ML17059C390

Amend 84 to License NPF-69,changing TS to Update Terminology & References to 10CFR50.55a(f) & (G) Consistent with 1989 Edition of Section XI of ASME BPV Code & Consistent with Second 10-year Interval of ISI & IST Program Plans for NMP2

ML17059C390
Person / Time
Site:	Nine Mile Point
Issue date:	12/03/1998
From:	Bajwa S NRC (Affiliation Not Assigned)
To:
Shared Package
ML17059C391	List: ML17059C390 ML17059C394
References
NUDOCS 9812100159
Download: ML17059C390 (36)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NIAGARAMOHAWKPOWER CORPORATION DOCKET NO. 50-410 NINE MILEPOINT NUCLEAR STATION UNIT2 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No.

84 License No. NPF-69

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

A.

The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated February 5, 1998, 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 1; B.

The'facility willoperate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) 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 willnot 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.

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. NPF-69 is hereby amended to read as follows:

98i2i00i59 PSi203 PDR ADGCK 050004iO

,j P

PDR

f>

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No.84 are hereby incorporated into this license.

Niagara Mohawk Power Corporation shall operate the facilityin accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance to be implemented within 30 days.

h FOR THE NUCLEAR REGUL'ATORYCOMMISSION I

S. Singh Bajwa, Director Project Directorate l-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 3,

1998

OO

ATTACHMENTTO LICENSE AMENDMENTNO.

TO FACILITYOPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Replace the following page of the Appendix A Technical Specifications with the attached page.

The revised page is identified by Amendment number and contains vertical lines indicating the areas of change.

Remove Insert X

xiv 3/4 0-2 3/4 3-87 3/4 4-14 3/4 4-24 3/4 4-26 3/4 4-35 3/4 4-36 3/4 8-11 3/4 10-7 B3/4 4-5 B3/4 10-1 5-9 X

xiv 3/4 0-2 3/4 3-87 3/4 4-14 3/4 4-24 3/4 4-26 3/4 4-35 3/4 4-36 3/4 8-11 3/4 10-7 B3/4 4-5 B3/4 10-1 5-9

NDE OO IMITIN NDITI NS F R

PERATI N AND RVEILLAN E E

IREM NT EA T R

LANT Y TEM (Continued)

~PA E

3/4A.6 PRESSURE/TEMPERATURE LIMITS Reactor Coolant System..............'................'........

'3/44-24 Figure 3.4.6.1-1 Minimum Beltline Downcomer Water Temperature for Pressurization During Hydrostatic Testing and System Leakage Testing (Reactor Not Critical)............

3/4 4-26 Figure 3.4.6.1-2'inimum Beltline Downcomer Water Temperature for Pressurization During Heatup and Low-Power Physics Tests (Reactor Not Critical) (Heating Rate s 100 F/HR).........,....

)................

3/4 4-27 Figure 3.4.6.,1-3 Minimum Beltline Downcomer Water Temperature for Pressurization During Cooldown and Low-Power Physics Tests (Reactor Not Critical)

(Cooling Rate a 100'F/HR) 3/4 4-28 Figure 3.4.6.1-4 Minimum Beltline Downcomer Water Temperature for Pressurization During Core Operation (Core Critical) (Heatup at a Heating Rate a 100 F/HR)...........

3/44-29 Figure 3.4.6.1-5 Minimum Beltline Downcomer Water Temperature for Pressurization During Core Operation (Core Critical) (Cooldown at a Cooling Rate a 100 F/HR)

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3/4 4-30 Figure 4.4.6.1.3-1 Reactor Vessel Material Surveillance Withdrawal Schedule...........

Reactor Steam Dome Program-3/4 4-31 3/4 4-32 3/4.4.7 MAIN STEAM LINE ISOLATIONVALVES 3/4.4.8 STRUCTURAL INTEGRITY..........

3/4.4.9 RESIDUAL'HEATREMOVAL Hot Shutdown Cold Shutdown

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3/4 4-33 3/4 4-34 3/4 4-35 3/4 4-36 4.

EMER EN Y ORE C LIN Y TEM 3/4.5.1 ECCS - OPERATING 3/4.5.2 ECCS - SHUTDOWN 3/4.5.3 SUPPRESSION POOL

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NTAINMENT Y TEM 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity Primary Containment Leakage NINE MILE POINT - UNIT 2

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. 3/4 6-2 Amendment No. 9584

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INDEX E

0 0

S 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal...............................

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Multiple Control Rod Removal.... ~.................

3/4.9.11 RESIDUAL HEAT REMOVAL'ANDCOOLANT CIRCULATION PAAR 3/4 9-12 3/4 9-14 High Water Level..................

Low Watei Level'...........;......

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1 PE IALT EX EPTI N

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3/4.10.1 PRIMARY CONTAINMENTINTEGRITY......

3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM

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3/4 10-2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS 3/4.10.4 RECIRCULATION LOOPS 3/4.10.5 OXYGEN CONCENTRATION 3/4 10-3 3/4 'I0-4 3/4 10-5 3/4 10-7 3/4.10.6 TRAINING STARTUPS....... ~.............................

3/4 10-6 3/4.10.7 SYSTEM LEAKAGE AND HYDROSTATIC TESTING 4.11 RADIOACTIVEEFFLUENT 3/4.1 1.1 LIQUID EFFLUENTS Concentration

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Table 4.11.1-1 Radioactive Liquid Waste Sampling and Analysis Program....................................

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3/4 11-2 ose.........................

D Liquid Radwaste Treatment System.....

Liquid Holdup Tanks 3/4.11.2 GASEOUS EFFLUENTS 3/4 11-5

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D 3/4 11-8 NINE MILE POINT - UNIT 2 xiv Amendment No. 6, 6 84

INDEX A

E F RSE TI NS 4.

4.9 REFUELIN PERA N

~PA E

3/4.9.1, 3/4.9.2 3/4.9.3 3/4.9 4 3/4.9.5 3/4.9.6 3/4.9.7 REACTOR MODE SWITCH INSTRUMENTATION CONTROL ROD POSITION

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....... B3/4 9-1

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DECAY,TIME.....;....

COMMUNICATIONS

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........ B3/4 9-2 REFUELING PLATFORM

........ B3/4 9-2 CRANE TRAVEL-SPENT FUEL STORAGE POOL...................

B3/4 9-2 3/4.9.8 3/4.9.9 3/4.9.1 0 WATER LEVEL - REACTOR VESSEL AND WATER LEVEL-SPENT FUEL STORAGE POOL.....................

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~. B3/4 9-2 CONTROL ROD REMOVAL......:.......:........;........

B3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVALAND COOLANT CIRCULATION..........

B3/4 9-3 4.10 PE IALTE T EX EPTI N

3/4.10.1 PRIMARY CONTAINMENTINTEGRITY.

B3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM..........................

B3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.. ~..................

B3/4.10-1 3/4.10.4 3/4.10.5 3/4.10.6 3/4.10.7 RECIRCULATION LOOPS..................................

B3/4 10-1 OXYGEN CONCENTRATION.............................

B3/4 10-1 TRAININGSTARTUPS....................................

B3/4 10-1 SYSTEM LEAKAGE AND HYDROSTATICTESTING................

B3/4 10-1 4 11 DI A TIVE EFFL ENT 3/4.11.1 LIQUID EFFLUENTS Concentration..........................................

B3/4 11-1 ose

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D B3/4 11-1 Liquid Radwaste Treatment System..............

B3/4 11-2 Liquid Holdup Tanks............;.............. ~..... ~...

B3/411-2 NINE MILE POINT - UNIT 2 Amendment No. 6, 6 84

RVEILLAN E RE ENT 0

4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other.

conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval:

4.0.3 Failure to perform a Surveillance Requirement within the. allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABIL'ITY requirements for a Limiting Condition for Operation.

The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed.

The ACTION requirements may be delayed for,up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Surveillance Requirements'do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.

This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements.

4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:

a.

Inservice testing of ASME Code Class 1,'2, and 3 pumps and valves shall be performed, in accordance with Section XI of the ASME Boiler and Pressure-Vessel Code and applicable addenda is required by 10CFR50.55a(f), except where specific written relief has been granted by the Commission pursuant to.10CFR60.55a(f)(6)(i).

Inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 60.55a(g)(6)(i).

b.

Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable addenda shall be applicable as follows in these Technical Specifications:

4 NINE MILE POINT - UNIT 2 3/4 0-2 Amendment No. 6, 4], 6 84

A '

CIDENT IVIONIT RING IN TR MENTATI N RVEILLANCEREQ IREMENTS TABLE NOTATION

. Excludes'sensors; sensor comparison shall be done in lieu of sensor calibration.

• +

Using sample gas containing:

a.

One volume percent hydrogen, balance nitrogen.

b.

Four volume percent hydrogen, balance nitrogen.

The CHANNEL CALIBRATIONshall consist of position indication verification using the criteria specified. for the Inservice Testing Program.

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The CHANNEL CALIBRATIONshall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

Red, Green or other indication shall'be verified as indicating valve. position'.

NINE MILE POINT - UNIT 2 3/4 3-87 Amendment No. 684

REA TOR OLA EA R

OLANT Y TEM LEAKA E PERATI NAL LEAKA E LIMITN NDITI F

R PERATI N

e.

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With one or more of.the required interlocks shown in Table 3A.3.2-3 inoperable, restore the inoperable interlock to OPERABLE status within 7 days or isolate the affected heat exchanger(s) from the RCIC steam supply by closing and deenergizing heat exchanger valves 2RHS*MOV22A and 2RHS*MOV80A or 2RHS*MOV22B and 2RHS*MOV80B, as appropriate.

With any reactor coolant system leakage greater than the limit in 3A.3.2.e above, identify the source of leakage within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

RVEILLANCE RE UIREMENTS 4.4.3.2.1 The RCS leakage shall be demonstrated to be within each of the above limits by:

a.

Monitoring the primary containment airborne particulate radioactivity at least once per 12

hours, b.

Monitoring the drywell floor drain tank and equipment drain tank fillrate at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, C.

Monitoring the primary containment airborne gaseous radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and d.

Monitoring the reactor'vessel head flange leak detection system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.3.2.2 Each RCS pressure isolation valve specified in Table 3 4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5, using the method and acceptance criteria specified in the Inservice Testing Program, and verifying the leakage of each valve to be within the specified limit:

a.

b.

At least once per 18 months, and Before returning the valve to service following maintenance, repair, or replacement work on the valve.

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.

NINE MILE POINT - UNIT 2 3/4 4-14 Amendment No. Ak 84

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EA LAN M

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4.4 PRE R

EMPERAT RE LI IT A T LANT Y M

LIMITIN NDITI NS F R

PERATI N

3A.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 for hydrostatic or system leakage testing.

Figure 3.4.6.1-2 for heatup by non-nuclear means.

Figure 3.4.6.1-3 for cooldown following a nuclear shutdown and low-power PHYSICS TESTS; and Figures 3.4.6.1-4 and 3.4.6.1-5 for operations with a critical core other than low-power PHYSICS TESTS, with:

a.

A maximum heatup of 100 F in any 1-hour period, b.

A maximum cooldown of 1004F in any 1-hour period, C.

A maximum temperature change of less than or equal to 20 F in any 1-hour period during hydrostatic and system leakage testing operations above the heatup and cooldown limitcurves, and d.

The reactor vessel flange and head flange temperature greater than or equal to 704F when reactor vessel head bolting studs are under tension.

~AT~IN:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains acceptable for continued operations, or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

URVEILLAN E REQUIREMENT 4.4.6.1.1 During system heatup, cooldown, and system leakage and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figures 3.4.6.1-1, 3.4.6.1-2, 3 4.6.1-3, 3A.6.1-4, and 3.4.6.1-5 as applicable, at least once per 30 minutes.

NINE MILE POINT - UNIT 2 3/4 4-24 Amendment No. 684

NINE MILE POINT- 'UNIT 2 NON-CRlTlCAL, HYDROTEST 1400 C7l Ch~ C7 CC I C7 cn. v IC CC O

1200 1000 800 639 600 woo 312 200 NON-CRtTICAI.

OPERATloN MNHvLM TBvPERATURE FOR BOG~

70 F 70 100 0

50 100 5 50 200 250 300 350 DOWNCOMER WATER TEMPERATURE(F)

FIGURE 3.4.6.1-1

, MINIMUMBELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING HYDROSTATIC TESTING AND SYSTEM LEAKAGE TESTING (REACTOR NOT CRITICAL) FOR UP TO 12.8 EFFECTIVE FULL POWER YEARS OF OPERATION NINE MILE POINT - UNIT 2 3/4 4-26 Amendment No. EE 84

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EA T R

LAN EM 4.4.

RE ID AL EAT EM VAL T

H TD WN LIMITING ONDITIONS FOR PERATI N

3.4.9.1 Two* shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be ln operation**, t with each loop consisting of at least:

a.

b.

One OPERABLE RHR pump, and.

One OPERABLE RHR heat exchanger.

0 'N LC tO 3.

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H cut-in permissive setpoint..

~CTIOM:

a.

With less than the above required RHR shutdown cooling mode loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible, Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.t t b.

With no RHR shutdown cooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

SURVEILLANCE REQUIREMENTS 4.4.9.1 At least one shutdown cooling mode loop of the residual heat removal system or alternative method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Ok One RHR shutdown cooling mode loop may be inoperable 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 provided the other loop is OPERABLE and in operation.

The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8-hour period provided the other loop is OPERABLE.

The RHR shutdown cooling mode loop may be removed from operation during hydrostatic and system leakage testing.

Whenever two or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat-removal methods.

NINE MILE POINT - UNIT 2 3/4 4-35 Amendment No. 684

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REA T R

LAN RESIDUAL HEAT REMOVAL 0~

LD H TD WN LIMITIN NDITI N F

R PERATI N

3A.9.2 Two* shutdown co'oling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation** t with each loop consisting of at least:

a.

b.

One OPERABLE RHR pump, and One OPERABLE RHR heat exchanger.

0 ATIO A CO OM 4.

~CT~IN:

a.

P ll With less than the above required RHR shutdown cooling mode loops OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at. least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop.

b.

With no RHR shutdown cooling mode loop in operation, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish reactor coolant circulation by an alternate method and monitor reactor coolant temperature and pressure at least once per hour.

RVEILLANCE RE UIREMENT 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or alternative method shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

One RHR shutdown cooling mode loop may be inoperable 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 provided the other loop is OPERABLE and in operation.

The shutdown cooling pump may be removed, from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> every 8-hour period provided the other loop is OPERABLE.

The shutdown cooling mode loop may be removed from operation during hydrostatic and system leakage testing.

NINE MILE POINT - UNIT 2 3/4 4-36 Amendment No. Ek 84

II

LET I L

WE JltllSC&

R E

PER TN VEILLAN E RE IREMEN 4.8.1.1.2.e (Continued) 12.

Verifying that the automatic load timer relays are OPERABLE with the interval

'between each load block within a 10% of its design interval for diesel generators EDG*1 and EDG*3.

13.

Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:

a) For Divisions I and II, turning gear engaged and emergency stop.

b) For Division III, engine in the maintenance mode and diesel generator lockout.

At least once per 18 months verify each diesel generator. starts and accelerates to at least 600 RPM within 10 seconds for EDG*1 and EDG*3, and 870 RPM within 10 seconds for EDG*2. The generator voltage and frequency for EDG*1 and EDG*3 shall be 4160 a 416 volts and 60 a 3.0 Hz within 10 seconds and 4160 a 416 volts and 60 a 1.2 Hz within 13 seconds after the start signaI.

The generator voltage and frequency for EDG*2 shall be 4160 a 416 volts and 60 a 1.2 Hz within 15 seconds after the start signal

~ This test shall be performed within 5 minutes of shutting down the diesel generator after the diesel generator has operated for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 4400 kW or more for EDG*1 and EDG*3 and 2600 kW or more for EDG*2. For any start of a diesel, the diesel must be loaded in accordance with manufacturer's recommendations.

Momentary transients due to changing bus loads shall not invalidate this test.

g h.

-At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all three diesel generators simultaneously, during shutdown, and verifying that all diesel generators EDG*1 and EDG*3 accelerate to at least 600 rpm and EDG*2 accelerates to at least 870 rpm in less than or equal to 10 seconds.

At least once per 10 years by:

1. Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, and

2. Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section XI Article IWD-5000.

I 4.8.1.1.3 All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2, within 30 days.

Reports of diesel generator failures shall include the information recommended in Position C.3.b of RG 1.108, Revision 1, August 1977.

If the number of failures in the last 100 valid tests, on a per nuclear unit basis, is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Position C.3.b of RG 1.108, Revision 1, August 1977.

NINE MILE POINT - UNIT 2 3/4 8-11 Amendment No. IM 84

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PE I LTE TEX I

3 4.10.7 YSTEM LEAKAGE AND HYDR STATI TE TIN h

IMITIN NDI N F R

PERATI N

..",. f 3.10.7 When conducting system leakage or hydrostatic testing, the average reactor coolant I

temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased above 200'F, and operation considered not to be in OPERATIONAL CONDITION 3, to allow performance of a system leakage or hydrostatic test provided the maximum reactor coolant temperature does I

not exceed 212'F and the following OPERATIONAL CONDITION 3 LCO's are met:

a.

3.3.2, "Isolation Actuation Instrumentation", Functions 1.a.2, 1.b, and 3.a and b of Table 3.3.2-1; b.

C.

3.6.5.1, "Secondary Containment Integrity";

3.6.5.2, "Secondary Containment Automatic Isolation Dampers"; and d.

3.6.5.3, "Standby Gas Treatment System."

APPLICABILITY: OPERATIONAL CONDITION 4, with average reactor coolant temperature 200oF

~A~TI N:

With the requirements of the above specification not satisfied, immediately enter the applicable condition of the affected specification or immediately suspend activities that could increase the average reactor coolant temper'ature or pressure and reduce the average reactor coolant

'emperature to a200'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLAN E REQUIREMENTS 4.10.7 Verify applicable OPERATIONAL CONDITION 3 surveillances for specifications listed in 3.10.7 are met.

NINE MILE POINT - UNIT 2 3/4 10-7.

Amendment No. 6 84

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REA T R

LAN BA E

4.4.

PRE URE EMPERAT RE LIMIT All components in the reactor coolant system are designed to withstand the effects of cyclic loads from temperature and pressure changes in the system.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The operating limitcurves of Figures 3A.6.1-1 through 3.4.6.1-5 are derived from the fracture toughness requirements of 10CFR50, Appendix G, and ASME Code Section III, Appendix G. The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components.

Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Subsection 5.3.1.5, "Fracture Toughness."

The reactor vessel materials have been tested'to determine their initial RTNDT. The results of these tests are shown in Bases Table B3/4.4.6-1.

Reactor operation and resultant fast neutron (E

greater than 1 MeV) irradiation willcause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and nickel content of the material can be predicted using Bases Figure B3/4.4.6-1 and the recommendations of RG 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating irradiated specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the specimens and vessel inside radius are essentially identical, the irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figures 3.4.6.1-1 through 3.4.6.1-5 shall be adjusted, as required, on the basis of the specimen data and recommendations of RG 1.99, Revision 2. Data obtained after removal of the first surveillance capsule will be used to adjust the fluence of Bases Figure B3/4.4,6-1.

The pressure-temperature limit lines shown in Figures 3A.6.1-1 through 3.4.6.1-5 for hydrostatic testing and system leakage testing for critical operations have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50.

The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are provided in Table 4.4.6.1.3-1 to assure compliance with the requirements of Appendix H to 10CFR50.

NINE MILE POINT - UNIT 2 B3/4 4-5 AMENDMENTNO. 6 84

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4 1 PECIAL T EPTIONS BA E

4.1 1

PRIMARY ONTAINMENTINTE RI The requirement for PRIMARY CONTAINMENTINTEGRITY is not applicable during the period

.when open vessel tests are being performed during the low-power PHYSICS TESTS.

4'1,2 D

E EN E

TR Y

In order to perform the tests required in the Technical Specifications it is necessary to bypass the sequence restraints on control rod movement.

The additional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.

4.1 H TD WN MAR IN DEM N TRA I

N Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur.

These additional restrictions are specified in this Limiting Condition for Operation.

4.10.4 RE IRCULATION L OP This special test exception permits reactor criticality under no-flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

4.1 OXYGEN ON ENTRATI N Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment duiing,the initial startup and testing phase of operation.

Without this access, the startup and test program could be restricted and delayed.

4.1

.6 TRAINING TAR P

This special test exception permits training startups to Pe performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize the discharge of contaminated water to the radioactive waste disposal system.

4.1

.7 Y TEM LEAKA E ND HYDR TATI TE TIN This special test exception allows reactor vessel system leakage and hydrostatic testing to be performed in OPERATIONAL CONDITION 4 with the maximum reactor coolant temperature not exceeding 212~F.

The additionally imposed OPERATIONAL CONDITION 3 requirement for secondary containment operability provides conservatism in the response of the unit to an operational event.

This allows flexibilitysince temperatures approach 190'F during the testing and can drift higher because of decay and mechanical heat.

Additionally, because reactor vessel fluence increases over time, this testing will require coolant temperatures

>200'F.

NINE MILE POINT - UNIT 2 B3/4 10-1 Amendment No. 6, h@4

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EA T R

Y LI RTRAN IENTLIMT ANDDE I

N Y LE RT AN IENT YCLIC R TRAN IENT LIIVIIT'20 heatup and cooldown cycles 80 step change cycles 198 reactor trip cycles

$ 30 hydrostatic and system feakage tests DESI N

Y LE R TRANSIENT 70'F to 565'F to 70'F Loss of feedwater heaters 100% to 0% of RATED THERMAL POWER Pressurized to a930 psig and a1250 psig NINE MILE POINT - UNIT 2 5-9 Amendment No. 84

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