Text

Proposed Tech Specs Re Admin Changes Which Will Correct Typographical Errors,Provide Clarifications or Make Editorial Changes
	ML20092G483
Person / Time
Site:	Millstone
Issue date:	09/11/1995
From:	Debarba E, Opeka J; NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	;
Shared Package
ML20092G463	List: B15349, Application for Amend to License DPR-65,proposing Changes to Ts,Re Admin Changes Which Will Correct Typo Errors, Provide Clarifications or Make Editorial Changes; ML20092G483;
References
	NUDOCS 9509190322
	Download: ML20092G483 (27)
	v • d • e

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TABLE 2.2-1 March 20, 1989 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS *

.5 FUNCTIONAL UNIT ,

TRIP SETPOINT ALLOWABLE VALUES

10. Thermal Margin / Low Pressure (1) lf ,

g Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to not i

'y Operating exceed the limit lines of exceed the limit lines of '

u Figures 2.2-3 and 2.2-4 (4). Figures 2.2-3 and 2.2-4(4). .

11. Loss of Turbine--Hydraulic 2 500 psig 2 500 psig Fluid (3) Pressure - Low 7.- TABLE NOTATION (1) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER .

@i + % of RATED THERMAL POWER.

m (2) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically -

f, removed at or above 780 psia. '.

(3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER IS 115% of RATED THERMAL POWER. ;

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(4) Calculations of the trip setpoint includes measurements, calculational and processor uncertainties, and dynamic allowances.

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l ,E., (5) Each of four channels actuate on the auctioneered output of two transmitters, one from each steam generator.

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M 9509190322 950911 .

hDR ADOCK 05000336 PDR

TABLE 3.3-3 (Costinued)_ .

ENGINttRED SAFETT FEATURE ACTURil0N STSTEM INSTRtNIDITATION 2

P MINIMUM G TOTAL NO. CflAINitLS CHANNELS APPLICABLE fyIICTI0lWit_WIIIT_ 0F CHAIWlELS_ 10 TRIP OPERA 8tt MODES _ ACTION e 7. CONTAIIINENT PURGE gg l61 g V3LVE ISOLATION ,

d i

s. Contalement Radf ation -

" Nigh .

5. 6 l61 2.,(d)

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seseous Monitor 1d) 1 3 j Particulate Monitor 2. g 1 d) 1 -

3

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8. LOSS OF POWER w

1 a. 4.16 tw Deergency tus I y Undervoltage (Under-s voltage relays) -

1evel one 4/ bus 2/ sus 3/ bus 1.2,3 2 lt

b. 4.15 kg Deergency Bus Undervoltage (Under- i voltage relays) - l -

level two 4/Sts 2/Ses 3/ Sus 1.2,3 2 L

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i 10/7/94 2. - 1 i - G TABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < 1750 psia; bypass shall be automatically removed when pressurizer pressure is 11750 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed below 600 psia; bypass shall be

-v automatically removed at or above 600 psir..

Tach channel has two sensors, high radiathlevel on either sensor will jnitiate containment purge valve isolation.

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

I ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

I a. < 1750 psia; imediately place the inoperable channel in the bypassed condition; restore the inoperable channel to i OPERABLE status prior to increasing the pressurizer pressure above 1750 psia.

b. 2.; 1750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied:

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1. All functional units receiving an input from the bypassed channel are also placed in the bypassed condition.

2. The Minimum Channels OPERABLE requirement is met; however, one r.dditional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing par Specification 4.3.2.1 provided one of the l inoperable channels is placed in the tripped condition. i 1

MILLSTONE - UNIT 2 3/4 3-16 Amendment No. 777 179 ]

one 1

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i TABLE 3.3-3 (Continued) p w a h w at M chewel5 N S ACTION 3 - With('ene ;r ::r: :t,;nn:1: inepe-: tic, ;;;retien .., centingr

' ;; e.ided the containment purge valves are aintained closed.

l ss than the Total ACTION 4 - With the number of OPERABLE channels one Number of Channels and with the pressurizer pressure:

)

a. < 1750 psia: immediately place the inoperable channel in i

the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing the pressurizer i pressure above 1750 psia.

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b. 11750 psia, operation may continue with the ' inoperable s

j cond on i saffffejdcondition,providedthefollowing f The Minimum Channels OPERABLE requirement is met; 1.

however, one additional channel may be removed from i service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing

! per Specification 4.3.2.1 provided EDIB of the j inoperable channels are placed in the bypassed f

condition.

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Amendment No. JJF 179 NILLSTONE - UNIT 2 3/4 3-17 0129

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j TABLE 3.3-5 (Continued) l ENGINEERED SAFETY FEATURES RESPONSE TIMES i

INITIATING SIGNAL AND FUNCTION , RESPONSE TIME IN SECONDS

3. Containment Pressure - Hiah j

a. Safety Injection (ECCS)

) -. 1) High Pressure Safety Injection 5 25.0*/5.0**

2) Low Pressure Safety Injection 1 45.0*/5.0** l l

3) Charging Pumps s35.0*/35.0**

I 4) Containment Air Recirculation System 126.0*/15.0**

j b. Containment Isolation 1 7.5

c. Enclosure Building Filtration System s 45.0*/45.0**

f

! d. Main Steam Isolation s 6.9 I e. Feedwater Isolation 1 14

4. Containment Pressure--Hiah-Hiah ,

a. Containment Spray 1 35. 6 *"'/16. 0**"'

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.)

) 5. Containment Radiation-Hioh i a. Containment Purge Valves Isolation f Counting period

plus 7.5

6. Steam Generator Pressure-Low

a. Main Steam Isolation 1 6.9 l

h. Feedwater Isolation s 14

! 7. Refuelina Water Storaae Tank-Low i

a. Containment Sump Recirculation s 120

3. Steam Generator Level-Low $ go

^

! a. Auxiliary Feedwater System'*' [240*/240**]

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l l N111 stone Unit No. 2 3/43-22 Amendment No. J. F. pl.

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- ** 6/7/94

.. TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES

'(

t TABLE NOT,1 TION

Diesel generator starting and sequence' loading delays included.

** Diesel generator starting and sequence loading delays Ani included.

Offsite power available.

I (1) Header fill time not included. blad ^

(2) udes 3-minute time delay.) k _,___

(3) For Cycle 12 only, OPERABILITY of the auxiliary feedwater (AFW) automatic 4

initiation logic will rely on operator action to ensure successful initiation of AFW. Prior to startup for Cycle 13, modifications to the ,

,' automatic initiation logic for AFW will be implemented to eliminate the I reliance on operator action.

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, Millstone Unit No. 2 3/43-22a Amendment No. J. J. M.

ese N. M1.

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Table 3. 3-8 METEOROLOGICAL MONITORING INSTRUMENTATION

INSTRUMENT MINIMUM

. MINIMUM CHANNELS

<. INSTRUMENT LOCATION ACCURACY OPERABLE

1. WIND SPEED

a. Nominal Elev.142 ft. 1 0.22 m/sec* 1

b. Nominal Elev. 374 ft. 1 0.22 m/sec* 1

2. WIND DIRECTION

a. Nominal Elev.142 ft. 1 5' 1

. b. Nominal Elev. 374 ft. 1 5' 1

3. AIR TEMPERATURE - DELTA T

a. Nominal Elev.142 ft. 1 0.18'F 1 i

b. Nominal Elev. @ ft. 1 0.18*F 1

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Starting speed of anemometer shall be < 0.45 m/sec.

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MILLSTONE - UNIT 2 3/4 3-37 A mtdmht A o. N '

.Fcbruary 26, 1991 j Y

o NEACTOR COOLANT SYSTEM l l SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION i

3.4.8 The specific activity of the primary coolant shall be limited to:' i

a. $ 1.0 pCi/ gram DOSE EQUIVALENT I-131, and \#g\s - '!

b. S 100/1 pCi/ gram. 'd l

g .0 g p ;

l APPLICABILITY: MODES 1, 2, 3, 4, and 5. g d%k x5

! ACTIDN: d o N d !

6R sod MODES 1, 2, and 3*: '# #[c r

a. With the specific activ y of the pr ry coolant > 1.0 poi / gram j

, DOSE EQUIVALENT I-131 t within t allowable limit (below and to '

l' the left of the line shown on Fi re 3.4-1, operation may continue i for up'to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Entry into _ 0PERATIONAL MODE or other _

! s >ecified condition is permittedjursuant to specification 3.0.

Qien subject to this ACTIDM statement. f i b. With the specific activity of the primary coolant > 1.0 pCi/ gram h

DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be j

in H0T STAND 8Y with T,,, < 515'F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ,

c. With the specific activity of the primary coolant > 100/E pC1/ gram, be in HOT STANDBY with T,,, < 515'F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . !

MODES 1, 2, 3, 4 and 5

d. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E #C1/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-2 until the specific activity of the primary coolant is restored to within its limits. .

With T,,9 2 515'p, .

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} MILLST.0NE - UNIT 2 3/44-13 Amendment No. J. JJJ, JJJ,151 l

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d w l August 1, 1975 REFUELING OPERATIONS CONTAINMENT RADIATION MONITORING LIMITING CONDITION FOR OPERATION i h e ,4e,ae,c_.w wet w h o(. (as ews 'A-o T^"K*^b 3.9.9 The -aa+=4 - a+ e :Nibb$((r

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ct y monitors which initiate containment purge valve isolation shall be OPERABLE.

APPLICABILITY: MODE 6.

ACTION:

With less than the above required instrumentation systems OPERABLE, either suspend all operations involving CORE ALTERATIONS and movement of fuel within the containment building or close all penetrations providing direct access from the containment atmosphere to the outside atmosphere, then CORE ALTERATIONS and/or fuel movement within the containment building may proceed for up to 7 days subject to ACTION requirements of Specifi-cation 3.3.3.1, as applicable.

SbRVEILLANCEREQUIREMENTS 4.9.9.1 The specified instrumentation shall be demonstrated OPERABLE by perfonnance of the surveillance requirements of Specification 4.3.3.1.

4.9.9.2 All penetrations providing direct access from the containment atmosphere to the outside atmosphere shall be verified closed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS or feel movement within the containment building when less than the above required instrumentation

, systems are OPERABLE.

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5 MILLSTONE - UNIT 2 3/4 9-9 l

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' March 30,1992 p . :.

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j 3/4.2 POWER DISTRIBtJTJ0N LIMITS

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i BASES k j k 9

3/4.2.1 LINEAR HEAT RATE , J 1

1 The limitation on linear heat rate ensures that in the event of a LOCA, }

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the peak temperature of,the fuel cladding will not exceed 2200'F. -

i Either of the two core power distribution monitoring systems, the Excore .

Detector Monitoring System and the Incore Detector Monitoring System, provide i adequate monitoring of the core power distribution and are capable of verify- '

. ing that the linear heat rate does not exceed its limits. The Excore Detector

! Monitoring System performs this function by continuously monitoring the AXIAL l SHAPE INDEX with two OPERABLE excore neutron flux detectors and verifying that '

. the AXIAL SHAPE INDEX is maintained within the allowable limits" f F:;;re Nu.waf h '

s M4 using the Power Ratio Recorder. The power dependent limits of .the Power omuco ! --

?

Ratic Recorder are less than or equal to the limits;7 f:;;re L:-:. In\$b '

f conjunction the AXIAL SHAPEwith INDEXthe use limits, insertion limits of Specifications -3.

of theassumptions the following excore monitoring system are made: 1) the CEA and in esta

.3.2, 3.1.3.5 and 3.1.3.6 are satisfied, A c.cac.

2) th; fle:: ;: kMg :27::t:tica fe;te e ere ee :h=, in Fi;;r; 4.21,'al) the MN !' ;

AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 4%

i M) the TOTAL UNR00DED INTEGRATED RADIAL PEAXING FACTOR does not exceed the l g lici,ts of Specification 3.2.3. I

} The Incore Detector Monitoring System continuously provides a direct 1 measure of the peaking factors and the alarus which have been established for 4 the individual incore detector segments ensure that the peak linear heat rates j will be maintained within the allowable Elimits f 7:;;r; L?-L The setpoints N d h ,., g

for these alarms include allowances, set in the conservative directions, for DM1 l 1)? flux peaking augmentation factort: : d:= i: Fi;;r: 4.2-3, 2) a measure- L% !

ment-calculational uncertainty factor - f 1.07, 3) an engineering uncertainty %R3 i L factor c' !.02, 4) an allowance ef 1.01 for ' axial fuel densification 'and

[; thermal h02. Note theexpansion, Items (1) and (4) aboveand are 5) only a THERMAL applicable to fuel POWER batches w measu i i

A* through "L". W ns % '

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3/4.2.3 and 3/4.2.4 TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTORS AZIMtJTHAL POWER TILT - T_ y gggFT y ,

T The limitations on F and T are provided to 1) ensure that the assump- !

tions used in the analysis for %stablishing the Linear Heat Rate and Local I" power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits, and, 2) ensure that the assumptions used in the analysis establishing the DNB Margin LCO, and Thermal

Margin / Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. If F or T exceed their basic ,;

li::itations, operation may continue under the addit onal 9estrictions imposed I 4

by the ACTION statements since these additional restrictions provide adequate :

provisions to assure that the assumptions used in establishing the Linear Heat ! l Rate. Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS i

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MILLSTONE . UNIT 2

"* B 3/4 2-1 Amendment No. M JJ, JJ2. l JM ,199 155 P

b --

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REACTOR C0OLANT SYSTEM BASES evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice )

inspection of steam generator tubing also provides a means of characterizing the l nature and cause of any tube degradation so that corrective measures can be taken. l The plant is expected to be operated in a manner such that the secondary '

coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the

~l limitation of steam generator tube leakage between the primary coolant system

, and the secondary coolant system (primary-to-secondary leakage = 0.10 GPM,

persteamgenerator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand

the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of 0.10

gallon per minute can readily be detected by radiation monitors of steam I generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and i

plugged.

2 Wastage-type defects are unlikely with proper chemistry treatment of the '

secondary coolant. However, even if a defect should develop in service, it will i be found during scheduled inservice steam generator tube examinations.

i Plugging or sleeving will be required for all tubes with imperfections exceeding

the plugging limit of 40% of the tube nominal wall thickness. Sleeving repair will be limited to those steam generator tubes with a defect between the tube sheet and the first eggerate support. Tubes containing sleeves with imperfections exceeding the plugging limit will be plugged. Steam generator

. tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be immediately reported to the

! Commission pursuant to 10 CFR 50.72. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection,

and revision of the Technical Specifications, if necessary.

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7b MILLSTONE - UNIT 2 8 3/4 4 h Amendment No. 77, 77, JJ JJ, JJ, m 111,111,138,

February 3', 1987 i.

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! REACTOR COOLANT SYSTEM BASES l

i 3/4.4.7 CHEMISTRY _

, The limitations on Reactor Coolant System contaminants ensure that

' corrosion of the Reactor Coolant System is minimized and reduce the poten-i tial for Reactor Coolant System leakage or failure due to stress corrosion.

Maintaining the concentrations of the contaminants within the Steady State i

Limits shown on Table 3.4-1 provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life

of the plant. The associated effects of exceeding theCorrosion oxygen, chloride and studies show i

2 fluoride limits are time and temperature dependent.

i that operation may be continued with contaminant concentration levels in l

excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval ,

l permitting continued operation within the restrictions of the Transient . j i Limits provides time for taking corrective actions to restore the contaminant l

concentrations to within the Steady State Limits.

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The surveillance requirements provide adequate assurance that concentra- l tions in excess of the limits will be detected in sufficient time to take . !

l corrective action. )

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3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure l l that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an l

i appropriately small fraction of Part 100 limits 55"-M following a steam steedy generator 5 tete Pe*y 'a-tube rupture accident.f :: jenttf en ef th :P l

nd:ry :t::r ;;r: rat:r h :h ;: 7:t: Of 1.0 CP" : d : ::::;rr::t h:: Of j Off:it: :h:trf::1 ; .::r l i

f The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 uti/ gram

' DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, l j

accomodates possible iodine spiking phenomenon which may occur following l changes in THERKAL POWER. l l

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, 1 B 3/4 4-4 Amendment No.115 1

MILLSTONE - UNIT 2 YS N7 i

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b RADIOACTIVE ETFLUENTS BASES e i

j 3/4.11.2 GASEOUS EFFLUENTS j

j 3/4.11.2.I' DOSE RATE

( This specification is provided to ensure that the dose rate at anytime 5

from gaseous effluents from all units on the site will be within the annual F

dase limits of 10 CFR Part 20 for all areas offsite. The annual dose limits

] cre the doses associated with the concentrations of 10 CFR Part 20, Appendix

B Table II. These limits provide reasonable assurance that radioactive caterial discharged in gaseous effluents will not result in the exposure of 4

en individual offsite to annual average concentrations exceeding the limits 1

specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase 4

e in the atmospheric diffusion factor above that for the site boundary. The spscified release rate limits restrict, at all times, the corresponding

3: ama and beta dose rates above background to an individual at or beyond the site boundary to S 500 arem/ year to the total body or to 5 3000 mrem / year to 4

the skin. These release rate limits also restrict, at all times, the l

corresponding thyroid dose rate above background to{ inh .t via the w /wunN j ashe&adant pathway t 3 1500 arem/ year, hr th: :::,:t a - - c.--.- .

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r. w orne oth h c o-b 3/4.11.2.2 DOSE, NOBLE GASES l i

This specification is provided to implement the requirements of Sections ! l l II.B. III.A and IV.A of Appendix I, 10 CFR Part 50. The Limiting condition ',

i for Operation implements the guides set forth in Section II.B of Appendix 1. -

l. The ACTION statements provide the required operating flexibility and at the

! sene time implement the guides set forth in Section IV.A of Appendix I to

, assure that the releases of radioactive material in gaseous effluents will ,

i l be kept "a3 low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section IJJ.A of Appendix 1 that conform with l

{

the guides of Appendix 3 to be shown by calculational procedures based on

! codals and data such that the actual exposure of an individual through the '

eppropriate pathways is unlikely to be substantially underestimated. The i

' dose calculations established in the ODCM for calculating the doses due to tha actual release rates of radioactive noble gases in gaseous effluents will I i

?

b2 consistent with the methodology provided in Regulatory Guide 1.109,

' " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," j Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating i

Attespheric Transport and Dispersion of Gaseous Effluents in Routine Releases frca light Water-Cooled-heactors," Revision 1, July 1977.

The ODCM equations provided for determining the air doses at the site j

! . bsundary are based upon utilizing successively more realistic dose l

! calculational methodologies. More realistic dose calculational methods are i

r MILLSTONE - UNIT 2 'n !

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B 3/4 11-2 Amendment No. 104 l

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i Docket No. 50-336 B15349 i

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Attachment 4 Millstone Nuclear Power Station, Unit No. 2 Proposed Technical Specifications Revision Administrative Changes to Technical Specifications Retyped Technical Specifications l

September 1995

l em S r-6 TABLE 2.2-1 -

l k REACTOR PROTECTIVE INSTRUNENTATION TRIP SETP0 INT LIMITS E

FUNCTIONAL UNIT TRIP SETP0 INT ALLOWABLE VALUES

10. Thermal Margin / Low Pressure (1)

[

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

11. Loss of Turbine--Hydraulic 1 500 psig 1 500 psig

Fluid (3) Pressure - Low l

TABLE NOTATION I

(1) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER l 1s 1 5% of RATED THERMAL POWER. l En (2) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically i removed at or above 780 psia.

(3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER IS 1 15% of RATED THERMAL POWER.

! (4) Calculations of the trip setpoint includes measurements, calculational and processor uncertainties, and dynamic i

allowances.

g E (5) Each of four channels actuate on the auctioneered output of two transmitters, one from each steram generator.

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, TABLE 3.3-3 (Continued) I h / ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION R /

i MINIMUM g TOTAL NO. CHANNELS CHANNELS APPLICABLE ;

q FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION l l 7. CONTAINMENT PURGE VALVE ISOLATION

a. Containment Radiation-High 5, 6 Gaseous Monitor 2 1 1 3 Particulate Monitor 2 1 1 3

( , 8. LOSS OF POWER s

a. 4.16 kv Emergency Bus

, Undervoltage (Under-a voltage relays) -

level one 4/ bus 2/ Bus 3/ bus 1, 2, 3 2

b. 4.16 kv Emergency Bus Undervoltage (Under-d voltage relays) -

[ 1evel two 4/ Bus 2/ Bus 3/ Bus 1, 2, 3 2 E

R a

.F m___ __ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _

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16BLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < 1750 psta;

bypass shall be automatically removed when pressurizer pressure is 21750 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed below 600 psia; bypass shall be automatically removed at or above 600 psia.

l (d) Deleted l (e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

ACTION STATEMENTS I With the number of OPERABLE channels one less than the Total ACTION 1 -

Number of Channels, restore the inoperable channel to OPERABLE i status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

4 l ACTION 2 - With the number of OPERABLE channels one less than the Total i

Number of Channels and with the pressurizer pressure:

< 1750 psia; immediately place the inoperable channel in a.

the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing the pressurizer

pressure above 1750 psia. l 4 i

b. 2 1750 psia, operation may continue with'the inoperable !

channel in the bypassed condition, provided the following !

conditions are satisfied:

j ;

j 1. All functional units receiving an input from the i bypassed channel are also placed in the bypassed

! condition.

1 i 2. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing i per Specification 4.3.2.1 provided one of the inoperable channels is placed in the tripped condition.

l

. NILLSTONE - UNIT 2 3/43-16 AmendmentNo.JJp,J77, 0210

O TABLE 3.3-3 (Continued) r.CTION 3 - With less than the minimum channels OPERABLE the containment purge valves are to be maintained closed.

ACTION 4 - With the number of OPERABLE channels one less than the Total

Number of Channels and with the pressurizer pressure:

a. < 1750 psia: immediately place the inoperable channel in 9 the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing the pressurizer 1

pressure above 1750 psia,

b. 21750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following condition is satisfied:

1. The Minimum Channels OPERABLE requirement is met; however, one additional cnannel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided HQIB of the inoperable channels are placed in the bypassed condition.

NILLSTONE - UNIT 2 3/4 3-17 Amendment No. J77. J77, 0210

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1

3. Containment Pressure - Hiah

a. Safety Injection (ECCS)

1) High Pressure Safety Injection 1 25.0*/5.0**

2) Low Pressure Safety Injection 1 45.0*/5.0**

3) Charging Pumps s 35.0*/35.0**

4) Containment Air Recirculation System 1 26.0*/15.0**

b. Containment Isolation s 7.5

c. Enclosure Building Filtration System s 45.0*/45.0**

d. Main Steam Isolation 1 6.9

e. Feedwater Isolation 1 14

4. Containment Pressure--Hiah-Hiah

a. Containment Spray 1 3 5. 6*"'/16. 0* *"'

5. Containment Radiation-Hiah

a. Containment Purge Valves Isolation s Counting period plus 7.5

. 6. Steam Generator Pressure-Low

a. Main Steam Isolation 1 6.9

b. Feedwater Isolation s 14

7. Refuelina Water Storaae Tank-low

a. Containment Sump Recirculation 5 120

8. Steam Generator Level-Low

a. Auxiliary Feedwater System s 240 l Millstone Unit No. 2 3/4 3-22 Amendment No. J. J. 97, om 91, 191 H F,

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES TABLE NOTATION

Diesel generator starting and sequence loading delays included.

- Diesel generator starting and sequence loading delays agi included.

j Offsite power available.

(1) Header fill time not included.

(2) Deleted l

. (3) For Cycle 12 only, OPERABILITY of the auxiliary feedwater (AFW) automatic l initiation logic will rely on operator action to ensure successful

initiation of AFW. Prior to startup for Cycle 13, modifications -to the i automatic initiation logic for AFW will be implemented to eliminate the reliance on operator action.

s 9

l i

1 N111 stone Unit No. 2 3/4 3-22a Amendment No. J 7, 77,

- - ' 77,177,J77,

i Table 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT MINIMUM MINIMUM CHANNELS INSTRUMENT ACCURACY OPERABLE

! LOCATION

1. WIND SPEED

a. Nominal Elev.142 ft. 0.22 m/sec* 1

b. Nominal Elev. 374 ft. 0.27 m/sec* 1

2. WIND DIRECTION

a. Nominal Elev.142 ft. 5' 1

b. Nominal Elev. 374 ft. 5' 1

3. AIR TEMPERATURE - DELTA T

a. Nominal Elev.142 ft. 0.18'F 1

b. Nominal Elev. 374 ft. 0.18'F 1 l

i 3

4

4 Starting speed of anemometer shall be < 0.45 m/sec.

MILLSTONE - UNIT 2 3/4 3-37 Amendment No. JJ, 0179

. REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. s 1.0 yC1/ gram DOSE EQUIVALENT I-131, and

b. i 100/E pC1/ gram.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2, and 3*:

a. With the specific activity of the primary coolant > 1.0 yC1/ gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Specification 3.0.4 is not applicable. Entry into an OPERATIONAL MODE or other specified condition is permitted in accordance with the ACTION statements.

b. With the specific activity of the primary coolant > 1.0 yCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T,y, < 515'F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. With the specific activity of the primary coolant > 100/E pCi/ gram, be in HOT STANDBY with T,,, < 515'F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

. MODES 1, 2, 3, 4 and 5:

d. With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-2 until the specific

activity of the primary coolant is restored to within its limits.

I

With T,,, 1 515'F.

MILLSTONE - UNIT 2 3/4 4-13 Amendment No. J. JJJ, JJ), JJJ, 0212

I l l i I ,; I I.

REFUELING OPERATIONS CONTAINNENT RADIATION MONITORING LIMITING CONDITION FOR OPERATION I l

1 3.9.9 A minimum of one channel each of gaseous and particulate airborne radioactivity monitors which initiate containment purge valve isolation shall be 0FERABLE.

APPLICABILITY: MODE 6.

ACTION:

With less than the above required instrumentation systems OPERABLE, either suspend all operations involving CORE ALTERATIONS and movement of fuel within the containment building or close all penetrations providing direct access from the containment atmosphere to the outside atmosphere, then CORE ALTERATIONS and/or fuel movement within the containment building may proceed for up to 7 days subject to ACTION requirements of Specifi-cation 3.3.3.1, as applicable.

SURVEILLANCE REQUIRENENTS 4.9.9.1 The specified instrumentation shall be demonstrated OPERABLE by performance of the surveillance requirements of Specification 4.3.3.1.

4.9.9.2 All penetrations providing direct access from the containment atmosphere to the outside atmosphere shall be verified closed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS or fuel movement within the containment building when less than the above required instrumentation systems are OPERABLE.

NILLSTONE - UNIT 2 3/4 9-9 Amendment No.

0211

i

3/4.2 POWER DISTRIBUTION LINITS BASES I 3/4.2.1 LINEAR HEAT RATE 1

The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200*F.

Either of the two core power distribution monitoring systems, the Excore

Detector Monitoring System and the Incore Detector Monitoring System, provide

adequate monitoring of the core power distribution and are capable of verifying

! that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with two OPERABLE excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits specified in

! the Core Operating Limits Report using the Power Ratio Recorder. The power i dependent limits of the Power Ratio Recorder are less than or equal to the limits specified in the Core Operating Limits Report. In conjunction with the l use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX l limits, the following assumptions are made: 1) the CEA insertion limits of ,

Spscifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT l restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL UNR00DED

INTEGRATED RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.3.

l The Incore Detector Monitoring System continuously provides a direct

measure of the peaking factors and the alarms which have been established for ,

- the individual incore detector segments ensure that the peak linear heat rates ,

l will be maintained within the allowable limits specified in the Core Operating l Limits Report. The setpoints for these alarms include allowances, set in the

conservative directions, for I) a flux peaking augmentation factor, 2) a measurement-calculational uncertainty factor, 3) an engineering uncertainty ;

, factor, 4) an allowance for axial fuel densification and thermal expansion, and l 5) a THERMAL POWER measurement uncertainty factor specified in the Core

! Operating Limits Report. Note the Items (1) and (4) above are only applicable to fuel batches "A" through "L".

3/4.2.3 and 3/4.2.4 TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTORS FJ MQ AZIMUTHAL POWER TILT - T q l The limitations on FT and T are provided to 1) ensure that the assump-tions used in the analys[s for $stablishing the Linear Heat Rate and Local

, power Density - High LCOs and LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits, and, 2) ensure that the

assumptions used in the analysis establishing the DNB Margin LCO, and Thermal 4 Margin / Low Pressure LSSS setpoints remain valid _ during operation at the various allowable CEA group insertion limits. If F or T exceed their basic limitations, operation may continue under the addi$1onal @estrictions imposed by the ACTION statements since these additional restrictions provide adequate

! provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin / Low Pressure and Local Power Density - High LCOs and LSSS 1

NILLSTONE - UNIT 2 83/42-1 Amendment No. 77, pg. 172,

= 119,199,199 i

1______

I I J e 1 e

i REACTOR C0OLANT SYSTEM i

! BASES i

j evidence of mechanical damage or progressive degradation due to design, eanufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary 1 coolant will be maintaimd within those chemistry limits found to result in

negligible corrosion of the steam generator tubes. If the secondary coolant !

l chemistry is not maintained within these limits, localized corrosion may likely l 4

result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 0.10 GPM, i per steam generator). Cracks having a primary-to-sesondary leakage less than j this limit during operation will have an adequate margin of safety to withstand

the loads imposed during normal operation and by postulated accidents. ,

Operating plants have demonstrated that primary-to-secondary leakage of 0.10 i

gallon per minute can readily be detected by radiation monitors of steam j i generator blowdown. Leakage in excess of this limit will require plant shutdown ;

' and an unscheduled inspection, during which the leaking tubes will be located and !

plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the

! secondary coolant. However, even if a defect should develop in service, it will I

be found during scheduled inservice steam generator tube examinations.

Plugging or sleeving will be required for all tubes with imperfections exceeding

the plugging limit of 40% of the tube nominal wall thickness. Sleeving repair

will be limited to those steam generator tubes with a defect between the tube j sheet and the first eggerate support. Tubes containing sleeves with i j imperfections exceeding the plugging limit will be plugged. Steam generator tube inspections of operating plants have demonstrated the capability to reliably j

detect degradation that has penetrated 20% of the original tube wall thickness.

! Whenever the results of any steam generator tubing inservice inspection j C-3, these results will be immediately reported to the fall into Category Commission pursuant to 10 CFR 50.72. Such cases will be considered by the

Commission on a case-by-case basis and may result in a requirement for

! analysis, laboratory examinations, tests, additional eddy-current inspection, j

and revision of the Technical Specifications, if necessary.

i NILLSTONE - UNIT 2 B 3/4 4-2b Amendment No. 17, 77 77, 77, pp, l

- Ill,111,lif, l

l

~

REACTOR COOLANT SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System contaminants ensure that corrosion of the Reactor Coolant System is minimized and reduce the poten-tial for Reactor Coolant System leakage or failure due to stress corrosion.

Maintaining the concentrations of the contaminants within the Steady State Limits shown on Table 3.4-1 provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limit s, up to the Transient Limits, for the specified limited time intervals without having a significant effect oc the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Ste Wy State Limits.

The surveillance requirements provide adequate assurance that concentra-tions in excess of the limits wil' be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident. l The ACTION statement permitting POWER OPERATION to continue for limited l time periods with the primary coolant's specific activity > 1.0 uCi/ gram !

DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

B 3/4 4-4 Amendment No. JJJ, NIglSTONE-UNIT 2

a- RADI0 ACTIVE EFFLUENTS BASES 4

3.4.11.2 GASE0US EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose rate at anytime from gaseous effluents from all units on the site will be within the annual

' 1

dose limits of 10 CFR Part 20 for all areas offsite. The annual dose limits J are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table - II. These limits provide reasonable assurance- that radioactive material discharged in gaseous effluents will not result in the exposure of an

, individual offsite to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).

For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the

atmospheric diffusion factor above background to an individual at or beyond the site boundary to s 500 mrem / year to the total body or to s 3000 mres/yecr

- to the skin. These release rate limits also restrict, at all times, the corresponding thyroid or any other organ dose rate above background to a child via the inhalation pathway to s 1500 mrem / year.

3/4.11.2.2 DOSE. N0BLE GASES This specification is provided to implement the requirements of Sections II.B. III.A and IV.A of Appendix I,10 CFR Part 50. The Limiting Condition ,

-for Operation implements the guides set forth in Section II.B of Appendix I. r

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as reasonably achievable." The Surveillance Requirements ,

implement the requirements in Section III.A of Appendix I that conform with the guides of Appendix I to be shown by calculational procedures based on

models and data such that the actual exposure of an individual through the

' dose calculations established in the ODCM for calculating the doses due to the 4

actual release rates of radioactive noble gases in gaseous effluents will be ;

consistent with the methodology provided in Regulatory Guide 1.109, ;

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"

Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases ,

from Light-Water-Cooled-Reactors," Revision 1, July 1977.

E

~

The ODCM equations provided for determining the air doses at the site boundary are based upon utilizing successively more realistic dose calculational methodologies. More realistic dose calculational methods are s

Millstone Unit 2 B 3/4 11-2 Amendment No. Jpf, 0215 4 r

,-_

ML20092G483

Text

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