ML19338C766

Proposed Revisions to Tech Specs Required to Support Plant Operation During Cycle 4 Reload

ML19338C766
Person / Time
Site:	Millstone
Issue date:	08/29/1980
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML19338C765	List: ML19338C764 ML19338C766
References
TAC-11348, TAC-11561, TAC-12505, TAC-42410, TAC-42846, NUDOCS 8009050105
Download: ML19338C766 (28)

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Docket No. 50-336 Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Proposed Revisions to Technical Specifications August,1980 0090go ggy

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• Technical Specification Changes - Cycle 4 Reload P. 1-7 Revised shutdown margin in Mode 5.

P. 2-2 Delete words referring to Figure P 2.1-1.

P. 2-4, 5 Revised LSS Setpoints and Allowable Values.

P. B2-1 W-3 correlation, DNBR of 1.30. Deleted Reference to Figure B 2.1-1.

P. B2-2 Delete Figure B2.1-1.

P. B2-3 DNBR of 1.30.

P. B2-4 Revise Bases 2.2.1 Power Level High to Reflect the Revised Trip Setpoint.

P. 82-5 DNBR of 1.30.

P. B2-6 DNBR of 1.30.

P. B2-8 DNBR of 1.30.

P. 3/4 1-1 Revised Charging Pump Flow Rate.

P. 3/4 1-3 Shutdown Margin, Revised Charging Pump Flow Rate.

P. 3/4 1-5 ModeratorTempegaturegoefficientlessnegative than -2.4 x 10- oK/K F.

P. 3/4 1-24a Delete page.

P. 3/4 2-5 As shown.

P. 3/4 9-1 40 gpm. Revised Charging Pump Flow Rate.

P. 3/4 10-1 40 gpm. Revised Charging Pump Flow Rate.

P. B3/4 1-1 Shutdown Margin.

P. B3/4 1-3 Shutdown Margin.

P. B3/4 7-2 Automatic auxiliary feedwater system.

-s TABLE 1.1_ .

OPERATIONAL MODES REACTIVITY  % P.ATED AVEP. AGE COOLANT CONDITION, K E!ER:'AL P_0' DER

• T EMPEP.ATURE liODE df

1. POWER OPERATION > 0.99 > 5% > 300*F

2. STARTUP > 0.99 < 5% > 300*F 3.' HOT STANDBY < 0.99 0 > 300*F

4. POT SHUTDOWN - < 0.99 0 3nn*F>

> 200*F T"V9

5. 0 < 200*F COLD SHUTDOWN .

<< 0.98 0.99fI2 0 f 200*F

6. REFUELING ** 1 0.90 0 1 140'F

.m qt Excluding decay heat.

Reactor vessel head unbolted or rerr.oved and fuel in the vessel.

(1) When the RCS is fuli or when the RCS is partially drai.ned and the core is in the "ALL RODS IN" configuration.

(2) When the RCS is partially drained.

!!!LLSTONE - UNIT 2 1-7

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Reactor Core Thermal Margin Safety Limit - Four Reactor Coot.mt Pumps Operating FIGURE 2.1 1

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l TABLE 2.2-1 h REACTOR PROTECTIVE INSTRUMEtiTATION TRIP SETP0 INT LDtITS v, ,

13 TRIP SETPOINT ' ALLONABLE VALUES bI FUNCTIONAL UNIT '

Not Applicable not Applicable

[j l. tianual Reactor Trip

% Power Level-High n, 2.

< 9.60Y above THERMAL POWER, < 9.7% above THERMAL POWER, with a Four Reactor Coolant Pumps Opera ting

~ith a minimum setpoint of minimum of _14.7% of RATED TilERMAL POWER, and a Nximum of < 106.7%

< 14.6% of RATED TilERMAL ~

POWER, and a maximum of Gf RATED THERMAL POWER.

< 106.6% of RATED THERMAL POWER. i

3. Reactor Coolant Flow - Low (1) y 91.7% of reactor coolant > 90.1% of reactor coolant flow Four Reactor Coolant Pumps with 4 pumps operating *.

n, flow with 4 pumps operating *.

A Opera ting

4. Reactor Coolant Pump > 823 rpm Speed - Low > 830 rpm 1 2400 psia 1 2408 psia

5. Pressurizer Pressure - High 1 5 23 psig

6. Containnent Pressure - liigh 1 4.75 psig

7. Steam Generator Pressure - > 492 psia

> 500 psia Low (2) (S)

> 36.0% Wa ter Level - each > 35.2% Water Level - each steam

8. Steam Generator Water generator level - Low (5) steam generator

• Design Reactor Coolant flow with 4 pumps operating is 370,000 gpm.

TABLE 2.2-1 (Continued) .

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS _

L';

O TRIP SETPOIN_T_

ALLOWABLE VALUES FUNCTIONAL UNIT Trip setpoint adjusted to Trip setroint adjusted to not Local Power Density - High (3) (6) exceed the limit lines of

[] 9.

not (6) exceed the limit 1 lines of Figures 2.2-1 and Figures 2.2-1 and 2.2-2.

n, 2.2-2,

10. Thermal Margin / Low Pressure (1)

Trip setpoint adjusted to not Four Reactor Coolant Pumps Trip (setpoint (6) exceed the limit lines of Figures Operating not 6) exceed adjustedthe limit to 2.2-3 and 2.2-4.

lines of Figures 2.2-3 and 2.2-4.

11. Loss of Turbine - Hydraulic 2: 500 psig Fluid (3) Pressure - Low 1500 psig 7'.

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TABLE NOTATION (1) Trip may be bypassed below 5% of RATED THERMAL POWER; byp, ass shall be automatically removed when THERMAL POWER is > 5% of RATED THERMAL POWER.

(2) Trip may be manually bypassed below 600 psia; bypass shall be automatically removed j at or above 600 psia, .

t (3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically l

removed when THERMAL POWER is > 15% of RATED THERMAL POWER. I Each of four channels actuate on the auctioneered output of two transmitters, one from (5)

' each steam generator.

(6) Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances.

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- . . . . - - -- =. -~ . - . . . . _ - _ . . -. . . _ - .

May 12, 1979 r

, 2 .1 SAFETY LIMITS BASES

(

REACTOR CORE 2.] .1 be restrictions of this safety limit prevent overheating of the fuel pladding and possible cladding perforation which would result in the i release of fission products to the reactor coolant. Overheating of

the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/ft. Centerline fuel melting will r.ot occur for this peak linear heat rate. Overheating of the fuel cladding is

prevented by restricting fuel operation to within the nucleate boiling l regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB). and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measur-able parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to -

l40 predict the CNB flux and the location of DNB for axially uniform and g

non-untform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would chuse DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, i

normal operational transients, and anticipated transients is limited to ' 10. This value corresponds to a 95 percent probability at a 95 l40 per ent confidence level that DNB will not occur and is chosen as as appropriate margin to DNB for all operating conditions.

' The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant P mps operating for which the minimum DNBR ,

is no less than 1.30 80 I l

- The limits in Figure 2.1-1 were calculated for reactor coolant inlet temperatures less than or equal l to 580'F. The dashed line at 580'F coolant inlet temperatures is not 50 l a safety limit; however, operation above 580'F is not possible because ,

of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at i

THERMAL POWER levels higher than 112% of RATED THERMAL POWER is

! prohibited by the high power level trip set point specified in Table 4n 1 2 . 2-1. The area of safe operation is below and to the left of these lines. ,

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MILLSTONE - UNIT 2 8 2-1

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f MILLSTONE - UNIT 2 B 2-2 l

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May 12, 1o79 SAFETY LIMITS BASES 40 The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.

The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent riny anticipate"i :om-bination of transient conditions for reactor coolant system temperature, pressure, and thermal power level tnat would result in a DNBR of less than 1.30.and preclude the existence of flow instabilities. l40 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

( The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of component design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code require-ments.

The entire Reactor Coolant System is hydrotested at 3125 psia to l demonstrate integrity prior to initial operation.

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MILLSTONE - UNIT 2 B 2-3 1

February 2,1076 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SET POINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at wMch the Reactor Trips are set for each parameter. The Trip Values have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a Trip Setpoint less conservative than its setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed to occur for each trip used in the accident analyses.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automati:

protective instrumentation channels and provides manual reactor trip capability.

Power Level-High The Power Level-High trip provides reactor core protection against-reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin / Low Pressure trip.

The Power Level-Hig,h trip setpoint is operator adjustable and can be set no nigner than 9.im ,s'uove Une leidicated THETd%L FGWER level. Operator action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL power decreases. The trip setpoint has a maximum value of 10~6.6% of REED I _

THERMAL POWER and a minimum setpoint of 1k.6% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 112% of RATED THERMAL POWER, which is the value used in the accident analyses.

Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to permit i

MILLSTONE - UNIT 2 8 2-4 1

May 12, 1970 L,lMITING SAFETY SYSTEM SETTINGS BASES Reactor Co lant Flow-Low (Continued) operation of the reactor at reduced power if one or two reactor coolant ,

pumps are taken out of service. The low-flow trip setpoints and Allowable j Values for the various reactor coolant pump combinations have been  ;

derived in consideration of instrument errors and response times of l equipment involved to maintain the DNBR above 1.30 under normal operation 1 40 j and expected transients. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flow-Low trip set-  ;

points, the Power Level-High trip setpoints, and the Thennal Margin / Low  :

Pressure trip setpoints are automatically changed when the pur.p condition selector switch is manually set to the desired two- or three-pump position.

Changing these trip setpoints during two and three pump operation prevents the minimum value of DNBR from going below 1;30 during normal operational ld0 transients and anticipated transients when only two or three reactor coolant pumps are operating.

Pressurizer Pressure-High

(

The Pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurtzer code safety valves and its concurrent operation with the power-operated relief valves avo;ds the undesirable operation of the pressurizer code safety valves.

Containment Pressure-High The Containnent Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpoint for this trip is identical to the safety injection setpoint.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against :a excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor ct. saat . The setting of 500 psia is sufficiently below the full-load operating point of 815 psia so as not to interfere with nonnal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used win an uncertainty factor of + 22 psi in the accident analyses.

MILLSTO,tE - dNIT 2 B 2-5

,>w-

May 12, 1979 LIMITING SAFETY SYSTEM SETTINGS BASES 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. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide a margin of more than 10 minutes before auxiliary feedwater is required.

Local Power Density-High 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 consequence of axial power ma1 distributions. 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 functior,of THERMAL POWER level with the allowed CEA group position being inferred from the THERMAL POWER level. The trip is automatically bypassed below 15 percent power.

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 opera-tional occurrence prior to a Power, Level-High trip is assumed.

Thermal Margin / Low P' essure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.30.. kn blLLSTONE-UNIT 2 B 2-6

May 12, 1979 LIMITING SAFETY SYSTEM SETTINGS BASES Underspeed - Reactor Coolant Pumps The Underspeed - Reactor Coolant Pumps trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant pump speed (with resulting decrease in flow) on all four reactor 40 coolant pumps. The trip setpoint ensures that a reactor trip will be generated, considering instrument errors and response times, in sufficient time to allow the DNBR to be maintained above 1.30 following a 4 pump loss of flow event.

40 MILLSTONE - UNIT 2 B 2-8

April 19, 1978 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T,yg > 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > 3.20% ak/k.

_ l32 ADPLICABILITY: MODES 1, 2* , 3 and 4.

ACTION:

- With the SHUTDOWN MARGIN < 3.20% ok/k, immediately initiate and continue boration at 2. 40 gpm until the required SHUTDOWN MARGIN is restored. ll32 SURVEILLANCE REQUIREMENTS 4

4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be > 3.20!. ok/k: l32

a. Immediately upon detection of an inoperable CEA. If the inoperable CEA is immovable or untrippable, the SHUTDOWN MARGIN, required by Specification 3.1.1.1, shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA.

b. When in MODES 1 or 2 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying

. that CEA group withdrawal is within the Transie .t Insertion Limits of Specification 3.1.3.6.

c. Prior to initial operation above 5% RATED THERMAL POWER after each refueling, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

See Special Test Exception 3.10.1.

MILLSTONE - UNIT 2 3/4 1-1

., o REACTIVITY CONTROL SYSTEMS avg 6 200"F SHUTDOWN MARGIN - T LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be:

a. > 1.0% A k/k when the reactor coolant system is full.

b. .2.2.0% ok/k or 21% A k/k and in the All Rods in configuration when the sysFem is partially drained.

APPLICABILITY: MODE 5.

~

ACTION:

If the SHUTDOWN MARGIN conditions are not met, immediately initiate and continue boration at 2 40 gpm until the required SHUTDOWN MARGIN is l restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be in accordance with LIMITING CONDITION FOR OPERATION statement 3.1.1.2a or 3.1.1.2b (as applicable):

a. Within one hour after detection of an inoperable CEA. If the inoperable CEA is immovable or untrippable, within one hour

the SHUTDOWN MARGIN required by Specification 3.1.1.2 shall be increased by an amount at least equal to the withdrawn worth i of the immovable or untrippable 'EA.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by consideration of the following l factors:

1. Reactor coolant system boron concentration,

' - 2. CEA position,

3. Reactor coolant temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration,

6. Samarium concentration, and l

7. Reactor coolant system fluid volume.

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l MILLSTONE - UNIT 2 3/4 1-3 l

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April 19,1978 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT (MTC)

LIMITING CONDITION F'OR OPERATION 3 .1.1. 4 The moderator temperature coefficient (MTC) shall be:

a. Less positive than 0.5 x 10' ak/k/*F whenever THERftAL POWER is < 70% of RATED THERMAL POWER,

b. Less positive than 0.2 x 10-4 ok/k/*F whenever THERMAL POWER is > 70% of RATED THERMAL POWER, and

c. Less negative than -2.4 x 10" ak/k/*F at RATED THERMAL POWER.

APPLICABILITY: MODES 1 and 2**

ACTION:

With the mTderator temperature coefficient outside any one of the above limits, 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 />.

SURVEILLANCE REQUIREMENTS 4 .1.1. 4 .1 The MTC shall be determined'to be within its limits by con-firmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the predicted values.

32

• With K,ff > l.0.

1. See Special Test Exemption 3.10.2.

MILLSTONE - UNIT 2 3/4 1-5 J

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MILLSTONE - UNIT 2 3/4 1-24a

1.0L0 l.oSO

/.040 Augmentation Distance from the Factor Bottom of Core ca p 4 O 1.056 136.6 in y $

f.o3o 1.047 100.0 cn g 1.0355 60.0 1.025 20.0 s 1.014 10.0 3 0.0 0.0 g /.oao i

1.06 0 Ib00 j

i 60 1,0 /00 /.10 /YO 0 .to #

DISTA p' r- Af06 00770H 0E CCC t , I UCH & i Figure 4.2-1 Augmentation Factors Vs. Distance From Bottom of Core

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. *4 August 1. 1975 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.' 9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure aK equivalent to no reater than 0.90 with all full length CEAs (sh8bownandregulating fully inserted.

APPLICABILITY: MODE 6*.

ACTION:

With the requirements of the above specification not satisfied, insnedi-ately suspend CORE ALTERATIONS and initiate and continue boration at >

40 gpm until K is reduced to < 0.90. The provisions of SpecificatTon 3.0.3arenotIhhlicable.

SURVEILLANCE REQUIREMENTS 4.9.1 The boron concentration of the refueling pool shall be detennined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

See Special Test Exception 3.10.4.

MILLSTONE - UNIT 2 3/4 9-1

a e

May 12, 1979 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at-least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s).

APPLICABILITY: MODES 2 and 3.

i ACTION:

a. With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, tannediately initiate and continue boration at > 40 gpm of l MO

> 1720 ppm boric acid solution or its equivaleiit until the 5110TDOWN MARGIN required by Specification 3.1.1.1 is restored,

b. With all full length CEAs inserted and the reactor subcritical by

(

less than the above reactivity equivalent, immediately initiate and continue boration at > 40 gpm of > 1720 apm boric acid l 20 solution or its equivalent until the SliUTDOWi MARGIN reqJired by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length CEA required either partially or fully withdrawn shall be detennined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped fran at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

MILLSTONE - UNIT 2 3/4 10-1

3/4.1 -REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be-made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion RCS boron concentration, and RCS T The most restrictive cond'if. ion occurs at E0L with Tav at no l $0. operating temperature,andisassociatedwithapostul!tedsteamlinebreakaccident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 3.2% o k/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN reovired by Specification 3.1.1.1 is based upon this limiting condition ar.d is con-sistent with FSAR accident analysis assumptions. For earlier periods during the fuel cycle, this value is conservative. With T 6 200 F, the reactivity transients resulting from any postulated acMent are minimal and a 1% A k/k shutdown margin provides adequate protection when the RCS is full.

If the RCS is drained for maintenance work (approximately 4828 cubic feet) the reactor shall be maintained at 2% o k/k SHUTDOWN MARGIN or in the All Rods In configuration.

3/4.1.1.3 BORON DILUTION AND ADDITION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration changes in the Reactor Coolant System.

A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 10,060 + 700/-0 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron concentration changes will be within the capability for operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurance that the coefficient will be maintained within acceptable values throughout each fuel cycle.

MILLSTONE - UNIT 2 B 3/4 1-1 i

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1 REACTIVITY CONTROL SYSTEMS

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BASES 3/4.1.2 B0 RATION SYSTEMS (Continued)

The boron capability required below 2000F is based upon providing a 1% A k/k SHU1DOWN MARGIN at 1400F during refueling with all full and part '

length control rods withdrawn and the primary system full. (approximately i 9260 cubic feet without the pressurizer) This condition requires either 5,050 gallons of 6.25% boric acid solution from the boric acid tanks or )

57,000 gallons of 1720 ppm borated water from the refueling water storage tank.

If the system is drained for maintenance work (approximately 4828 cubic feet) the reactor shall be maintained at 2% ak/k SHUTDOWN MARGIN or in the All Rods In configuration.

3/4.1. 3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (,1) acceptable power distribution limits are z.;ntained. (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the pot.".ial effects of a CEA ejection accident are limited to acceptable levelt I The ACTION statements w ch permit limited variations from the basic requirements are accompanied iy additional restrictions which ensure that the original criteria are met. i The ACTION statements applicable to an immovab13 or untrippable CEA and I

> 20 steps) of two or mre CEAs. require a prompt to a largeofmisalignment shutdown the reactor s (Tnce either of these cd.tfons may be indicative i j

of a possible loss of mechanical functional capat elity of the CEAs and in  ;

the event of a immovable or untrippable CEA. the loss of SHUTDOWN MARGIll.

For small sisalignments (< 20 steps) of t ie CEAs, there is 1) a small degradation in the peaking factors relative t( those assumed in generating

, LCOs and LSSS setpoints for DNBR and linear h rat rate, 2) a small effect on the time dependent long term power distrib stions relative to those used i

in generating LCOs and LSSS setpoints for DNEi and linear heat rate, 3) a small effect on the available SHUTDOWN MARC IN, and 4) a small effect i on the ejected CEA worth used in the safety a talysis. Therefore, the I ACTION statement associated with the small misalignment of a CEA permits  !

1 a on) hour time interval during which attempts may be made to restore the CEA to within its alignment requirements prior to initiating a reduction i in THERMAL POWER. The one hour time limit is sufficient to (1) identify  !

causes of a misaligned CEA. (2) take appropriate corrective action to reelign the CEAs and (3) minimize the effects of menon redistribution. l I

I Overpower margin is provided to protect the core in the event of a l large misalignment (> 20 steps) of a CEA. However, this misalignment l would cause distortion of the core power distribution. The reactor l

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l NILLSTONE - UNIT 2 B 3/41 3

May 12, 1979 PLANT SYSTEMS 1

BASES 107 = Power Level-High Trip Setpoint for two loop operation j X = Total relieving capacity of all safep valves pet steam line in 1bs/ hour = 6.35 x 10 lbs/ hour on l

Y = Maximum relieving capacity of any one safety valve l in 1bs/ hour = 7.94 x 10 lbs/ hour 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS ,

The OPERABILITY of the auxiliary feedwater pumas ensures that the l Reactor Coolant System can be cooled down to less tun 300*f from normal l operating conditions in the event of a total loss of off-site power.

Either motor driven pump or the' steam-driven mp have the required capacity to provide suf'ficient feedwater ow to remove reactor decay heat and reduce the RCS taperature to < -300*F where the shutdown cooling system say be placed into operation fiir continued cooldown.

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 300*F in the event of a total loss of off-site power. The minimum water volume is sufficient to 2-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.

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 8 3/4 7-2 l - . . .- . . - . . , . . . .. .... . . . . . . _ . . . _ .

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.~ - p Docket No. 50-336 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Discussion of Proposed Revisions to Technical Specifications i August, 1980

,. .6 proposed Technical Specification Changes The purpose of this Attachment is to provide an explanation for each of the changes proposed in support of Cycle 4 operation of Millstone Unit No. 2.

(1) In Reference (11), NNEC0 reported to the NRC Staff that the calcula-ted shutdown margin required by Technical Specifications in Mode 5 at Millstone Unit No. 2 was based on a Reactor Coolant System (RCS)

Volume of 9,500 ft 3. However, during Mode 5, the RCS can be drained to the centerline of the hot legs which correspond to an RCS volume of 4,828 ft .3This smaller volume requires a larger shutdown margin to result in acceptable consequences during a postulated boron dilution event in Mode 5.

A reanalysis of the boron dilution event in Mode 5 using a conserva-tive RCS volume of 4,828 ft3was performed. The results of this analysisillustratethatashutdownmarginof2%%with"AllRods Out" of the core will adequately provide the required time of 15 minutes for operator action to mitigate the dilution event. The shutdown margin in Mode 5 with the RCS partially drained can be reduced to 1% A with the core in the "All Rods In" configuration.

The "All Rods In" configuration assures a shutdown margin of 2% AK.

K The analysis of the boron dilution event in Mode 5 for Cycle 4 is provided in References (6) and (7).

Technical Specification pages 1-7, 3/4 1-3, B 3/4 1-1, and B 3/4 1-3 have been revised accordingly to reflect the results of the reanalysis

. of the boron dilution event in Mode 5.

(2) The infomation presented on Figure 2.1-1 on page 2-2 has been revised to reflect the deletion of Figure B 2.1-1 on page B 2-2. The axial power distributions illustrated in Figure B 2.1-1 were valid for the previous vendor's analyses. The Westinghouse axial power distribution

._. . _ . _ . . . _ . _, . _ _ . . _ ..w__. .

methodology is utilized to produce Figure 2.1-1. It is therefore, inappropriate to include Figure B 2.1-1 in the Cycle 4 Technical Specifications.

(3) The Reactor Protection System (RPS) Setpoints listed in Table 2.2-1 on pages 24 and 2-5 have been updated and the allowable values have been revised to include the maximum expected drift assumed to occur (between surveillance intervals) for each trip 11 the accident analyses.

The revised Allowable Values are consistent with Section 2.2.1 of the Limiting Safety System Settings (LSSS)' Bases. Tlie LSSS Bases on page B 2-4 have also been revised to reflect changes nude to Table 2.2-1.

a. Upon review of the bases for the uncertainties utilized in the Power Level-High trip by the NSSS vendor, it was detennined that Technical Specification Table 2.2-1 required revising to be consistent with these bases. Specifically, while the uncer-tainties remain the same (5%) for the Power Level-High trip, the uncertainty shall be combined multiplicatively rather than additively as was the case in previous cycles.

b. The Reactor Coolant Pump (RCP) Speed-Low trip setpoint has been updated to 830 RPM. The present trip setpoint provided to the Staff in Reference (12), as suppiomented by Reference (13),

was calculated by utilizing a nominal RCP operating speed which was derived from measurements of RCS flow rate. This was required as no data on RCP nominal operating speed existed at the time of the Reference (13) submittal.

j

. During Cycle 3 operation, measurements of actual RCP speed at nominal conditions were made. The measured value of the RCP Speed was slightly greater than that derived from the RCS flow rate. This slight deviation in the measured vs. calculated speed of the RCP's resulted in the revised trip setpoint for the Reactor Coolant Pump Speed-Low function.

It is noted that although the Millstone Unit No. 2 Safety Analysis requires the Reactor Coolant Pump Speed-Low trip setpoint be 800 RPM, the actual trip setpoint for Cycle 3 was significantly more conservative (nominally 845 rpm) as will be the case for Cycle 4.

(4) The Safety Limit Bases on pages B 2-1 and B 2-3 and the Limiting Safety System Settings Bases on pages B 2-5, B 2-6, and B 2-S have been revised to reflect the use of the W-3 correlation in the Cycle 4 Safety Analyses. The W-3 correlation relates reactor power, reactor coolant temperature and pressure to DNBR and 1 hits the calculated minimum DNBR to 1.30 which provides a 95/95 probability /

confidence level that DNB will not occur.

(5) In Reference (14), Northeast Nuclear Energy Company reported to the NRC Staff that in-service inspections at Millstone Unit No. 2 had shown that under test conditions, the positive displacement charging pumps deliver approximately 43 gpm. These pumps were previously reported to deliver 44 gpm in the FSAR.

The ECCS/LOCA analyses performed in support of Cycle 4 subsequently utilized a conservative value of 40 gpm on the flow delivered by these pumps and acceptable results were obtained.

Pages 3/41-1, 3/4 9-1, and 3/410-1 have been revised to reflect the conservative charging pump flow of 40 gpm as used in the safety analyses.

(6) The Moderator Temphrature Coefficient (MTC) has been recalculated based on Cycle 4 core kinetics characteristics. page 3/4 1-5 of the Millstone Unit No. 2 Technical Specifications are proposed to be revised accordingly.

,P P (7) Page 3/4 1-24a is proposed to be deleted as the requirements of this specification no longer apply to Millstone Unit No. 2 and the Technical Specifications are clarified accordingly.

(8) Figure 4.2-1'on page 3/4 2-5 has been revised to reflect the'augmen-tation factors which will apply to the Westinghouse Batch F fuel during Cy,le 4. The proposed augmentation factors presented in the attached Figure 4.2-1 envelop the previous cycles' factors and will therefore be utilized for the entire Cycle 4 core.

(9) In Reference (15) and (16), NNECO provided the Staff with the proposed design for the control grade automatic auxiliary feedwater system to be installed at Millstone Unit No. 2. Analyses supporting the design have illustrated that either electric motor driven auxiliar feedwater pump on the steam driven auxiliary feedwater pump will provide adequate feedwater to the steam generators to remove reactor decay heat. The Plant Systems Bases Section 3/4 7.1.2 referring to the Auxiliary Feedwater Pumps on page B 3/4 7-2 are proposed to be modified accordingly.