ML20096G259

Amend 194 to License DPR-65,changing Listed TS Sections, Tables & Bases Sections

ML20096G259
Person / Time
Site:	Millstone
Issue date:	01/17/1996
From:	Mckee P NRC (Affiliation Not Assigned)
To:
Shared Package
ML20096G263	List: ML20096G259 ML20096G265
References
NUDOCS 9601250079
Download: ML20096G259 (16)

Text

_. _. _ _. _

nng g p

UNITED STATES g

j NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D.C. 20eeH001 i

NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT Als POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET N0. 50-336 NILLSTONE NUCLEAR POWER STATION. UNIT NO. 2 ANENDMENT TO FACILITY OPERATING LICENSE Amendment No. 194 License No. DPR-65 l.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee) dated September 11, 1995, as supplemented November 15, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Consission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9601250079 960117 PDR ADOCK 05000336 P

PDR

~

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-65 is hereby amended to read as follows:

(2)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 194, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.-

3.

This license amendment is effective as of the date of issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMISSION Phillip. McKee, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 17, 1996 c'

, 'G ;

aw,

t

l ATTACMENT TO LICENSE AMENDMENT NO.194 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET N0. 50-336 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert l

2-5 2-5 3/4 3-14 3/4 3-14 3/4 3-16 3/4 3-16 3/4 3-17 3/4 3-17 3/4 3-22 3/4 3-22 3/4 3-22a 3/4 3-22a 3/4 3-37 3/4 3-37 3/4 4-13 3/4 4-13 3/4 9-9 3/4 9-9 l

B 3/4 2-1 B 3/4 2-1 B 3/4 4-2a B 3/4 4-2b 8 3/4 4-4 B 3/4 4-4 8 3/4 11-2 B 3/4 11-2 1

+

4 9

Bi!!

r-5 Taste z.z-i l

REACTOR PROTECTIVE INSTIMENATION TRIP SETP0IN LIMITS i

t E

FINICTIONAL INIIT TRIP SETP0 INT ALLOMABLE VALUES

10. Thennal Margin / Low Pressure (1) r M

Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted Operating exceed the limit lines of to not 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 2 500 psig 2 500 psig Fluid (3) Pressure - Low TAsLE NOTATION l

(1) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER.

li Ei (2) Trip may be manually bypassed below 780 psia when all CEAs are fully inserted; bypass shall be automatically 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 2 15% of RATED THERMAL POWER.

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

g E+

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

l 9

M.

f

=

=

i w

a

t l

Em TABLE 3.3-3 (Continued) t r-h ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 9

m MINIMUM g

TOTAL NO.

CHANNELS CHANNELS APPLICABLE l

q FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION I

7.

CONTAlletENT PURGE VALVE ISOLATION

[

a.

Containment Radiation-High 5, 6 Gaseous Monitor 2

1 1

3 Particulate Monitor 2

1 1

3 i

8.

LOSS OF POWER j

,s i

i a.

4.16 kv Emergency Bus Undervoltage (Under-voltage relays) -

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

b.

4.16 kv Emergency Bus Undervoltage (Under-voltage relays) -

(

level two 4/ Bus 2/ Bus 3/ Bus 1, 2, 3 2

=

n y.

1 4

~

t C

c G

s m

e J

TABLE 3.3-3 (Continued)

TABLE NOTATION

]

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

bypass shall be automatically removed when pressurizer pressure is i 1750 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.

(d) Deleted l

(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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

ACTION 2 -

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

b.

11750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied:

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 additional channel may be removed from service for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.2.1 provided one of the inoperable channels is placed in the tripped condition.

NILLSTONE - UNIT 2 3/43-16 Amendment No. JM, JM,194 0210

TABLE 3.3-3 (Continued)

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

b.

1 1750 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following condition is satisfied:

l.

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

NILLSTONE - UNIT 2 3/4 3-17 Amendment No. JJJ 17J.194 0210

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 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 1 35.0*/35.0**

4)

Containment Air Recirculation System 1 26.0*/15.0**

• b.

Containment Isolation 1 7.5 c.

Enclosure Building Filtration System 1 45.0*/45.0**

d.

Main Steam Isolation 1 6.9 e.

Feedwater Isolation 1 14 4.

Containment Pressure--Hiah-Hiah a.

Containment Spray 5 3 5. 6*"'/16.0**"8 5.

Containment Radiation-Hiah a.

Containment Purge Valves Isolation s Counting period plus 7.5 6.

Steam Generator Pressure-Low

a. Main Steam Isolation s 6.9 b.

Feedwater Isolation 5 14 7.

Refuelina Water Storace Tank-Low a.

Containment Sump Recirculation 1 120 8.

Steam Generator level-Low a.

Auxiliary Feedwater System

s 240 l

i Millstone Unit No. 2 3/4 3-22 Amendment No. 1 J. 97, anu 91, 197 119,194

l 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 ag.1 included.

Offsite power available.

(1) Header fill time not included.

j (2) Deleted l

(3) For Cycle 12 only, OPERABILITY of the auxiliary feedwater (AFW) automatic 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 reliance on operator action.

l l

1 l

l l

l l

l N111 stone Unit No. 2 3/4 3-22a Amendment No. 1,7,97,94 91, 191, 119,1 osoe

1

)

Table 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT MINIMUM MINIMUM CHANNELS INSTRUMENT LOCATION ACCURACY OPERABLE 1.

WIND SPEED

a. Nominal Elev.142 ft.

i 0.22 m/sec*

1

b. Nominal Elev. 374 ft.

0.22 m/sec*

1 2.

WIND DIRECTION

a. Nominal Elev.142 ft.

1 5' I

b. Nominal Elev. 374 ft.

5*

1 I

3.

AIR TEMPERATURE - DELTA T

a. Nominal Elev.142 ft.

0.18'F 1

b. Nominal Elev. 374 ft.

0.18'F 1

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

MILLSTONE - UNIT 2 3/4 3-37 Amendment No. (),194 017e

REACTOR C0OLANT SYSTEN SPE IFIC ACTIVITY LIMITING COMITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. 1 1.0 #C1/ gram DOSE EQUIVALENT I-131, and

b. s 100/E pC1/ gram.

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

ACTION:

NODES 1, 2, and 3*:

With the specific activity of the primary coolant > 1.0 pCi/ gram a.

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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Specification 3.0.4 is not applicable.

l b.

With the specific activity of the primary coolant > 1.0 #C1/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T,,< 515'F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

With' the specific activity of the primary coolant > 100/E pC1/ gram, c.

be in HOT STANDBY with T,,, < 515*F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

NODES 1, 2, 3, 4 and 5:

d.

With the specific activity of the primary coolant > 1.0 pCf/ 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.

• With T,,, 1 515'F.

NILLSTONE - UNIT 2 3/44-13 Amendment No. J. JJJ, JJJ, 171, 194 0212

REFUELING 0PERATIONS CONTAIMENT RADIATION N0NITORING LIMITING COMITION FOR OPERATION 3.9.9

- A minimum of one channel each of gaseous and particulate airborne radioactivity 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 butiding may proceed for up to 7 days subject to ACTION requirements of Specifi-cation 3.3.3.1, as applicab1?.

SURVEILLANCE REQUIREMENTS 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS or fuel movement within the containment building when less than the above required instrumentation systems are OPERABLE.

MILLSTONE - UNIT 2 3/4 9-9 Amendment No;94 4111

3/4.2 P0WER DISTRIBUTIpp LIMITS BASES

)

3/4.2.1 LINEAR HEAT RATE 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 perfoms 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 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 limits, the following assumptions are made:

1) the CEA insertion limits of j

Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.3.

The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for i

the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits specified in the Core Operating Limits Report.

The setpoints for these alarms include allowances, set in the conservative directions, for 1) 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

5) a THERMAL POWER measurement uncertainty factor specified in the Core Operating Limits Report.

Note the Items (1) and (4) above are only applicaSle i

to fuel batches "A" through "L".

3/4.2.3 and 3f4.P.o TOTAL UNRODDED INTEGRATED RADIAL PEAKING FACTORS FJ AND AZIMUTHAL POWE tTT

- T.,

The limitations on FI. 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 DN8 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 limitations, operation may continue under the addi;ional Eestrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Themal Margin / Low Pressure and Local Power Density - High LCOs and LSSS MILLSTONE - W IT 2 B 3/4 2-1 Amendment No. M. M. 172, em IM, IM, IM, 194

REACTOR C00LANT SYSTEN 3

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 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 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 limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage 0.10 GPM, per steam generator).

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 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 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 i

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 i

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

i 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.

MILLSTONE - LMIT 2 B 3/4 4-2b Amendment No. 77,77,51,77,77, em 111, 111, 119, 194

REACTOR C0OLANT SYSTEM BASES 3/4.4.7 CHEMISTRY t

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 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 permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentra-tions in' excess of the limits will 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 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.

MILLSTONE - UNIT 2 8 3/4 4-4 Amendment No. JJJ,194 0213

RADI0 ACTIVE EFFLUENTS BASES i

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

The annual dose limits 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 mre# year to the total body or to 5 3000 are# year 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 are# year.

3/4.11.2.2 DOSE. NOBLE 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.

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 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.

The ODCM equations provided for determining the air doses at the site boundary are based upon utilizing successively more realistic dose i

calculational methodologies. More realistic dose calculational methods are Millstone Unit 2 8 3/4 11-2 Amendment No. JpJ,194 0215

-