B19015, Changes to Technical Specifications Bases
| ML033240469 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/07/2003 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| B19015 | |
| Download: ML033240469 (8) | |
Text
i,;3 DOm-inion, Dominion Nuclear Connecticut, Inc.
Millstone Power Station Rope Ferry Road Waterford. CT 06385 NtV 7 20 Docket No. 50-336 B19015 RE: 10 CFR 50.59 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Power Station, Unit No. 2 Changes 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 (NRC) Staff with changes to Millstone Unit No. 2 Technical Specifications (TS) Bases Sections 2.2.1 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. This letter supersedes the DNC letter dated July 31, 2003.(1) provides the retyped pages of the Technical Specifications Bases for Millstone Unit No. 2.
There are no regulatory commitments contained within this letter.
(1)
J. Alan Price to the Nuclear Regulatory Commission, "Millstone Power Station, Unit No.
2, Changes to Technical Specifications Bases," dated July 31, 2003 (B1 8951).
R-DC)
U.S. Nuclear Regulatory Commission B1 9015/Page 2 If you should have any questions regarding this submittal, please contact Mr. David W.
Dodson at (860) 447-1791, Ext. 2346.
Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.
J. Alpn)Price Site Wce 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 B19015 Millstone Power Station, Unit No. 2 Changes to Technical Specifications Bases Retvped Pages
U.S.
Nuclear Regulatory Commission~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
U.S. Nuclear Regulatory Commission B19015/Attachment 1/Page 1 Millstone Unit No. 2 Bases Pages Section No.
Pa e No.
2.2.1 B 2-6 3/4.7.1.2 B 3/4 7-2, 7-2a, 7-2b
LIMITING SAFETY SYSTEM SETTINGS February 20. 2003 Mambos 2-21-02 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-Hiah 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 Margin/Low Pressure I
The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is below the 95/95 limit for the DNB correlation.
MILLSTONE - UNIT 2 0901 B 2-6 Amendment No. 1g, oil, M, ox, Alp, 7fl
PLANT SYSTEMS TSCR 2-4-02 October18, 2002 BASES 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of the auxiliary feedwiter..pumps ensures that the Reactor Coolant System can be cooled down to less than 3000 r from normal operating conditions in the event of a-total loss of off-site power.
The:FSAR.Chapter 14 Loss. df Norral Feediatdr. (LONF)n.aaiyais.6valuates the *eVent.odcurring with a'nd. without.offtite. power available, ahd -a si§ijle acti've'failure. This analysis.has determined that on'e.m6tor driven AFW. pump.
iso~tsuffici ient to meet the 'acceptance criteria.
Therefore,.two AFW.pumps (two.tr-deiV.n-AFWpumps,...or...ole'imoto r-diven. 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 failurieof onb'e.pump; Thisis-consisteft..w ith.the limiting~onidi'tioni for operation and action'statements of Technical:Specification.3.:7..1.2.
Al~though.not par4t.'of. the b's'eKsof..Tdch~ical.Specif.iatibn 3;7.1.2, the
- less. conseivat.ive FSAR Chapter LO Best.Estimate Analys of the LONF event was performed'. to..demonstfate. that oise.ot r.driven. AF pump is.adequate..to.
rdmove decay heat, pievent steamgenera'tor
.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 singlbemotor-driven AFW pump has sufficient capacity to. reduc'ethe RCS.temperature-to,300 F (in. addition to decy: heat.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 *hjydraulic 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. L
, Zing 0899 g
gi
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 flowj, pressure,.and speed.'(turbine drivhen'pumip only) measurement. instrument jni cc.racies are'idt refle
.1ed in the Technical Specification acceptance criteria. OPER.
The Auxiliary Feed Water- (AFW) system-.is.OPERBLE when the AFWpumps
'and flow paths required to provide AFW to the steam'g nerators.- reOPEABLE.
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. discharge..header requirements, 'and -
common' dischargeheader to the steam generators flow path requirements
- "There' are two':'AFW pump-suction flow.paths from the.:.Condensate.Stdrage
- '. Tank to 'the AFW pumps'.
One flow path to the-turbine drienAFW pump, 'and one'flbw-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 flow 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-besUpplied 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 t1o an actuation signal, or failure -of the piping.
In the case. of an inoperable
- automatic AFW regulating valve a(2-FW-43A.or B),-flow path.OPERABILITY'cah-be 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
^
. ' ^.
Octobert:18, 2002 BASES.
3/4.7.1.2 AUXILIARY FEEDWATER PUMPS (Continued)
AFW imust be capable of. being'deliveered to both Steam-generiators :for.
design basis accident. mitigation;' Certain design basis events, such as. a m.main. steam line break.or steam generator. tube rupture, require that the
- affctted steam.generat'6r be isblatedKdr the RRCS-`decay heatiremeoval' safety fuhction be. 'satisfi~ed.;by;...:feei'ding.and'...ste'am'i':thee'm nnaffected.'ctam
't::
.generatior..
If a faiure ina'-.AFW discharge'flow...path: 'foro'th,'o mmnon discharge header, to a 'steam gehiat pr. prevents.delivery 'of AFW to. a.sttam generator, then the. design basis events 'may nbt be effectively mitigated.
.In thi; s.ituation;'th&:ACTION'requiremenjts. of;Tecjhnical 'Speoei.ficationi.3.0.3 are applicae and an immei~diate plant' shutdown is *appropriate.
'Two..inoperable AFW. System discharge.flow..paths -from. the common
. dischirhge6hfeade-tb both`,Steam !ge'nertrs
.wi ll: eult. in -a co'mpete.loss..of the ability to supply AFW.fl'owvtothe sthnieraVors.
Tnh-this sittation, all three AFW.pumps are inoperable and the ACTION requirements of Technical Sp0etifitat.i.on.. 3.7..2':are appli.cable;,.
Immediat&,
. c6rrecfi'V¢'..'action; is
- required;:/ However, a plant,:shuftdown (is'.not-:app 6pri~ate`.until. a: discha'rg'e
.,' ' flow path. from the -commn.dis1charg header. to one 'steam.generator is restored.
, During quarterly surveillance testing of the turbine driven AFW pump, valveZ2-CN-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'A!
pump is aligne'd 'to the Conddnsate'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 condensatelstorage 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. pg A7, A, 0789 J
7,