ML20073M288

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Proposed Tech Specs Changing Cycle 6 Reload Operating Characteristics
ML20073M288
Person / Time
Site: Millstone Dominion icon.png
Issue date: 04/30/1983
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20073M234 List:
References
TAC-49798, NUDOCS 8304210465
Download: ML20073M288 (14)
v • d • e


Text

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3 Docket No. 50-336 Attachment 3 Millstone Nuclear Power Station, Unit No. 2 Proposed Revisions to Technical Specifications Cycle 6 Reload

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r304210465 830413 PDR ADOCK 05000336 PDR i p

Technical Specification Changes - Cycle 6 Reload P. 2-2 Reduced Reactor Coolant Flow Rate

p. 2-4 Reduced Reactor Coolant Flow Rate
p. 3/4 1-26 CEA Drop Time
p. 3/4 2-4 Axial Shape Index Tent
p. 3/4 2-6 Revised planar radial peaking factor, Fxy.
p. 3/4 2-8 Revised Fxy, Fr limit curves
p. 3/4 2-9 Revised ir.tegrated radial peaking f actor, Fr.
p. 3/4 2-14 Reduced Reactor Coolant Flow Rate
p. 3/4 7-4 Auxiliary Feedwater Pumps

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' TABLE 2.2-1 .

1 REACTOR PROTECTIVE INSTRtMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT _ Trip SETPOINT ALL5WA8LEVALUES

1. Manual Reactor Trfp , Not Applicable Not App 1tcable
2. power Level-Hfgh 4 Four Reactor Coolant pumps Operating < 9.6% above THERMAL POWER.

, wtth a minimum setpoint of < 9.7% above THERMAL POWER. with I

< 14.6% of RATED THERMAL a alntmum of < 14.7% of RATED TOWER, and a maximum of THERMAL POWER 7 and a mantmum of l

< 106.6% of RATED THERMAL i 106.7% of RATED THERMAL POWER.

1 F0WER. -

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Reactor Coolant Flow - Low (1)

e. Four Reactor Coolant. Pumps i a Operating > 91.7% of reactor coolant > 90.15 of reactor coolant flow i

71ow wtth 4 pumps operatirs*. with 4 pumps operating *.

1 4. Reactor Coolant Pump

Speed - Low > 830 rps

> 823 rpm j 5. presserfrer Pressure - Nfgh , < 2400 psia 1 2408 psia

5. Containment Pressure - Hfgh 1 4 75 psig < 5.23 psig
7. Steam Generator pressare - > 500 psia > 492 ps1a low (2) (5) ~ ~

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8. Steam Generator Water

> 36.0% Water' Level - each Level - Low (5) steam generator > 35.25 Water Level - each steam generator

9. Local power Density - High (3) Trip setpoint adjusted to not Trip setpoint adjusted to not exceed the limit lines of exceed the Ilmit lines of

. Figures 2.2-1 and 2.2-2 (4). Figures 2.2-1 and 2.2-2 (4).

1

MA:TIVITY CONTROL SYSTEMS .

CIA DROP TIME 4

L341T!NG CONDITION FOR OPERATION

.s.l.3.4 The individual full length (shutdown and control) CEA drop time, from a fully withdrawn position, shall be 4 2.75 seconds from I when electrical power is interrupted to the CEA :Trive mechanism until the CEA reaches its 90 percent insertion position with:

a. T,yg t 515'F, aM
b. All reactor coolant pumps operating.

APPLICABILITY: MODE 3.

ACTION:

a. With the drop time of any full length CEA detemined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeaing to MODE.1 or 2.
b. With the CEA drop times within limits but detemined at less than full reactor coolant flow, operation may proceed provided THERMAL POWER is restricted to less than or equal to the maximum THERMAL POWER 1evel allowable for the reactor coolant pump combination operating at the time of CEA drop time determination.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality:

a. For all CEAs following each removal of the reactor vessel head,
b. For specifically affected individual CEAs following any inain-tenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs,and
c. At least once per 18 months. ,

MILLSTONE - UNIT 2 3/4 1-2G

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  • FIGURE 3.2 2 AX1AL SHAPE INDEX w Fraction of Allomeble Power Lewt per Specifistion 4.2.1.2c MILLSTONE . UNIT 2 3/4 2-4 ,

i POWER DISTRIBUTION LIMITS TOTAL PLANAR RADIAL PEAKING FACTOR - Fh LIMITING CONDITION FOR OPERATION .

3.2.2Thecalculatedva~1ueofF}y.definedasFfy=Fm(1+T q ). shall be limited to 1 1 615.

l APPLICABILITY: MODE 1*.

ACTION:

With Ffy > 1.615, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

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a. Reduce THERMAL POWER to bring the combination of THERMAL POWER andFfy to within the limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Tern Steady State -

Insertion Limit of Specification 3.1.3.6; or

b. Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 T Ffy shall be calculated by the expression F ,= F,yW ,) and F,Ty shall be detennined to be within its limit at the following intervals:

s. Prior to operation above 70 percent of RATED THERMAL POWER after i each fuel loading,
b. At least once per 31 days of'acetanulated operation in MODE 1 and -
c. Within four hours if the AZIMUTHAL POWER TILT q(T ) is > 0.02. .

5ec Special Test Exception 3.10.2.

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  • f gun E 3.2-3 Total Radlel Pesking Factor Versin Alloweble Frection of RATED THERMAL POWER

F.WER DISTRIBUTION LIMITS T

TOTAL INTEGRATED RADIAL PEAKIIG FACTOR - F p LIMITING CONDIT1DN FDR OPERATION 3.2.3 The calculated value of FT, defined as FT = F (1+Tq ), shall be r 7 limited to 1 1 565. l APPLICABILITY: MODE 1*.

ACTION:

With FT > 1.565, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

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a. Reduce THERMAL POWER to bring the con 61 nation of THERMAL POWER .

T and F, to within the limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Term 5,teady State Insertion Limits of Specification 3.1.3.6; or

b. Be in at least HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 FT7 shall be calculated by the expression F,T = F 7(1+Tq) and Ff shall be determined to be within its limit at the following intervals:

s. Prior to operation above 70 percent of RATED THERMAL POWER after each fuel loading.
b. At least once per 31 days of accumulated operation in NME 1, and
c. Within four hours if the AZIMUTHAL POWER TILT (T ) is > 0.020.

4.2.3.3 F shall be determined each time a calculation of F7 is required r

by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pep Combination.

4.2.3.4 T shall q be determined each time a calculation of F is required -

T and the value of T, used to determine F, shall be the measured value of a .

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TABLE 3.2-1 DNB MARGIN ,

.- LIMITS Four Reactor Coolant Pumps Parameter Operating Cold Leg Temperature 1 549'F Pressurizer Pressure t 2225' psia

  • Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THEbiAL POER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

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PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two feedwater pumps, each capable of being powered from separate OPERABLE emergency busses, and
b. One feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

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a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b. With two auxiliary feedwater pumps inoperable be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. .

i SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE: -

e. At least.once per 31 days by: ,
1. Star. ting each pump from the control room. -
2. Verifying that:

a) Each motor driven pump develops a discharge pressure of >1070 osio on recirculation flow. and

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b) The steam turbine driven pump develops a discharge pressure of > 1080 psig on recirculation flow when ~

the secondary steam supply pressure is greater than 800 psig. The provisions of Specification 4.04 are not applicable for entry into Mode 3.

3/47-4

. e Docket No. 50-336 Attachment 4 Millstone Nuclear Power Station, Unit No. 2 Discussion of Proposed Revisions to Technical Specifications i

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4 April,1983

. . . . _ , - . . _ - , _ , . . _ . . . _ _ _ _ . - - . ~ . _ . - _ - . . _ _ . . , _ _ _ . . _ _ . . . . . . _ . . _ . _ _ _ . - , . _ _ . _ . , . _ . . _ . . . . _ _ . . _ . _ , . . _ .

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Proposed Technical Specification Changes The purpose of this attachment is to provide an explanation for each of the changes proposed in Attachment 3.

o Reduced Reactor Coolant Flow Rate - This proposed change effects pp.

2-2, 2-4, and 3/4 2-14 of the Technical Specifications. It involves lowering the required primary coolant flow rate from 362,500 gpm to 350,000 gpm. This new lower flow is established to correspond to a plugging level of 2500 steam generator tubes.

o CEA Drop Time - This proposed change to p. 3/41-26 of the Technical Specifications involves a revision of the CEA drop time. At the beginning of Cycle 3, four small flow hole test assemblies were put into the core under CEA locations in an effort to mitigate the guide tube wear problem. At that time, the CEA drop time was changed from 2.75 seconds to 3.1 seconds due to a larger dashpot effect realized with the reduced flow holes. This design is no longer being used as the " guide

-tube wear" fix at Millstone Unit 2 and the four test assemblies will be removed from the core during the 1983 refueling. As such, the CEA drop

' time is proposed to be changed back to the original value. The measured value of CEA drop thne determined during startup testing for past refuelings has been well below 2.75 seconds.

o' . New Axial Shape Index Tent - The change to p. 3/4 2-4 involves a new axial shape index (ASI) monitoring tent for figure 3.2-2 of the Technical Specifications. This tent is used to verify the kw/ft limit of 15.6 which is the input to the LOCA analyses. Operation within the tent ensures that the maximum local power is Ic2s tbn 15.6 kw/ft. and thus satsifies the TS surven lance requirement. It should be noted that mder normal conditions . the kw/ft surveillance limit is verified with the incore monitoring system and the only time the ASI tent is used is if the incore system is inoperable, o - Revised tot:1 pianar peaking factor, Fxy, curve - This change effects pp.

3/4 2-6 and 3/4 2-8 of the Technical Specifications and involves restoring the planar radial peaking factor, Fxy, monitoring limits back to the original Beginning of Cycle (BOC) 5 values. The Cycle 6 licensing analyses support this proposed revision.

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( o Revised total radial peaking factor (Fr) curve - This proposed change effects pp. 3/4 2-8 and 3/4 2-9 of the Technical Specifications. In comparing the BOC 5 values to BOC 6 values, the required primary flow is being reduced by 5.4% (370,000 gpm to 350,000 gpm). Although the current licensed primary coolant flow rate is 362,600 gpm, BOC 5 values l are being used since these ' values correspond with those of the last l transient analysis. The Cycle 4 Reload Safety Analyses have shown that the DNB analysis penalty which results from a reduction of 2% in primary flow can be offset with an approximate 1% reduction in Fr.

Therefore, the 4% reduction in allowable Fr more than offsets the penalty associated with a 5.4% reduction in primary flow. The Cycle 6 licensing analyses support this proposed revision.

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o- Auxiliary Feedwater Pumps. - These proposed changes make Millstone Unit 2 Technical Specifications, specifically p. 3/4 7-4, consistent with NUREG-212, Revision 2 Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors. In addition, the proposed revision modifies the Technical Specifications to reflect the actual plant conditions applicable to Mode 4 under which there is insufficient steam to allow the steam turbine driven auxiliary feedwater pump to meet the required discharge pressure.

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