ML20202G375

Proposed Tech Specs Pages,Reflecting References to 10CFR50.55a(f) & (G) as Well as Terminology Used in Second Ten Year Isi/Ist,Beginning on 980405

ML20202G375
Person / Time
Site:	Nine Mile Point
Issue date:	02/05/1998
From:	NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML20202G346	List: ML20202G342 ML20202G364 ML20202G375
References
NUDOCS 9802200115
Download: ML20202G375 (33)

Text

ATTACHMENT A

', NIAGARA MOHAWK POWER CORPORATION

. LICENSE NO. NPF 69 DOCKET NO. 50 410 finanand Channes to Technical Specifications Replace the existing pages with the attached revised pages as listed below. The page.s

, have been retyped in their entirety with marginal markings to indicate the changes.

i EB1811DR.Etats Revised Panes i X X l xiv xiv l

XX XX 3/4 0 2 3/4 0 2 3/4387 3/4 3-87 3/4414 3/4 4-14 3/4424 3/4424 3/4426 3/4426 3/4435 3/4435 3/4436 3/4436 3/4811 3/4811 3/4107 3/4107 B3/4 4 5 B3/4 4 5 B3/410-1 B3/4101 59 59 I

9802200115 900205 0 PDR ADOCK 0500 P

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

4  % INDEX

( LIMITING CONDITIONS FOR OPERATION AND SURVElLLANCE REQUIREMENTS PAGE REACTOR COOLANT SY51EM (Continued) 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 24 Figure 3.4.6.1 1 Minimum Beltline Downcomer Water Temperature for Pressurization During Hydrcstatic Testing and System Leakage Testing (Reactor Not Critical) . . . . . . . . . . . . 3/4426 Figure 3.4.6.12 Minimum Beltline Downcomer Water Temperature for Pressurization During Heatup and Low Power Physics Tests (Reactor Not Critical)(Heating Rate s 10 0 F/H R ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4427 Figure 3.4.6.13 Minimum Beltline Downcomer Water Temperature for Pressurization During Cooldown and Low-Power Physics Tests (Reactor Not Critical)

(Cooling Rate s 100 F/HR) . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4 2 8 Figure 3.4.6.14 Minimum Beltline Downcomer Water Temperature for Pressurization During Core Operation (Core Critical) (Heatup at a Heating Rate s 100 F/HR) . . . . . . . . . . . 3/4429 Figure 3.4,6.15 Minimum Beltline Downcomer Water Temperature for Pressurization During Core Operation (Core Critical)(Cooldown at a Cooling Rate s 100 F/HR)......................................... 3/4430 Figure 4.4.6.1.3 1 Reactor Vessel Material Surveillance Program -

Withdrawal Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 31 Reactor Steam Dome . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 3 2 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 33 3/4.4.8 STRUCTUR AL INTEGRITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4 34 3/4.4.9 RESIDUAL HEAT REMOVAL H ot S hu t d o w n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4 3 5 Cold Shutdown ............................................ 3/4436 3/4.5 EMERGENCY CORE COOLING SYSTEMS

.3/4.5.1 ECCS - OPERATING ....... ................................. 3/4 5 1 3 /4. 5. 2 E C C S - S H UTDOWN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5 7 3/4.5.3 SUPPRESSION POOL ........................................ 3/4 5 9 3/4.6 CONTAINMENT SYSTEMS I

3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity . . . . . , . . . . . . . . . . ................. 5/4 6-1 Primary Containment Leakage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6 2 NINE MILE POINT UNIT 2 x Amendment No, dd

". INDEX i

. 1 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

. PAGE 3/4.9.10 CONTROL ROD REMOVAL Single Control Rod Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9 12 Multiple Contiol Rod Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9 14 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION Hig h Water Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 9 16 Low Wa ter Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9 17 3/4.10 SPECIAL TEXT EXCEPTIONS >

3/4.10.1 PRIMARY CONTAINMENT INTEGRITY , . . . . . . . . . . . . . . . . . . . . . . . . 3/4101 ,

3/4.10.2 ROD SEQUENCE CONTROL SYSTEM . . . . . . . . . . . . . . . . . . , . . . . . . . 3/410 2 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS ......... ............ 3/4103 i I

3/4.10.4 R ECI R C U L ATI O N LO O PS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 10 4 3/4.10.5 O XY G E N C O N C E NT R ATIO N , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-5 3/4.10.6 TR AINING STARTUPS . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10 6 3/4.10.7 SYSTEM LEAKAGE AND HYDROSTATIC TESTING . . . . . . . . . . . . . . , , 3/410-7 l 3/4.11 RADIOACTIVE EFFLUERIE  ;

i 3/4.11.1 LIQUID EFFLUENTS Concentration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 1 1 Table 4.11.1 1 Radioactive L .sid Waste Sampling and Analysis 1 Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 1 2 Dose......................................... . . . . . . . 3 /4 1 1 - 5 Liquid Radwaste Treatment System . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/411 6 Liquid Ho: dup Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 1 7 L 3/4.11.2 GASEOUS EFFLUENTS D o se l a t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 1 8 i

NINE MILE POINT - UNIT 2 xiv - Amendment No. //, d$

1

. . INDEX BASES FOR SECTIONS 3.0/4.0 PAGE

. 3/4.9 REcUELING OPERATIONS 3/4.9.1 R EACTO R MO D E SWITCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 9 1 3/4.9.2 INSTRUMENTATION ...................................... B3/4 9 1 3/4.9.3 CONTROL ROD POSITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 9 1 3/4.9.4 D E C AY TI M E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 9 2 3/4.9.5 C O M M U N I C ATI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 9 2 3/4.9.6 RE FUELIN G PLATFO RM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. B3/4 9 2 3/4.9.7 CRANE TRAVEL SPENT FUEL STORAGE POOL , . . . . . . . . . . . . . . . . . 83/4 9 2 3/4.9.8 WATER LEVEL REACTOR VESSEL AND WATER LEVEL -

3/4.9.9 SPENT FUEL STORAGE POOL , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 9 2 3/4.9.10 CO NTROL RO D R EMOVAL . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . B3/4 9 2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION .......... B3// 9 3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY COf JTAINMENT INTEGRITY . . . . . . . . . . . . . . . . . . . . . . . . . B3/4101 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . B3/4101 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS . . . . . . . . . . . . . . . . . . . . . B 3 /4.10- 1 3/4.10.4 RECIRCULe DN LOOPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3 /4 1 0 1 3/4.10.5 OXYG EN CONCENTR ATION , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 10 1 3/4.10.6 TR AINING STARTUPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 10-1 3/4.10.7 SYSTEM LEAKAGE AND HYDROSTATIC TESTING . . . . . . . . . . . . . . . . B3/410-1 l 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS C onc entration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... B3/411 1 D o s e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1 1 - 1 Liquid Radwaste Treatment System . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 11 -2 Liquid Holdup Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 1 1 2 NINE MILE POINT - UNIT 2 xx Amendment No. //, dd

EURVEILLANCE REQUIREMENTS 4.0.1 Survr.illance Requirements shall be met during the OPERATIONAL CONDITIONS or other condithns 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 Faiiare to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requiremergs for a Limiting Condition for Operation. The timo 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 where the allowable outage time limits of the ACTION requirements are less than 24 l hours. Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 En;ry into an OPERATIONAL CONDITION or other specified applicable condition shall not l be modo unless the Surveillance Requirement (s) associated with the Limiting Condition for l Operation have been performed within the applicable surveillance interval or as otherwise i

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

4.0.5 Ourveillance 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 as required by 10CFR50.55a(f), except where specific written relief has been granted by the Commission pursuant to 10CFR50.55a(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 50.55a(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:

i NINE MILE POINT UNIT 2 3/4 0 2 Amendment No. /d, dd, dd

TABLE 4.3.7.5.1 lCcntinued)

, AC_CIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMEf!TS TABLE NOTATIONS

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

• • Us!ng sample gas containing:

a. One volume percent hydrogen, balance nitrogen,

b. Four volume percent hydrogen, balance nitrogen.

The CHANNEL CAllBRATION shall consist of position indication verification using the criteria specified for the Inservice Testing Program, t The CHANNEL CAllBRATION shall 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, tt Red, Green or other indication shall be verified as indicating valve r,osition.

i NINE MILE POINT - UNIT 2 3/4387 Amendment No, dd I

REACTOR COOLANT SYSTEM

• i REACTOR COOLANT SYSTEM LEAKAGE QEERATIONAL LEAKAGE LIMITING CONDITIONS FOR OPERATION

e. With one or more of the required interlocks shown in Table 3.4.3.2 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*MOV228 and 2RHS*MOV808, as appropriate.

f. With any reactor coolant system leakage greater than the limit in 3.4.3.2.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 />.

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The RCS leakage shall be demonstrated to be within each of the above limits by:

a. Monitoring the primary containmen, .rborne particulate 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 />,

b. Monitoring the drywell floor drain tank and equipment drain tank fill rate 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 3.4.3.21 shall be demonstrated -

OPERABLE by leak testine wrsuant to Specification 4.0.5, using the method and acceptance criteria specified in the ins le Testing Program, and verifying the leakage of each valve to De within the specified limit:

a. At least once per 18 months, and

b. Before returning the valve to servica following maintenance, repair, or replacement work on the valve.

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

NINE Mll.E POINT UNIT 2 3/4414 Amendment No. //

I REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM i

LIMITING CONDITIONS COR OPERATION- '

3.4.6.1 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 l

-3.4.6.12 for heatup by non-nuclear means. Figure 3.4.6.13 for cooldown following a nuclear shutdown and 'ow power PHYSICS TESTS: and Figures 3.4.6.14 and 3.4.6.15 for operations with a critical core other than low power PHYSICS TESTS, with:

a. A maximum lieatup of 100*F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

b. A maximum cooldown of 100*F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

c. A maximum temperature chanbe of less than os equal to 20 F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period during hydrostatic and system leakage testing operations above the heatup and l cooldown limit curves, and

d. The reactor vessel flange and head flange temperature greater than or equal to 70 F when reactor vessel head botting studs are under tension.

APPLICABILITY: At all times.

ACTION:

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-iimit condition on the structuralintegrity of the reactor coolant system; determine that the reactor coolant system remains accr'eble 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 - 90WN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.6.1.1 Durin;) system heatup, cooldown, and system leakage and hydrostatic testing l operations, the reactor coolant system temperature and pressure shall be determined to be within the aoove 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, 3.4.6.1-4, and 3.4.6.1 5 as applicable, at least once per 30 minutes.

l NINE MILE POINT - UNIT 2 3/4424 Amendment No. dd

NINE MILE POINT UNIT 2 NON-CRITICAL HYDROTEST 1400 1200

.~.

.P w l $, 1000 r ,

5o m P' 800

===- NON-CRITICAL "I 639 OPERATION a 600 as 1

n6 03 D y 400 mI G

312 MNuJM Te m Tugt 200 .. pon ecx.np 70 F 0 79 3co 0 50 100 150 200 250 300 350 DOWNCOMER WATER TEMPERATURE (P)

FIGURE 3.4.6.1 1 MINIMUM BELTLINE 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 3/4426 Amendment No. Id

REACTOR COOLANT SYSTEM 3/4.4.9 RESIDUAL HEAT REMOVAL HOT SHUTDOWR LIMITING CONDITIONS FOR OPERATION _

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 in operation **, t with each loop consisting of at least:

a. One OPERABLE RHR pump, and

b. One OPERABLE RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut in permissive setpoint.

ACTION:

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 altemate 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 retum 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 attemate 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 attemative 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 RH81 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 suveillance 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.

t The RHR shutdown cooling mode loop may be removed from operation during hydrostatic and system leakage testing. l tt 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 3/4435 Amendment No. dd I

REACTOR COOLANT SYSTEM RESIDUAL HEAT REMOVAL COLD SHUTDOWN

.. LIMITING CONDITIONS FOR OPERATION l '3.4.9.2 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 in operation ** t with each loop consisting of at least:

a. One OPERABLE P.HR pump, and l

b. One OPERABLE RHR heat exchanger.

APPLICABILITY: OPERATIONAL CONDITION 4.

ACTION:

a. With less than the above required AHR 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.

SURVEILLANCE REQUIREMENTS 4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system or Elternative method shell 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 />.

s

• 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.-

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

NINE MILE POINT UNIT 2 3/4436 Amendment No. Id

• S ELECTRICAL POWER SYSTEMS AC SOURCES AC SOURCES - OPERATING SURVEILLANCE REQUIREMENTS 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 i 10% of its design interval for diesel generators EDG*1 and EDG*3.

13. Verifying that the following diesel generator lockout features provent diesel generator starting only when required:

a) For Divisions I and ll, turning gear engaged and emergency stop, b) For Division ill, engine in the maintenance mode and diesel generator lockout.

f. 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 i 416 volts and 6013.0 Hz withire 10 seconds and 41001416 volts and 60 i 1.2 Hz within 13 seconds after the start signal. The generator voltage and frequency for EDG*2 shall be 41601416 volts and 60 i 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 duo to changing bus loads shall not invalidate this test,

g. 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 equai to 10 seconds.

h. At least once per 10 years by:

1. Draining each fuel oil storage tank, removing the accumu' sed 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 lil, subsection ND of the ASME Code in accordance with ASME Code Section XI Article IWD 5000. l 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 diese! generator failures shallinclude 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 3/4811 Amendment No, dd

SPECIAL TEST EXCEL 2TIONS 3/4.10.7 SYSTEM LEAKAGE AND HYDROSTATIO TESTING l LIMITING COND! TION FOR OPERATION . 3 3.10.7 When conducting system leakage or hydrostatic testing, the average nactor coolant l temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased above 200 F, and operation considered not to be in OPERATICNAL CONDITION 3, to allow performance of a system leakage or hydrostatic test provided the maxims.m reactor coolant temperature does l not exceed 2 F and the following OPERATIONAL CONDITON 3 LCO's are met:

a. 3.3.2, " Isolation Actuation Instrumentation", Functiens 1.a.2,1.b, and 3.a and b of

) Table 3.3.21;

b. 3.6.5.1, " Secondary Containment integrity";

c. 3.6.5.2, " Secondary Containment Automatic Ist _ .pers"; and

d. 3.6.5.3, " Standby Gas Treatment System."

APPLICABILITY: OPERATIONAL CONT:' TIC' . with average reactor coolant temperature

> 200*F.

ACTION:

With the requirements of the above specification not satisfied,immediately enter :he applic able

, condition of the affected specification or immediately suspend activities that could increase the averago reactor coolant teii.parature or pressure and reduco the as arage reactor coolant temperature to s200 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS

)

\

4.10.7 Wrify applicable OPERATIONAL CONDITION 3 surveillances for specifications listed in 3.10.7 are met, j NINE MILE POINT - UNIT 2 3/4 10-7 Amendment No. d$

q _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

REACTOR COOLANT SYSTEM BASES 2&L6 PRESSURE / TEMPERATURE LIMITS i All components in the reactor coolant system are designed to withstand the effects of cyclic loads from temperature and pressure changes ir, 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 teatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The operating limit curvas of Figures 3.4.6.1-1 through 3.4.6.* -5 are derived frcm the fractu e toughness requirements of 10CFR50, Appendix G, and ASMc Code Section lit, Appendix G. The curves are based on the RTNDT and stress intensity factL ;rmation 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 han been tested to dete mine their initial RTNDT. The results of these teste are shown in Bases Table B3/4.4.61. Retctor operation and resultant faat neutron (E greater than 1 McV) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperatJre, based upon the fluence, sepper content, and nickel content of the material can be predicted using Bases Figuro B3/4.4.6-1 and the recommendations of RG 1.99, Revision 2,

" Radiation Embrittlernent of Reactor Vessel Materials."

The actual shift in RTNDT of the vessel material will be established periodicany during operation by removing and evaluating irradiated specimens installed near the inside wall of the reactor vesselin the core area. Since the neutron spectra u the specimens arid vesselinside radius are essentially identical, the irradiated specimens can b3 used with confidence in predicting reactor vessel material transition temperature shift. The operating limit cwves of Figures 3.4.6.1-1 through 3.4.6.1-5 shell 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 3.4.6.1-1 through 3.4.6.15 for hydrostatic testing end system leakage testing for critical operations have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50.

The umber of reactor vessel irradiation surveillance capsules and the frequencies for removing i and ;est5g the spacimens in these capsules are provided in Table 4.4.6.1.3 1 to assure complicace with the requirements of Appendix H to 10CFR50.

NINE MILE POIN'i - UNIT 2 B3/4 4-5 AMENDMENT NO. did

& 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 PRIMARY CON TAINMENT INTEGRITY l

The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low-power PHYSICS TESTS.

3/4.10.2 ROD SEQUENCE CONTROL SYSTEM in order to perfctm the tests requi ed in the Technical Specifications it is necessary to bypass t7e sequence restraints on control rod movement. The additional surveillance requirements ensure that the specifications on heat generation rates and shutdov.n margin requiraments are not exceedad during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety anclysis.

3/4.10.3 SHUTDOWN MARGIN DEMONSTR ATIONS 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 Lir .iting Condition for Operation.

3/4.10.4 RECIRCULATION LOOPS 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.

3/4.10.5 OWGEN CONCENTRATICE Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase of operation. Without this access, the startup and test program could % r65tricted and delayed.

3/4.10.6 TR AINING STARTUPS l

This special tes: exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperaturc while centrolling RCS tamperature with one RHR subsystem aligned in the shutdown ecoling modo in o:dar to minimize the discharge of contaminated water to the radiocctive waste disposal su cem.

3/4.10.7 SYSTEM LEAKAGE AND HYDROSTATIC TESTjjEi l This cpecial test exception allows reactor vessel system leakage and hydrostatic testing to be I nerformed in OPERATIONAL CONDITION 4 with the maximum reactor coolant temperature not

.xceeding 212 F. The additionally imposed OPERATIONAL LONDITION 3 requirement for secondary containment operability provides conservatism in the respr.,nse of the unit to an operational event. This allows flexibility since temperatures appr. .ch 190 F during the testing and can drift higher because of decay and mechanical lieat. Adduionally, because reactor veseel fluence increases over ti.ne, this testing will require coolant temperatures > 200oF.

NINE MILE POINT - UNIT 2 B3/410-1 Amenhent No. d[ dd

l l*V TABLE 5.7.1 1

, BEACTOR CYCLIC OR TRANSIENT LIMITS AND E'ESIGN CYCLE OR TRANSIENT CYCLIC OR TRANSIENT LIMIT DISlRN CYCLE OR TRANSIENT l

120 heatup and cocidown cycles 70*F to 565'F to 70oF 80 step change cycles Loss of feedwater heaters 198 reactor trip cycles 100% to 0% of RATED THERM AL POWER h 130 hydrostatic and system leakage tests Pressurized to 2930 psig and s1250 psig l t

4 t

NINE MILE POINT - UNIT 2 5-9 Amendment No.

+ ATTACHMENT B *

, NIAGARA MOHAWK POWER CORPORATION LICENSE NO. WPF 69 DOCKET NO. 50-410 Suoourtino information and No Sionificant Hazards ConsiderPtigp Analysis INTRODUCTION The proposed Nine Mile Point Unit 2 (NMP2) Technical Specification (TS) changes contained herein present revisions to the following NMP2 TS pages:

Section 4.0.5, Surveillance Requirements for inservice inspection and testing; Table 4.3.7.5-1, " Accident Manitoring instrumentation Surveillance Requirements"; Section 4.4.3.2.2, Reactor Coolant System Leakage surveillance requirements: Sections 3.4.6.1 and 4.4.6.1.1, Reactor Coolant System Prossure/ Temperature Limits: Figure 3.4.6.1 1, Nine Mile Point Unit 2 non-critical hydrotest; Section 3.4.9.1, Residual Heat Removal Hot Shutdown; Section 3.4.9.2 'iesidual Heat Removal Cold Shutdown; Section 4.8.1.1.2.h.2, Electrical Power Systerns ~ 'ources - Operating; Section 3/4.10.7, System Leakage and Hydrostatic Testing; Bases b/4.4.6, Pressure / Temperature Limits; Bases 3/4.10.7, System Leakage and Hydror atic Testing: Table 5.7.1-1, " Reactor Cyclic or Transient Limits and DLsign Cycle or Transient."

10CFR50.55a requires that the inservice inspections (ISI) and inservice tests (IST) conducted during successive 10-yeu intervals comply with the requirements in the latest edition and addenda of Section XI of the ASVi Beiler and Pressure Vessel Code that was in effect twelve months prior to the start of the 10-year interval. NMP2 will begin its second 10-year ISI/IST interval on April 5,1998. The effective edition of the ASME Code for this interval is the 1989 edition. Some terminology used in the present program is being changed in the second 10 year ISI/IST Program Plan to reflect the terminology changes from the 1983 edition of the ASME Code, which was used in the current interval, ,

to the 1989 edition. These changes do not represent a change to the inspections or tests performed under the ISI/IST Program. The proposed changes will ensure that TS reflect

} the correct 10CFR references and the terminology of the second NMP210-year ISI/IST prograrn.

Niagara Mohawk Power Corporation requests approval of this Application for Amendment for implementation on April 5,1998.

EVALUATION 10CFR50.55a requires that the inservice inspections and inservice tests cenducted during successive 10-year intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Boiler and Pressure Vessel Code that was in effect twelve months prior to the start of the 10-year interval. NMP2 will begin its second 10-year ISI/IST interval on April 5,1998. The effective adition of the ASME Code for this interval is the 1989 edition. Some terminology used in the present p ogram is being changed in the st.cond 10-year ISI/IST Program Plan to reflect the terminology changes Page 1 of 3

from tha 1983 edition of the ASME Code, which was used in the current interval, to the 1989 edition. These changes do not represent a change to the inspections or tests

, performed under the ISl/IST Program. The proposed revisions will ensure that TS reflect the 10CFR references and the terminology of the second NMP210-year ISI/IST program.

' This program will be written to address 10CFR50.55a(f) (for the IST program) and 10CFR50.55a(g) (for the ISI program!. This change in the applicable 10CFR sections is raflected in the proposed revision to Section 4.0.5.

The proposed revisions to Table 4.3.7.51 and Section 4.4.3.2.2 replace references to ASME Section XI with references to criteria in the IST Program. The IST Program is based on Section XI of the 1989 edition of the ASME Code, which hr.; been approved by the NRC.

The proposed revisions to Sections 3.4.9.1 and 3.4.9.2 add the phrase " system leakage" to notes that identify testing conditions when the shutdown cooling mode loop may be removed from service. This addition is not changing a condition for shutdown cooling, it is incorporating additional testing flexibility allowed by the Code. A Section XI code case of the 1989 edition of the ASME Code allcws the flexibility to perform either a hydrostatic test or a system leakage test. The test conditions and frequency are not being changed by this proposed revision.

l The proposed revision to Section 4.8.1.1.2.h.2 corrects a typographical error. This section

! references ASME Code Section ll;it should read Section XI.

The remaining proposed revisions are editorial changes to make the references to hydrostatic testing and leak testing conform to the terminology that will be used in the second 10-year ISI/IST program.

CONCLUSIONE The proposed changes will revise sections of the NMP2 TS to ensure that the TS reflect the references to 10CFR50.55a(f) and (g) as well as the terminology used in the NMP210-year ISI/IST program. These changes to the TS will not affect the inspections or tests performed under the ISI/IST program.

ANALYSIS No Slanificant Hazards Consideration l

10CFR50.91 requires that at the time a licensee requests an amendment, it must provide to the Commission its analysis using the standards in 10CFR50.92 concerning the issue of no significant hazards consideration. Therefore, in accordance with 10CFR50.91, the following analyses have been performed with respect to the requested change:

Ihe operation of Nine Mile Point Unit 2. In accordance wlth the orocosed amendment, will ag_tinvolve a sianificant increase in the orobability or conseauences of an accident Draviousiv evaluated.

The changes to the TS will ensure that TS reflect the correct 10CFR references and the terminology of the second NMP210-year ISI/IST program. The proposed revisions replace references to ASME Saction XI with references to criteria in the inservice Testing Program.

The performance of system leakage testing is added to notes that identify conditions when Page 2 of 3

th3 shutdown cooling mode loop may be removed from service. The other changas are cditorial changes only to ensure that TS refisct the sncond 10-year ISl/IST program. One

, of the changes corrects a typographical error. These proposed changes do not affect the inspections or tests performed under the ISI/IST Program and will not result in any changes to the plant. None of the precursors of previously evaluated accidents are affected and therefore, the probability of an accident previously evaluated is not incraased.

The changes will not affect the safety function of any equipment covered by the ISI/IST program. Therefore, these changes will not involve e significant increase in the consequences of an accident previously evaluated.

The operation of Mine Mile Point Unit 2. In a.ccordance with the oronosed amendrnent, will not create the possibility of a new or different kind of accident from snv accident previously evaluated.

The changes to the TS will ensure that TS reflect the correct 10CFR refercnces and the terminology of the second NMP210-year ISI/IST program. One of the changes corrects a typographical error. No physical modification of the plant is involved and no changes to the methods in "/b'ch plant systema are operated arc required. These changes do not affect the inspections or tests performed under the ISl/IST Program. The changes do not introduce any new failure modes or conditions that may create a new or different accident.

Therefore, the changes do nct by themselves create the possibility of a new or different kind of accident previously evaluated.

The operation of N!ne Miie Point Unit 2. In accordance with the proposed amendment, will not involve a sinnificant reduction in a marain of safety.

The changes to the T3 will ensure that TS reflect the correct 10CFR references and the terminology of the second NMP210-year ISI/IST program. One of the changes corrects a typographical error. No physical modification of the plant is involved and no changes to g the methods in which plant systems are operated are required. The changes do not adversel/ effect any physical barrier to the release of radiation to plant personnel or to the public. Thase changes do not affect the inspections or tests performed Lnder the ISI/lST Program. Therefore, these changes do not involve a reduction in a margin of safety.

Accordingly, the proposed changes do not present a significant hazards consideration.

Page 3 of 3

ATTACHMENT C NIAGARA MOHAWK POWER CORPORATION LICENSE NO. NPF-69 ,

DOCKET NO. 50 410 Marked Cooy of Proposed Chennes to Current Technical Specifications The current version of pages listed belew of the NMP2 Technical Specifications have been hand marked-up to reflect the proposed changes.

2 61arked-un Paaes X

xiv XX 3/4 0 2 3/4387 3/4414 l 3/4 4-24 r 3/4 4-26 3/4435 3/4436 3/4 8-11 3/4 10-7 B3/4 4-5 B3/4 10-1 5-9

INDEX t!MITfMC c0NottteNs rop cort > TION awo SURvttttaNet troufREut1Ts p

. C E

REACTOR COOLANT. SYSTEM (Continued) 3/4.4.6 PRESSURE / TEMPER /TURE LIMITS Reactor Coolant System........................................ 3/4 4-24 ,/

r Figure 3.4.6.1 1 Minimum Beltline Downcomer water Temeerature for Pressurization During L ,m ..o Hydrostatic Testing and 4ew=- Testing (Reactor Not Critical) . . . .. . 3/4 4 26 H u s ost LLerr%E .

Figure 3.4.6.1-2 Minimum Beltline Downcomer Water Temperature for Pressurization During heatup and Low Power Physics Tests (Reactor Not Critical) (Heating Rate i 100 F/HR)..................................... 3/4 4-21 Figure 3.4.6.1 3 Minimum Beltline Downcomer Water Temocrature for Pressurization Ouring Cooldown and Low-Power Physics Tests (Reactor Not Critical) (Cooling Rate i 100 F/HR)..................................... 3/44-2d Figure 3.4.6.1-4 Minimum Beltline Downcomer Water Temperature for Pressurization During Core Operation (Core Critical) (Heatup at a Heating Rate i 100 F/HR)...... 3/4 4-29 Figure 3.4.6.1-5 Minimum Beltline Downcomer Water Temperature for Pressurization During Core Operation (Core l ' Critical) (Cooldown at a Cooling Rate i 100 F/HR) . . . 3/4 4-3C Table 4.4.6.1.3-1 Reactor Vessel Material Surveillance Program -

Withdrawal Schedule................................. 3/4 4-31

\

Reactor Steam 0ome............................................ 3/4 4-32' I

3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.............................. 3/4 4-33!

3/4.4.8 STRUCTURAL INTEGRITY........................ . . . . . . . . . . . . . . . . . ' ' ' 4- 34 '

3/4.4.9 RESIOUAL HEAT REMOVAL Hot Shutdown.................................................. 3/4 4-35, Cold Shutdown................................................. 3/4 4-36\

3/4.3 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING.............. ............................... 3/4 5-1 3/4.5.2 ECCS - SHUTD0MN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 5-7 3/4.5.3 SUPPRES$10N P00L.............................................. 3/4 5-9 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Cont?inment Integrity................................. 3/4 6-1 Primary Contai nment Le aitage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 6-2 s

i NINE. MILE POINT - UNIT 2 x AmendmentNo.g

7 -. -

4 th uM[ TING CONDf710Ng poM OpsRATION AND SURV8811 ANCE REautREMENTE PAGE 3/4.9.10 CON 1MOL ROD REMOVAL l Single Centrol Rod Removal . . . . . . . . . . . . . . .*. . . . . . . . . . . . . . . SM 9-12 Muhiple control Rod Memoval . . . . . . .. . . . . . . . . . . . . . . . . . . . . . SM 614 3M.S.11 RESIDUAL MEAT REMOVAL AND COOLANT CIRCULATION 680h Weest Lawel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . SM S.18 Lou Wets. Lowel . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . SM S-17 SM.10 SPECIAL TEXT EXcRPTIONS 3/4.10.1 MtiMARY CONTAINMENT INTEGRITY . . . . . . . . . .'. . . . . . . . . . . . SM 101 SM.10.2 ROD SEQUUNCE CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . SM 10 2 3/4.10.3 SHUTDOWN MARG:ta DEM0hSTRATIONS . . . . . . . . . . . . . . . . . . 3/410-3 3/4.10.4 RECIRCUMTicN LOOPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . SM 10 4 3/4.10.S CXYGEN CONCENTRA. TION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3M 10 5 3/4.10.8 TRAINING STARTUPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4104 3/4.10.7 4NSORviettSRIt AND HYDROSTATIC TSSTING . . . . . . . . . . . . . . . SM 10-7 l S TS*TE M LEM%6 33,11.ftADl0ACTlW EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 11 1 Tabis 4.11.1 1 Reseactive uguid Weste Sampline end Analysia Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1 1 2 Dese . . . . . . . . . . . '. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1 1.s uedd Redweste Treatment System . . . . . . . . . . . . . . . . . . . . . . . . SM 114 3M i i .,

u.uid N.mu, Tenn. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4.11.2 GASEQUS EFFLUENTS Does Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 114 NINE MILE POINT . UNIT 2 ziv Amendment No. $d [

I lhtDEY _.

miese po,e nuevioman s em_o 4-

- EMil sm.a atFUEUNG OPERATIONS SM.S.1 REACTOR MODE SWITCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 S.1 SM.S.2 INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 9 1 33/4 p.1 4M.S.3 CONTROL ROD POSITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4.9.4 DECAY Tl3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 9-2 3M.S.5 COMMUNICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 S.2 3M.9.6 P.EFUEUNG PLATPORM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 9 2 3/4.9.7 CRANE TRAVEL - SPENT PUEL STORAf5.E POOL . . . . . . . . . . . . . . 35/4 9 2 3/4.9.3 WATER L2 VEL - REACTOR VERSEL AND WATEP LEVEL - SPENT 3/4.9.9 PUEL STORAGE POOL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33M S.2 SM.9.10 COWTROL ROD REMOVAL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 9 2 3/4.9.11 RESIDUAL HEAT RIMOVAL AND COOL ANT CIRCULATION . . . . . 33/4 S 3 3/4.10 SPECl#L TEST EXCEPTIONS .

3/4.10.1 PRIMARY CONTAINnENT INTEGRITY . . . . . . . . . . . . . . . . . . . . . . SS/4101 SM.10.2 ROD SEQUENCE CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . B3/4101 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS . . . . . . . . . . . . . . . . . . . 33/4 10-1 3/4.10.4 RECIRCULATION LOOPP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4101 3/4.10.5 OXYGEN CONCENTRATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4101 3/4.10.6 TRAINING STAR 7UPS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53/4 10 1 3/4.10.7 SiOEFMGMemK AND HYDh0 STATIC TESTING . . . . . . . . . . . . . . . 33/4101 l SyShe LurdGG jji11 RADIOACTIVE EPPLuENTs 3/4.11.1 LIQUID EPPLUENTS Cas M . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33/4 1 1 -1 Does . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B3/4 1 1 1 Liquid Radweste Treatment System . . . . . . . . . . . . . . . . . . . . . . . . 83/4112 Liquid Holdup Tenks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83/4112 NINE MILE POINT- UNIT 2 xx Amendment No. dd [

&\

SURVEILLANCE REQUIREMENTS 4.D.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 i.dividual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified 0/

time interval with a maximum ellowable extension not to exceed 25% of the Y

surveillance interval. /

s

/

4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification '.9.2, shall constitute nonenepliance with the GPERABILITY requirements ior 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 Limitim Condition for Operation have teen performed within the applicable surv ;1ance interval or as othenvise 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 s Code Class 1, 2, and 3 components shall be applicable as follows:

a. g--Toting Inservice :inspection of ASME Code Class 1, 2, and 3 componentsdosemJee-

= ::d: Cla n _1. 2, n 3 5 - - "- " S shall be performiid in'accordi~nce with~$ection XI of the ASME Boiler and Pressura Vessel Code and applicable addenda a.= required by 10 CFR 50.55a(g), except where specif?c written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1). g

[ b. Surveillance intervals specified isSection XI of the ASME Boiler and Pressure Vessel Code and applicabit addenca for tne 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:

amlm~naw

[ 0 .s ~ e.c.

<}d4 dA W y- f95ME G r0- (L / 2, <-. ,/ 3 py w(

MniM D~6~' 4 P- a Q GL o-.~.4

< q u G wcnsasce,(r

' ( y& L Q ' u t t w L h! L k . y A Q) k $ l-~ mnd 2 to cm s2.cca (?)(t)(i).

(. .

NIME MILE POINT - IHTT 2 3/4 0-2 Amendment No. Z8 27,[

s

?

TABLE 4.3.7.5-1 (Centinued)

ACCIDErrr MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

(~

TABLE NOTATIONS

• Excludes sensors; sensor cornparison shall be done in lieu of sensor calibration.

• • Using sample gas containing:

s. One volume percent hydrogen, balance nitrogen,

b. Four volume percent hydrogen, balance nitrogen.

• • • The CHANNEL CAllBRATION shall consist of position indication verification using w9 qiet.uv . Xi "If/ S3094est-entesi) f/c M a s p - y 'l / p J 0 hwn -k t

%C P m .~.. y The CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a o,w point c.alibration check of the detector below 10 R/hr with an installed or portable gamma source.

[

tt Red, Green o* other indication shall be verified as indicating valve position.

NINE MILE POINT - UNIT 2 3/4 3-87 Amendmant I40.

REACTOR COOLANT SYSTEM MOR COOLA?R SYSTEM LEAKAGE

, OPERATIONAL T WAKAGE LIMmNG CONDmONS FOR OPERATION e.

With one or more of the required interlocks shown in Table 3.4.3.2-3 inoperable, restore the Inoperable interlock to OPERABLE status within 7 days or isolate the affected heat exchanger (s) from :Le RCIC steam supply by closing and deenergizing heat exchanger valves 2 RHS*MOV22A and 2RHS*MOV80A or 2RHS*MOV22B and 2RHSWOV80B, as appropriate, f.

With any reactor coolant system leakage greater than the limit in 3.4.3.2.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 HCyr 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 SHOTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

MNCE REOUIREMEN'TS 4.4.3.2.1 The RCS leakage shall be demonstrated to be within each of the above limits by:

a.

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

o.

Monitoring the drywell floor drain tank and equipment drain tank fill rate 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 e .ast once per 24 i hours.

4.4.3.2.2 Each RCS pressure isolation valve specified in Table 3.4.3.2-1 shall be eman*m_Llad OPERABLE by leak testing pursuant to Specifistina 4.0.5g:Ei.d in iD MM % '", i.regr;i"N-2427(bjAnd m

verifying the leakage of each valve to ,

be within the phd limit:

a. At least once per 18 months, and b.

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.

/

Ah W Q i a A G a 92 9 Pp,

i. .

NINE MILE POINT - UNIT 2 3/4 4-14 I AmendmentNo.[

I

_ . . _ . _ __-s

o.

REACTOR COOLANT SYSTEM 3/4.a.6 84ESStlDE/ TEMPERATURE LIMITS

, REACTOR C00LAN) ?YSTEM LfMfTfNC CONDfTIONS FOR ODERATION 675fE^

3.4.6.1 ~he reactor coolant system temperature and pressure shall be limited in a orcance with the limit lines shown on Figure 3.4.6.1 1 for hydrottatic

[I ,

l LMMM Q3.4.6.1-3testing: Figure 3.4.6.1-2 for heatup by non-nuclear means, Figure 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-pcwer PHYSICS TESTS, with;

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

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

c. A maximum tempe 3tu ange of less than or eaual to 20'F in any 1-hour period dJrin C :: y ydrostatic ar.d lee $t testing operations above the l neatuo and coolcown limit curves, and 95 % I EdW6f

d. The reactor vessel flange and head flange temperature greater than or equal to 70*F wnen reactor vessel head bolting studs are under tension. ,

APDLICABILITY: At all times.

ACTION:

l With any of the above limits exceeded, restore the temperature and/or pressure to within the l'..iitts 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 />.

/

SURVEf t_ LANCE REOutREWENTS WSM LMcAGE 4.4.6.1.1 During system heatup, cooldown, andN Lei.;u leaDand l hydrostatic testing operations, the reactor coolant system temperature and pressure snall 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, 3.4.6.1-4, 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,

+

(y

/

A NINE MILE POINT UNIT 2 l NON-CRITICAL HYDROTEST I 1400 -

1200 ,

j i

.P w /  :

.-g 2,* 1000 w

i ,

Eo 800 NON-CRITICAL

{

g}

W- 639 . OPERATION I 600 l' )

t a. -

4 et O2  !

I 4m

$ 400 a3 312 m I .,

as E 200 -

T"'F"T'u"n Pom soLT ',

70 F O 79 39o 0 50 100 150 200 250 300 350 i

\

i DOWNCOMER WATER TEMPERATUREW) i L

f t

l FIGURE 3.4.6.1-1 MINIMUM BELTLINE DOWNCOMER WATER TEMPERATURE FOR PRESSURIZATION DURING T ^ 27IC O HYDROSTATIC TESTING AND 1sas1FG (REACTOR NOT Ch1TICAL) FOR !!P TO 12.8 EFFECTIVE FULL POWER YEARS OF OPERATION '

Svs-tm L'Akn cM NINE MILE POINT - UNIT 2 3/4 4 20 Amerxbnent No.

REACTOR COOLANT SYSTEM 3/a.a.9 RESIOUAL MEAT REMOVAL HOT SHUTDOW LIMIT!NG' CON 0!YTONS FOR OPERATION 3.4.9.1 Two* shutdown cooling mode looos of the residual heat removal (RHR) systes shall be OPERA 4LE ans, unless at least one recirculation pumo is in ooeration, at least one snuteown cooling mose loop snail be in ooeration ",? a -

with each looo consisting of at least:

a. One OPERAELE RHR oumo, and

b. One OPERA 8LE RHR heat exchanger.

l APPLICA8!LITY: OPERATIONAL CON 0! TION 3, with reactor vessel pressues less tnan tne AMR cut-in permissive setpoint.

ACTION:

a. With less than the above reevired RNR shutdown cooling mode loops CPERASLE, immediately initiata corrective action to return the requird loops to 0PERA8LE 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 alternata method caoable of 6ecay heat removal for each inoperable F.HR shutdown cooling moes loop. Se in at least COLD SHUTDOWN within 24 hourf.tt

b. With no RNR shutdown :ooling mode loop in operation, immediately initiate corrective action to return at least one loop to operation as soon as possible. Witnin 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 alter-nata ee' hod and monitor reactor coolant temperature and pressure at least once per hour.

SURVE!LLANCE REQUIREMENTS 4.4.9.1 At least one twtdown cooling mode loop of the residual heat removal system er alternative method shall be detensined to be in operation and circulating reacter 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 AMR 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 optRASLE and in operation.

• • The shutdown cooling pues 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 4-hour period provided the other loop is OPERASLE.

t The RNR shutdown cooling mode loop any be removed from operation during W **'* E**9' -AWO sprem t.cskact it Whenever twa or more RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by thi: ACTION, maintain reactor coolant

, emperature as low as practical by use of alternata heat-remova) methces.

NINE MILE POINT - UNIT 2 3/4 4-35 l amendmentNo.//

EEACTOR COOLANT $YSTEM Rf5!DUALHEATREMOVAL CgLDSHUTOOWN L!fiff!NG CON 0TTIONS FOR OPERATION ,

n. m 3.4.9.2 Two* shuteown cooling mose loops of the residual heat removal (RHR) systas shell be OPERA 8LE and, unless at least one recirculation pumo is in operation, at least one snutsown cooling mooc loop shall be in operation ** ?

with eaco loop consisting of at least:

a. One OPERABLE RHR pump, and

b. One OPERA 4LE RHR heat exchanger.

APPLICA8!LITY: OPERATIONAL CON 0!T'ON 4.

ACTION:

a. 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 hors thereafter, demonstrata the operability of at least one alternate mached capaale cf decay heat removal for each inopersole RNR shutdown cooling mode loop.

b. With no RNR 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 alternata method and monitor reactor coolant temperatur and pressure at least once per hour.

SURVE!LI.ANCE REQUIREMENTS __

4.4.9.2 At least one shutdown cooling mode lesp of the residual heat remeval systes or alternative method shall be determines 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 RNR shutdown cooling mode loop any 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 serve 111ance testing provided the other loop is OPERABLE are in operation.

• • The shutdown cooling pump any 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-hove period provided the other loop is OPERABLE.

t The shutdown cooling mode loop may be removed free operation during testing. l hydrostatic , Q cy sySw LEMM MINE MILE POINT - UNIT 2 3/4 4-36

-t .y

SECTRICAL POWER EYSTEMS

. AC SOURCES .

AC SOUNCES - OPERA 11NG SURVFit i ANCE REOUIREMENTS 4.8.1.1.2.e (Corsonued)

12. Verifyin0 that the automatic load timer relays are OPDMBG with the interval between each load block witnin

• 10% of its design interval for desel generators EDG*1 and EDG*3.

13. Ventyine that the following diesel gerwator lockout foetures prevent diesel generator sterung only when regired:

a) Fc Weions I and H tumng gear engaged smd emergency stop.

b) For Devision Td, en0 i ne in the me' m tenance modo end dssel generator W *_

f. At least once per 18 monthe venfy each desel generator starts and accelerates to at least 6Cs0 RPM within 10 asconds for EDG*1 and EDG*3, and 870 RPM withm 10 seconde for EDG*2. The generstar voltage and frequency for EDG*1 and EDG*3 shes be 4160

• 416 [

volts and to

• 3.0 He withm 10 asconds and 4160

• 416 vt,.a and 60

• 1.2 He within 13 seconds afte* the start menal. The generator vattege and frome,. :y for EDG*2 shal be 4160

• 416 volts and 60 s 1.2 Hr within 15 seconds after the stort signsL This test shof be performed witnm 5 mmutes of stuten0 own d the diesel generator after the diesel generator hm , .

operated for se least 2 hou a at 4400 kW or more for EDG*1 and EDG*3 and 2600 kW or

{

more for EDG*2. For any sta:t of a desel, the diesel must the kaded in accordenas with manufactuier's recommendations. Momentory transients due to changing bus loads shat not Irwelidets this test.

g. At least once per 10 years or after any modfications which could affect desel generator

.nte, ependence by si e ai-- maioneousey, du,ino snu n, and p

venfyh Dthat e5 desel generatorr EDG*1 and EDG*3 accelerate to at least 600 rpm and EDG*2 secolorsers to at least 370 rpm in less then or equel to 10 seconds.

h.

M Atleast once per iu years by: l

1. Dra. rung m:h fuel oil storage tank, removing the accumdated sedmont and cieening tne tank mine a sodium hypochionte solucon, and I

2. Performing a preneurs test of those portons of the diesel fuel oil system deogned to Section III, subesetion ND of the ASME Code in acconhnce with ASME Code Se':tiondi l

Aracio IWD-6000.

)(/

4.8.1.1.3 A8 diesel generator faBues, veEd or norwelid, shou be reponed to the Commiselon pasuant to SpecHication 6.9.2, within 30 days. Reports of diesel generator fabures shot include the information recommended in Position C.3.b of RG 1.108, Revision 1, August 1977. W the number of failures in the lost 100 vand tests on a per nucieer unit basis, is greater then or equel to 7, the report ahat be s@pionnented to include the addlional Informenon recommended in Poeroon C.3.b of RG 1.106, Revision 1, August 1977.

k

/

Specuu. TEST FXCEPDONS S TS* A4WU

/

sma na " -- ... : - -- % ue m 3nomyAyie 71,y,Na l uMmHG CONDrT13NR SOR OPERATION ,

s y s 7s+. 4 s,4 r4 cs j ,

3.1C.7 When cooient temperature h hydrostatic testin0, the average reactor

.. in Table 1.2 for OPERATIONAL CONDmON 4 may be il increased above 200*F, and operation considered not hbe in OPERATIONAL  !

COND8 TION 3. to a3cw parformance C- r2 - 'D hydrostatic test provided the l' l

maximum reacter nochnt tempr sture does 2 - 112*F cnd the foBowine .

OPERATIONAL CONDfT'ON 3 LCO's are met: 4 7 yg., gg i

s. 3J.3, *lselstion Actuation instrumentation *, Functions 1.a.2,1.b, and 3.s and b i of Tanne 3.3.21;

b. 1.E.5.1, ' Secondary Contelament integrity *;

c. 3.4.16.2, ' Secondary Containment Automatic toolatica Dempers"; and  ; {

d. 3.8.5J. "Stenary Gas Trostment System." - l 4 '

APPUCAtitffY: OPERATIONAL CONDmON 4, with everage reactor cooNnt temperature'

> 200 *F. j t

ACIEBk #

}

1 With the requirements of the above specification not satisfied,'.e.a:":r', enter the applicable oorueton of the effected specificanon or immediately suspend actMties that f I could increses the avere0e reactor coolant temperature or presswo and reduce the l overage rm cocient temperature to s200*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. {

ntavmsI w

• RfoutREMENTE 4.10.7 Verify applicable OPERATIONAL CONDITION 3 surveulences (cf specificatione Ested in 3.10.7 are met. i i

l

\

\

\

\

k

,0 r.o T. . 10 7

- No.4

i y

~e

- REACTOR COOLANT SYSTEM

. 8? A cEE 3/a.4.6 'RE550RE/ TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads from temperature and pressure changes in the system.

Tnese cyclic loacs are introduced by normal load transients, reactor trips, and startup ano snutcown operations. The various categories of load cycles used for cesign purposes are provided in Section 3.9 of the FSAR. During startup ano shutcown, the rates of temperature and pressure enanges are limited so that tne maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

The operating limit curves of Figures 3.4.6.1-1 through 3.4.6.1-5 are aerived from the fracture toughness reauirements of 10CFR50, Appendtx G. and ASME Code Section III Appendix G. The curves are based on the RTN0t and stress Fracture intensity f actor information for the react

• vessel components.

toughness limits and the casts 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 RT NOT. The results of these tests are shown in Bases Table 83/4.4.6-1.

Reactor operation and resultan*, ?ast neutron (E greater than i Mev) trradiation will cause an increase in the RTNOT. Thecefore, an adjusted reference temperature, based upon the fluence, copper content, and nickel

' content of the material can be predicted using Bases Figure 83/4.4.6-1 and the recomendations of RG 1.99, Revision 2. " Radiation Embrittlement of Reactor Vessel Materials."

The actual shift in RTNOT of the vessel material will be established periodically during operation by removing and evaluating trradiated spectrr.dns Since installed near the inside wall of the reactor vessel in the core area.

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 Ilmtt 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 octained after removal of the first surveillance capsule will be used to adjust the fluence of Bases Figure 83/4.4.6-1.

The pressure-temperature limit lines shown in Figures 3.4.l.1-1 through 3.4.6.1-5 for $ r :T hydrostatic testing and testing for crtttcal  !

operations have been provided to assure compliance ith the minimum temperature requirements of Appendix G to 10CFR50. g g,, g g

• 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 83/4 4-5 imandmentNo.f

d*

SM.1o SPECIAL TEST EXCEPTIONS m

__A GER

BM.10.1[ PRIMARY CONTAINhlEMLINTEGl4TY The requirement for PRIMARY CONTAINMENT INTEGRTTY is not applicetne during the period when open veessi teste are being performed durin0 the new power PHYSICS TuSTs.

3M 10.2 rop SEQUENCE CONTROL SYSTyp en order to perfonn the tests required in the Technical Specifloodens it le necessary to bypnes the esquence restrainto an contral red movement. The addidonal survelmence requirements ensure that the specifiestions en heet generation rates and shutdown margin requwemerits are not amoeeded during the period when these terts me being performed and that individual red worths de not esoned the venues assumed in the enfaty er'elytis.

sM.1n.s sNuTDOWN MARGIN DEhDNSTRATIONG ,

Performance of shtArwn mergin demonstrctions with the vessel head removed requires addedonal restrictions in enter to ensure that criticedty does not occur. These addinonal restrictions are specified in this Lknitin0 Condition for Operation.

1M.10.4 REclRCULAT10N LOOPS ,

This special test exception permits reactor criticebty under nodlow constions and is required to perform certain startup and PHYSICS TESTS while et low THEfh., POWER levels.

sm.1o.s OXYGEN CONCENTRATION Reuef from the cuygen conoomraden specifications is necessary in order to provide soones to the primary containment durin0 the initial startup and testing pheme of operation. Without this access,the startup and test program could be vostricted and deisyed.

3M.10.8 TR41NING STARTUPR This specisa test exception permits training stortups to be performed with the reactor vessel depreneurtaed at bw THERMAL POWER and temperature while controEing RCS temperature with one RHR subsystem eligned in $e shutdown cooEn0 mode in order to menimise the discharge of co stominated water to the radioactive waste sEapaaal system, s>'s % t 6 AWA GE SM.10 E' ' J_--- ^ DAND WVDiliGimt. Tic TimielG

" .syvem uskKS bl 1

This special test exception atows reactor ._ r_ P -- '-r-mand hytirostatic ter:tirg to }

be performed in OPERATIONAL CONDmON 4 wi hNtmejilisen reactor cocient temperature not exceeding 212'F. The additionally imposed OPERATIONAL CONDm0N 3 requirement for escondary comminmem operabRity provides conservatism in the response of the unit te en operational event. This asews flaxiulity since temperatures approach 190*F during the tegen0 and can drift higher because of decay and mectionical host. Adetionomy, because reactor veneel fluence increases over time, this testme wiu voguire cocient temperatures >200'F. {

l NiNE witt rowT-uNrT 2 S 3/4 m1 Amendmwe No. H, /

TABLE 5.7.1 _1 REACTOR CYCLIC OR TRANSIENT LIMITS AND DESIGN CYCLE OR TRANSIENT CYCLIC OR TRANSIENT LIMIT DESIGN CYCLE OR TRANSIENT 120 heatup and cooldown cycles 70*F to 565'F to 70'F 80 step change cycles Loss of feedwater heaters ,

198 reactor trip cycles 100% to 0% of RATED THERMAL POWER 130 h nd Pressurized to >930 psig and

@ydrostatica tests <1250 psig l

sysn-1 Lenkna l

SG-e ee m

\

NINE NILE POINT - UNIT 2 5-9

____