B18951, Changes to Technical Specifications Bases
| ML032250119 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/31/2003 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| B18951 | |
| Download: ML032250119 (10) | |
Text
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-4.IF 0 0 Dominion Dominion Nuclear Connecticut, Inc.
Millstone Power Station Rope Ferry Road Waterford, CT 06385 JL31 mm Docket No. 50-336 B18951 Re: 10 CFR 50.59 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit No. 2 Chanaes to Technical Specifications Bases In accordance with the requirements of Technical Specification 6.23.d of Millstone Unit No. 2, Dominion Nuclear Connecticut, Inc. (DNC) is providing the Nuclear Regulatory Commission Staff with changes to Millstone Unit No. 2 Technical Specifications Bases Sections 2.2.1, 3/4.6.3 and 3/4.7.1.2. 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. provides the retyped pages of the Technical Specifications Bases for Millstone Unit No. 2.
There are no regulatory commitments contained within this letter.
obi
U.S. Nuclear Regulatory Commission B18951/Page 2 If you should have any questions regarding this submittal, please contact Mr. Ravi Joshi at (860) 440-2080.
Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.
J. Al Site President - Millstone Attachment cc:
H. J. Miller, Region I Administrator R. B. Ennis, NRC Senior Project Manager, Unit No. 2 Millstone Senior Resident Inspector
Docket No. 50-336 B18951 Millstone Power Station,-Unit No. 2 Changes to Technical Specifications Bases Retyped Pages
U.S. Nuclear Regulatory Commission B18951/Attachment 1/Page 1 Millstone Unit No. 2 Bases Pages Section No.
Page No.
2.2.1 B 2-6, 2-7 3/4.6.3 B 3/4 6-3b 3/4.7.1.2 B 3/4 7-2, 7-2a, 7-2b
LIMITING SAFETY SYSTEM SETTINGS February 20. 2003 L6bCfl 2-21-02 RA.FIC Steam Generator Water Level -
Low The Steam Generator Water Level-Low Trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded.
Local Power Densitv-Hih-The Local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local power density in the fuel-which corresponds to fuel centerline melting will not occur as a conse-quence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL-SHAPE INDEX exceeds the allowable limits of Figure 2.-2-2.
The AXIAL.
SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level. The trip is automatically bypassed below 15 percent power as sensed by the power range nuclear -instrument Level I bistable.
The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
Thermal Marain/Low Pressure I
The Thermal Margin/Low Pressure trip when the DNBR is below the 95/95 limit for is provided to prevent the DNB correlation.
operation MILLSTONE - UNIT 2 0901 B 2-6 Amendment No. I$,
0X, AZ, OX, XP,$
7179
LIMITING SAFETY SYSTEM SETTINGS Febrmjary_
- 0. 2003 L.FdbrL 2-21 -ok~
BASES Thermal Marain/Low Pressure (Continued)
The trip is initiated whenever the reactor coolant system pressure signal drops below either 1865 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.
In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.
Thermal Margin/Low Pressure trip setpoints are derived from.the core.
safety limits. A safety margin is provided which includes allowances for equipment response times, core power, RCS temperature, and pressurizer pressure measurement uncertainties, processing errors, and a further allowance to compensate for the time delay associated with providing effective termination of the occurrence that exhibits the-most rapid decrease in margin to the safety limit.
Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER as sensed by the power range nuclear instrument Level 1 bistable. This trip provides turbine protection, reduces the severity of the ensuring.transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip.
Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.
I MILLSTONE - UNIT 2 0901 B 2-7 Amendment No. I$, YZ. 11y, jAi, Ml, 3 ZU
LBDCR 2-16-02 CONTAINMENT SYSTEMS December 3, 2002 BASES 3/4.6.3 CONTAINMENT ISOLATION VALVES (continued)
(3) assuring that environmental conditions will not preclude access to close the valve and that this action will prevent the release of radioactivity outside the containment.
The appropriate administrative controls, based on the above considerations, to allow locked or sealed closed 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 and 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.
If a locked or sealed closed containment isolation valve is opened while operating in accordance with Abnormal or Emergency Operating Procedures (AOPs and EOPs), it is not necessary to establish a dedicated operator. The AOPs and EOPs provide sufficient procedural control over the operation of the containment isolation valves.
Opening a locked or sealed 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. As a general guideline, a locked or sealed closed containment isolation valve should not be opened longer than the time allowed to restore the valve to OPERABLE status, as stated in the action statement for LCO 3.6.3.1 "Containment Isolation Valves."
A discussion of the appropriate administrative controls for the containment isolation valves, that are expected to be opened during operation in MODES I through 4, is presented below.
Manual containment isolation valve 2-S1-463, safety injection tank (SIT) recirculation header stop valve, is opened to fill or drain the SITs and for Shutdown Cooling System (SDC) boron equalization.
While 2-SI-463 is open, a dedicated operator, in continuous communication with the control room, is required.
When SDC is initiated, SDC suction isolation remotely operated valves 2-SI-652 and 2-SI-651 (inside containment isolation valve) and manual valve 2-S1-709 (outside containment isolation valve) are opened. 2-S1-651 is normally operated from the control room. While in Modes 1, 2 or 3, 2-SI-651 is closed with manual disconnect switch NS1651 locked open to satisfy Appendix R requirements.It does not receive an automatic containment isolation closure signal, but is interlocked to prevent opening if Reactor Coolant System (RCS) pressure is greater than approximately 275 psia. When 2-SI-651 is opened from the control room, either one of the two required licensed (Reactor Operator) control room operators can be credited as the dedicated operator required for administrative control. It is not necessary to use a separate dedicated operator.
When valve 2-SI-709 is opened locally, a separate dedicated operator is not required to remain at the valve. 2-SI-709 is opened before 2-SI-651. Therefore, opening 2-SI-709 will not establish a connection between the RCS and the SDC System. Opening 2-SI-651 will connect the RCS and SDC System. If a problem then develops, 2-S1-651 can be closed from the control room.
MILLSTONE - UNIT 2 B 3/4 6-3b Amendment No. 2440, 21-5,236
PLANT SYSTEMS TSCR 2-4-02 October 18, 2002 BASES 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of the auxiliary feedwater 'pumps ensures'that the-Reactor Coolant System can be cooled down to less than 3O0o0 from normal operating conditions in the event of a total loss of off-site power.
The FSAR Chapter'14 Loss of Normal Feedwater (LONF) ahalysis evaluates the event occurring with and without offsite. power available, and a single active'failure.
This analysis has determined that one.motor driven AFW pump is not sufficient to meet the acceptance criteria. Therefore', two AFW. pumps (two.motor-drivenAFW:pumps.,, or.
..one' iotor-driven.
AFW.pump' and the steam-driven AFW -pump) are required to meet the acceptance 'criteria for' this moderate frequency event.
To meet the requirement of two AFW pumps available for mitigation,. All three pumps must be. OPERABLE.. to accommodate the fail'ure.of one pump.. This is.
consistent with the limiting:condition. for operation and action'statements of Technical'-Specification--3.7..1.2..
Although.not part of the bases..of Techni-cal. Specification 3.7.1.2, the lessconservative FSAR:Chapter 10 Best Estimate Analysis: of the'LONF 'event was performed'. to dem'onstrite that -one:motor.driven. AF pump is adequate':to..
remove decay heat prevent -steam -generator dryout, maintain Reactor Coolant
- System (RCS) subcooling, and prevent pressurizer level from exceeding acceptable limits.
From this best estimate -analysis of the LONF event, an evaluation.was performed to demonstrate-that a single-motor-driven AFW pump
-has-sufficient capacity to-.reducethe RCS.temperature to,300'-F (-in. addition to decay het.removal) where the Shutdown Cooling System may be placed into operation for continued cooldown.
As a result of these evaluations, one motor-driven AFW pump (or the steam-driven AFW pump which has twice the capacity of a motor-driven AFW pump) can meet the requirements to remove-decay heat, prevent steam generator dryout, maintain RCS subcooling, prevent the pressurizer from exceeding acceptable limits, and reduce RCS temperature to 300'F.
The' Technical Specification Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the accident analysis are met and that subsystem OPERABILITY is maintained.
The purpose of the auxiliary feedwater pumps differential pressure tests on recirculation, Surveillance Requirements 4.7.1.2.a.2.a and 4.7.1.2.a.2.b, is to ensure that the pumps have not degraded to a point where the accident analysis-would be adversely impacted.
The surveillance requirement acceptance criteria -for the motor driven auxiliary feedwater pumps was developed assuming a 5% degraded pump from the actual pump curves.
The surveillance requirement acceptance criteria for the turbine driven auxiliary feedwater pump was developed from high flow test data extrapolated to minimum recirculation flow, and can be adjusted to-account for the affect on pump performance -of variations in pump speed.
Flow and pressure measurement instrument inaccuracies have not been accounted for in the design basis hydraulic analysis for the motor driven auxiliary feedwater pumps. Flow, pressure, and speed-measurement instrument inaccuracies have not been accounted for in the design basis hydraulic analysis for the turbine driven auxiliary feedwater pump.
Corrections for flow, pressure, and speed (turbine driven pump only) measurement instrument inaccuracies will be applied to test data taken when verifying pump MILLSTONE - UNIT 2 B 3/4 7-2 Amendment No. p!,
g, ZJJ, 0899 Zoo Z
PLANT SYSTEMS TSCR 2-4-02 October 18, 2002 BASES 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS (Continued) performance in-the flow. ranges credited... in the accident analyses.
No
.corrections for flow, pressure, and -speed (turbine driven. pump'. only) measurement instrument inaccuracies will be applied to minimum recirculation flow type test data since this portion of the curve is not credited:.in the
. accident analyses....Corrections. for flow,.pressure,.and speed j(turbine driven pump only) measurement. instrument inaccuracies are no treflected in-the Technical Specification acceptance criteria.
The Auxiliary Feed Water -(AFW) system.is
-OPERABLE-when the AFW pumps and flow paths required to provide AFW-to-the steam-generators-are-OPERABLE.
Technical Specification 3.7.1.2 requires three AFW pumps to be OPERABLE and provides ACTIONS to address inoperable AFW pumps.
The AFW flow path
-. ' requirements are -separated into AFW-pump suction flow'path requirements, :AFW pump discharge flow -path to the common disctarge.header -requirements, :and.
common discharge.header to the steam generators 'flow path requirements.
'.. There are twoAFWpump suction flow paths from the'Condensate Storage
'. Tank -to the AFW pumps'.
One flow path.to the -turbine driven AFW pump, and' one flow path to both motor driven AFW pumps.
There are -three..AFW pump
.discharge flow paths to the common discharge header, one flow path from each of the three AFW pumps.
There are two AFW discharge flQw paths from the common discharge header to the steam generators, one flow path to'each-steam generator. With 2-FW-44 open (normal position), the discharge from any AFW pump Will be 'supplied to both steam 'generators through the. associated AFW regulating' valves.
.2-FW-44 should remain open when the AFW system is.
required -to be OPERABLE (MODES 1, 2, and 3).
Closing 2-FW-44 places the plant in a configuration not considered as an initial condition in the Chapter 14.
accident analyses.
Therefore, if 2-FW-44 is closed while the plant is operating in MODES 1, 2, or 3, two AFW pumps should be considered inoperable and the appropriate action requirement of Technical Specification 3.7.1.2 entered to limit.plant operation in this configuration.
A flow path may be considered inoperable as the result of closing a manual valve, failure of -an. automatic valve to respond correctly *to an actuation signal, or failure -of the piping.
In the case. of an'inoperable automatic AFW regulating valve (2-FW-43A or B), flow-path.OPERABILITY!cahnbe restored by use of -a dedicated operator stationed at the associated bypass valve (2-FW-56A or B).as directed by OP 2322..
Failure. of the common discharge header piping will cause both discharge flow paths to the steam generators to be inoperable.
An inoperable suction flow path to the turbine driven AFW pump will result in one inoperable AFW pump. -An inoperable suction flow path to the motor driven AFW pumps will result in two inoperable AFW pumps.- The ACTION requirements of Technical Specification 3.7.1.2 are applicable based on the.
number of inoperable AFW pumps.
An inoperable pump discharge flow path from an AFW pump to the common discharge header will cause the associated AFW pump to be inoperable.
The ACTION requirements of Technical Specification 3.7.1.2 for one AFW pump are applicable for each affected pump discharge flow path.
MILLSTONE - UNIT 2 0899 B 3/4 7-2a Amendment No.
PLANT SYSTEMS TSCR 2-4-02
-October18, 2002 BASES 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS (Continued)
AR! must be capable of being delivered to both steam generators for design basis accident mitigation.
Certain design basis events, such as-'as main. steam line break or steam generator tube rupture, require that the affected'steam.generator be isolated,'and the RCS'decay he-at removal safety function be s'atisfied'.'..by feedinig' and 'stemi'ng the unaffected st6m
.enerator.
If a failure in an`.AfW'd.ischarge'.'flowpath from the omion discharge header to a steam gene-'r~at~orS prevents -delivery of'AF to a steam' generator, then the design basis events may -not be effectively mitigated'.,
In this -s.it.uation; the,:,ACTION requirements of Technical.Specifi.cation.3.0.3 are applicable and an immediate plant shutdown is appropriate;
- Two. inoperable AFW. System discharge -flow..paths from the common distharge header to both:steam' generators 'will result. in a complete. loss: of the ability to supply AFWbflow"tofthe steam generators. In this situation, all th'ree AFW.pumps are inoperable and the ACTION requirements of Technical Specification. 3.7.1.2 are applicable.
Immediate, correcti'vWe-.action.. is'
.e-quired..- Howeve'r, a plant' shitdow6':.i-s not.approprilate
.uhnti.l
-a disc'harge.
flow 'path" from' the common.discharge. header.to one 'steam generator is Arestored.-
During quarterly surveillance testing of the turbine driven AF! pump, valve-2-0N-27A is closed' and valve 2-CN-28 is opened to prevent.overheating
.the water being circulated... In this configuration, the suction of. the turbine driven AFW pump is aligned 'to. the Condensate Storage Tank via the motor driven AFW pump suction flow path, and the pump minimum flow is directed to the Condensate Storage Tank by the turbine driven AFW pump suction path upstream of 2-CN-27A in the reverse direction.
During this surveillance, the suction path to the motor driven AFW pump suction path remains OPERABLE, and the turbine driven AFW suction path-is inoperable. In this situation, the ACTION requirements of Technical Specification 3.7.1.2 for one AFW pump are applicable.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 3007F in the event of a total loss of-off-site power. The minimum water volume is sufficient to maintain the *RCS at HOT STANDBY conditions -for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with steam discharge to atmosphere.
The contained water volume limit includes an allowance formwater not usable due to discharge nozzle pipe elevation above tank bottom, plus an allowance for vortex formation.
3/4.7.1.4 ACTIVITY The limitations on-secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction MILLSTONE - UNIT 2 B 3/4 7-2b Amendment No. pY, fl, Fit
-0789 771', 77,