsystem inoperability (except due to other equipment manipulations to support valve testing) since these valves are manually aligned in accordance with the Emergency Operating Procedures (EOPs to establish the recirculation flowpaths. It is 1

expected the valves will be returned to the norma pre-test position following termination of the surveillance testing in response to the accident. Failure to restore any valve to the normal pre-test position will be indicated to the Control Room Operators when the ESF status panels are checked, as directed by the EOPs. The EOPs direct the Control Room Operators to check the ESF status panels early in the event to ensure proper equipment alignment. Sufficient time before the recirculation flowpath is required is expected to be available for operator action to position any valves that have not been restored to the pretest position, including local manual valve operation.

Even if the valves are not restored to the pre-test position, sufficient capability will remain to meet ECCS ost-LOCA recirculation requirements. As a result, stroke testing of the ECCS f

recircu ation valves discussed below will not result in a loss of system independence or redundancy, and both ECCS subsystems will remain OPERABLE.

When performing the quarterly stroke test of 3SIH*MV8923A, the control switch for safety injection pump 3SIH*PlA is placed in the pull-to-lock position to prevent an automatic pump start with the suction valve closed. With the control switch for 3SIH*PlA in pull-to-lock, the Train A ECCS subsystem is inoperable and Technical Specification 3.5.2, ACTION a.,

applies. This ACTION statement is sufficient to administratively control the plant configuration with the automatic start of 3SIH*PlA defeated to allow stroke testing of 3SIH*MV8923A. In addition, the EOPs and the ESF status panels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath. Even if system restoration is not accomplished, sufficient equipment will be available to perform all ECCS and RSS injection and recirculation functions, provided no additional ECCS or RSS equipment is inoperable, and an additional single failure does not occur (an acceptable assumption since the Technical Specification ACTION statement limits the plant configuration time such that no additional equipment failure need be postulated). During the injection phase the redundant subsystem (Train B) is fully functional, as is a significant portion of the Train A subsystem. During the recirculation phase, the Train A RSS subsyste~ncan supply water from the coiltaiilrnent sump to the Train A MILLSTONE - UNIT 3 B 314 5-2b Amendment No. W, 44-7,4-57,

LBDCR NO. 05-MP3-004 April 21,2005 EMERGENCY CORE COOLING SYSTEMS BASES and B charging pumps, and the Train B RSS subsystem can supply water fiom the containment sump to the B safety injection pump.

When performing the quarterly stroke test of 3SIH*MV8923B, the control switch for safety injection pump 3SIH*PlB is placed in the pull-to-lock position to prevent an automatic pump start with the suction valve closed. With the control switch for 3SIH*PlB in pull-to-lock, the Train B ECCS subsystem is inoperable and Technical Specification 3.5.2, ACTION a,,

applies. This ACTION statement is sufficient to administratively control the plant configuration with the automatic start of 3SIH*PlB defeated to allow stroke testing of 3SIH*MV8923B. In addition, the EOPs and the ESF status panels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath. Even if system restoratioh is not accomplished, suffkient equipment will be available to perform all ECCS and RS8 injection and recirculation functions, pavidd no additional ECCS or RSS equipment is iaopmabb, and an additional single failure does not occur (an acceptable assumption since the Technical Specification ACTION statement limits the plant configuration time such that no additional equipment failure need be postulated). During the injection phase the redundant subsystem (Train-A) is fully functional, as is a significant portion of the Train B subsystem. During the recirculation phase, the Train A RSS subsystem can supply water from the containment sump to the Train A and B charging pumps and the Train A safety injection pump. The Train B RSS subsystem cannot supply water fiom the containment sump to any of the remaining pumps.

When performing the quarterly stroke test of 3SIH*MV8807A or 3SIH*MV8807B, 3SIH*MV8924 is closed first to prevent the potential injection of RWST water into the RCS through the operating charging pump. When 3SIH*MV8924 is closed, it is not necessary to declare either ECCS subsystem inoperable. Although expected to be open for post-LOCA recirculation, sufficient time is expected to be available post-LOCA to identi@ and open 3SIH*MV8924 either from the Control Room or locally at valve. The Ems and tfre ESP status pafiels will identify this abnormal plant configuration, if not corrected following the termination of the surveillance testing, to the plant operators to allow restoration of the normal post-LOCA recirculation flowpath. Even if system restoration is not accomplished, sufficient equipment will be available to perform all ECCS and RSS injection and recirculation functions, provided no additional ECCS or RSS equipment is inoperable, even if a single failure is postulated. The failure to open 3SIH*MV8924 due to mechanical binding or the loss of power to ECCS Train A could be the single failure. If a different single failure is postulated, restoration of 3SIH*MV8924 can be accomplished. The closure of 3SIH*MV8924 has no affect on the injection phase. During the recirculation phase, assuming 3SIH*MV8924 remains closed (i.e., the single failure), the Train A RSS subsystem can supply water fiom the containment sump to the Train A and B charging pumps, and the Train B RSS subsystem can supply water from the containment sump to the Train A and B safety injection pumps. If power is lost to ECCS Train A and 3SIH*MV8924 is not opened locally (i.e., the single failure ), cold leg recirculation can be accomplished by using RSS Train B to supply containment sump water via 3SIH*PlB to the RCS cold legs and 3SIL*MV8809B can be opened to supply containment sump water via RSS Train B to the RCS cold legs. Hot leg recirculation can be accomplished by using RSS Train B to supply containment sump water via 3SIH*PlB to the RCS hot legs and maintaining 3SIL*MV8809B open to supply containment sump water via RSS Train B to the RCS cold legs.

MILLSTONE - UNIT 3 Amendment No. W ,W , M , 1

LBDCR NO. 04-MP3-007 August 12,2004 CONTAINMENT SYSTEMS BASES The design of the Containment RSS is sufficiently independent so that an active failure in the recirculation spray mode, cold leg recirculation mode, or hot leg recirculation mode of the ECCS has no effect on its ability to perform its engineered safety function. In other words, the failure in one subsystem does not affect the capability of the other subsystem to perform its designated safety function of assuring adequate core cooling in the event of a design basis LOCA.

As long as one subsystem is OPERABLE, with one pump capable of assuring core cooling and the other pump capable of removing heat fiom containment, the RSS system meets its design requirements.

The L C 0 3.6.2.2. ACTION applies when any of the RSS pumps, heat exchangers, or associated components are declared inoperable. All four RSS pumps are required to be OPERABLE to meet the requirements of this LC0 3.6.2.2. During the injection phase of a Loss Of Coolant Accident all four RSS pumps would inject into containment to perform their containment heat removal function. The minimum requirement for the RSS to adequately perform this function is to have at least one subsystem available. Meeting the requirements of L C 0 3.6.2.2. ensures the minimum RSS requirements are satisfied.

Surveillance Requirements 4.6.2.1 .d and 4.6.2.2.e require verification that each spray nozzle is unobstructed following maintenance that could cause nozzle blockage. Normal plant operation and maintenance activities are not expected to trigger performance of these surveillance requirements. However, activities, such as an inadvertent spray actuation that causes fluid flow through the nozzles, a major configuration change, or a loss of foreign material control when working within the respective system boundary may require surveillance performance. An evaluation, based on the specific situation, will determine the appropriate test method (e.g., visual inspection, air or smoke flow test) to verify no nozzle obstruction.

MILLSTONE - UNIT 3 Amendment No.

LBDCR 05-MP3-028 November 30,2005 CONTAINMENT SYSTEMS BASES For the purposes of meeting this LCO, neither the containment isolation valve, nor any alternate valve on a closed system have a leakage limit associated with valve OPERABILITY.

The opening of containment isolation valves on an intermittent basis under administrative controls includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

The appropriate administrative controls, based on the above considerations, to allow containment isolation valves to be opened are contained in the procedures that will be used to operate the valves. Entries should be placed in the Shift Manager Log when these valves are opened or closed. However, it is not necessary to log into any Technical Specification ACTION Statement for these valves, provided the appropriate administrative controls have been established.

Opening a closed containment isolation valve bypasses a plant design feature that prevents the release of radioactivity outside the containment. Therefore, this should not be done frequently, and the time the valve is opened should be minimized. The determination of the appropriate administrative controls for containment isolation valves requires an evaluation of the expected environmental conditions. This evaluation must conclude environmental conditions will not preclude access to close the valve, and this action will prevent the release of radioactivity outside of containment through the respective penetration.

When the Residual Heat Removal (RHR) System is placed in service in the plant cooldown mode of operation, the RHR suction isolation remotely operated valves 3RHS*W8701A and 3RHS*MV8701B, andlor 3RHS*MV8702A and 3RHS*MV8702B are opened. These valves are normally operated from the control room. They do not receive an automatic containment isolation closure signal, but are interlocked to prevent their opening if Reactor Coolant System (RCS) preqsure is greater than approximately 412.5 psia. When any of these valves are opened, either one of the two required licensed (Reactor Operator) control room operators can be credited as the operator required for administrative control. It is not necessary to use a separate dedicated operator.

314.6.4 DELETED MILLSTONE - UNIT 3 Amendment No. %,43,M, 244,

LBDCR 05-MP3-028 November 30,2005 THIS PAGE INTENTIONALLY LEFT BLANK IILLSTONE - UNIT 3 Amendment No. %,a, 44,216,

LBDCR 05-MP3-028 November 30,2005 THIS PAGE INTENTIONALLY LEFT BLANK MILLSTONE - UNIT 3 Amendment No. 63,442, W,

LBDCR 05-MP3-028 November 30,2005 CONTAIMMENT SYSTEMS BASES 314.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM 3l4.6.S.l STEAM JET AIR EJECTOR The closure of the isolation valves in the suction of the steam jet air ejector ensures that:

(1) the containment internal pressure may be maintained within its operation limits by the mechanical vacuum pumps, and (2) the containment atmosphere is isolated from the outside environment in the event of a LOCA. These valves are required to be closed for containment isolation.

MILLSTONE - UNIT 3 Amendment No. 63,442,244,

LBDCR 04-MP3-013 November 29,2004 314.9 REFUELING OPERATIONS BASES 314.9.5 DELETED 314.9.6 DELETED 314.9.7 DELETED 314.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 314.9.8.1 HIGH WATER LEVEL BACKGROUND The purpose of the Residual Heat Removal (RHR) System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant and to prevent boron stratification. Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchanger(s), where the heat is transferred to the Reactor Plant Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR system for normal cooldown or decay heat removal is manually accomplished fiom the control room. The heat removal is manually acco~nplisl~ed from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the RHR heat exchanger(s) and the bypass. Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR system.

MILLSTONE - UNIT 3 Amendment No. W,