ML18038A177

Withdraws 860506 Request for Exemption from Section Iiic of 10CFR50,App J for 16 Relief Valves.Justification for Reverse Flow Testing Encl.Fsar Changes Will Be Incorporated in Next Amend.W/Two Oversize Charts

ML18038A177
Person / Time
Site:	Nine Mile Point
Issue date:	07/03/1986
From:	Mangan C NIAGARA MOHAWK POWER CORP.
To:	Adensam E Office of Nuclear Reactor Regulation
References
(NMP2L-0768), (NMP2L-768), NUDOCS 8607080207
Download: ML18038A177 (102)

Text

REQU ORY INFORMATION DISTRIBUT SYSTEN (R IDS)

ACCESSION NBR: 8607080207 DOC. DATE: 86/07/03 NOTARIZED:

NO DOCKET 0 FACIL:50-410 Nine Nile Point Nuclear Stations Unit 2.

Niagara Noha 05000410 AUTH. NANE AUTHOR AFFILIATION NANQAN>C. V.

Niagara Mohawk Power Corp.

REC IP. NANE REC IPIENT AFFILIATION ADENSANiE. Q.

BWR Prospect Directorate 3

SUBJECT:

Withdraws 860506 request for exemption from Section IIIC of 10CFR50i App J for 16 relief valves. Justification For reverse flow testing encl. FSAR changes will be incorporated in next amend. W/two oversize charts.

DISTRIBUTION CODE:

AO17D COPIES RECEIVED: LTR g ENCL g SIZE:

TITLE:

OR Submittal:

Append J Containment Leak Rate Testing NOTES:

RECIPIENT ID CODE/NAl"JE BWR ADTS BWR EICSB BWR PD3 LA HAUQHEYp N BWR RSB INTERNAL: ACRS 07 ELD/HDS3 08 NRR PWR-A ADTS NRR/DSRO/EIB EQ E

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7 HIIASA,IRK U MQHANK NIAGARAMOHAWKPOWER CORPORATION/300 ERIE BOULEVARDWEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474.1511

, JUly 3, 1986 (NMP2L 0768)

Ms. Elinor G. Adensam, Director BNR Project Directorate No.

3 U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue l<ashington, DC 20555

Dear Ms. Adensam:

Re: Nine Mile Point Unit 2 Docket No. 50-410 My letter of May 6, 1986 requested an exemption for 16 relief valves from Section IIIC of Appendix J to 10 CFR 50.

Niagara Mohawk withdraws that request for exemption.

Based upon further review, three of these valves can be reverse flow tested in accordance with Appendix J, and 13 valves discharge under water in the suppression pool.

Justification for reverse flow testing is provided in the attachment.

The discharge lines from the 13 relief valves will be modified prior to fuel load to seal weld closed the vacuum breaker lines.

This will ensure the discharge is below the suppression pool water level, even under expected minimum post-LOCA drawdown water level of the suppression pool.

This is based on a similar situation to other valves, shown in Notes 23 and 24 of Table 6.2-56 of the Final Safety Analysis Report, which are not included in the Type C testing as approved by your staff.

The Final Safety Analysis Report changes are attached which reflect this information.

These changes will be incorporated in the next amendment.

Very truly yours, NLR:ja 1740G Attachment 8607080207 860703 PDR ADOCK 050004l0 PDR C. V. Manga Senior fice President xc:

R. A. Gramm, NRC Resident Inspector Project File (2) ol I I

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II 0

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Niagara Mohawk Power Corporation

)

(Nine Mile Point Unit 2)

)

Docket No. 50-410 AFFIDAVIT C. V.

Man an

, being duly sworn, states that he is Senior Vice President of Niagara Mohawk Power Corporation; that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.

Subscribed and sworn to before me, a Notary Public in and for the State of New York and County of Alendn o, this 8 day of Ci 1986.

PfbtizhJL Q.coi)

Notary Public in and for o.

County, New York My Commi~ss'~]es:

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Nine Mile Point Unit 2 FSAR TABLE 6.2-56 (Cont)

'24'This line consists of inputs from the valves listed below.

The line discharges below the suppression pool water level and therefore is, not exposed to the primary containment atmosphere.

2RHS*SV34B and 2RHS*SV62B steam condensing line safety valves.

2RHS*RV56B -

RHR heat exchanger shell side relief valve.

2RHS*MOV26B and 2RHS~MOV27B -

RHR heat, exchanger vent line isolation valves.

2RHS*MOV26B and 2RHS*MOV27B are open only during steam condensing mode.

Valve position is indicated in the main control room to provide the operator confirmation of valve status.

These valves are included in Type C test in accordance with 20CFR50, Appendix J.

26 2RHS*V117 and 2RHS*V118 vacuum breaker line.

~ 2RHS*RVV35B and 2RHS*RVV36B vacuum breakers.

The above-listed relief, safety, and vacuum breaker valves are included in the Type A containment integrated leak rate test.

They are not included in Type C testing, based on the. design considerations discussed in Note 23.

26

'Deleted.

6'Penetrations Z-99A,B,C,D, and Z-100A,B,C,D contain lines for the hydraulic control of the reactor recirculation flow control valve.

These lines contain hydraulic fluid used to position the reactor recirculation flow control

valve, and are protected against the effects of pipe whip and jet impingement.

26 Amendment 26 23 of 24 May 1986

'Q-i )c,>x

Insert A This line consists of inputs from the applicable valves listed below.

The line discharges at elevation 195'-6", which is 2'-2" below the minimum water level in the suppression pool and, therefore, is not exposed to the primary containment atmosphere.

All of the valves are relief valves which provide relief for high/low pressure interface leakage, except 2RHS*RV108, which provides relief for upstream level control failure.

For penetration Z73 2RHS*RV108 2RHS*RV20C For penetration Z98A 2RHS*RV61A 2RHS*RY20A 2RHS*RV110 2CSL*RV123 2CSL*RV105 2RHS*RV139 For penetration Z98B 2RHS*RV61C 2RHS*RV61B 2CSH*RV114 2CSH*RV113 2RHS*RV20B The above-listed relief valves are included in the Type A containment integrated leak rate test for external leakage.

They are not included in Type C testing, based on the design considerations discussed in Note 23.

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Nine Mile Point Unit 2 FSAR QUESTION F480.37 (6.2.6)

Appendix J,

Section III.C.1 prescribes methods for conducting the containment isolation valve leak rate tests.

These requirements state that containment isolation valves should normally be leak tested with the test pressure appli'ed in the same direction the valve must function to preclude leakage in an accident condition.

Reverse direction testing is permitted only if it can be demonstrated that such testing yields results which are equivalent or more conservative than those obtained using same direction as post accident flow testing.

Iist the containment isolation valves for which Type C leak testing with reverse flow is used.

For each justify by means of test data or valve design arguments that this testing is equivalent or more conservative than "same direction as post accident flow" testing.

RESPONSE

Table 480. 37-1 includes the following I I RT test penetrations:

Containment penetration number.

2.

System in which the reverse tested valve is installed.

3.

Valve identification number.

Type of valve construction and the justification necessary to ensure the reverse test is as conservative as testing the valve in the forward direction.

Amendment 17 QErR F480.37-1 January 1985

Nine Mile Point Unit 2 FSAR TABLE 480.37-1 REVERSE TESTED CONTAINMENT ISOLATION VALVES Pene-tration No.

Z8A Z8B Z12 Z18 Z17 Z19 Z21A Z48 Z51 Z50 Z49 Z55A Z55B Z56A Z57A Z56B

, Z57B Z58 Z59 Z60A Z60C Z60D Z61C Z60E Z60G Z60H Z61F Z01A Z01B Zojc Z01D nba.~C 9 RHR RHR CHS ICS ICS ICS ICS CPS CPS CPS CPS HCS HCS HCS HCS HCS HCS CPS CPS CMS CMS CMS CMS CMS CMS CMS CMS MSS MSS MSS MSS MOV25A MOV25B MOV118 MOV143 MOV136 MOV122 MOV128 AOV108 AOV109 AOV107 AOV106 MOV4A MOV4B MOV6A MOVSA MOV6B MOV5B SOV122 SOV121 SOV61A SOV63A SOV33A SOV34A SOV61B SOV63B SOV33B SOV34B SOV97A SOV97B SOV97C SOV97D Split Disc Split Disc Split Disc Globe Split Disc Split Disc Split Disc Butterfly Butterfly Butterfly Butterfly Globe Globe Globe Globe Globe Globe Globe Globe Plug Plug Plug Plug Plug Plug Plug Plug Globe Globe Globe Globe Gate Gate Gate Gate Gate Gate Justi-fication Valve 26 Justification Notes:

1. Split disc gate valves may be tested using a test connection (TC) between the discs.

This is a

conservative test since both LOCA and non-LOCA seat leakage is measured.

Amendment 26 1 of 2 May 1986

lJ t Jg"

Insert B

Penetration Zll Z33B Z348

~Ss tern RHS CCP CCP Valve I.D.

RV152 RV170 RV171

~Te Relief Relief Relief Justification 1740G

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0 Nine Mile Point Unit 2 FSAR TABLE 480.37-1 (Cont)

2. Globe valves are oriented to ensure LLRT test pressure tends to unseat the valve, whereas LOCA pressure will tend to seat the valve.

This is conservative for testing.

3. Butterfly valves are reverse tested which will provide equivalent results since the seating area(s) and test pressure force(s) will be equal in either direction.

4. Plug valves are of the pressure-balanced bellows type.

By design, neither upstream nor downstream pressure can exert a force on the disc, and the spring force of the bellows is the only force tending to seat the valve disc.

Reverse flow testing is therefore equivalent to testing in the same direction as post-accident flow.

ZHstr4 Amendment 26 2 of 2 May 1986

~

Insert C

5.

Relief valves are nozzle type spring actuated relief valves.

The valves are orientated to ensure test pressure tends to unseat the valve, whereas LOCA pressure will tend to seat the valve.

This is conservative for testing.

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Div I 26 Div II Aaendaeat 26 15 of 24 haF 1986

Changes to Technical Specifications in Area of Rod Worth Minimizer

0

Subject:

Justification for changes to Technical Specifications in area of rod worth minimizer The requested changes to Technical Specifications are enclosed.

The current Technical Specification 3/4 10.2 allows suspension of constraints imposed by the rod sequence control system (RSCS) provided that the rod worth minimizer (RWM) is operable.

This suspension is allowed for those tests identified in the Technical Specification 3/4 10.2.

To perform those tests, however, it is also necessary to suspend the constraints imposed by RWM to allow for control rod movement.

Therefore, Niagara Mohawk requests changes to the Technical Specifications to allow bypassing the RWM, in conjunction with bypassing the RSCS, for those tests to be performed.

The enclosed changes are consistent with Hope Creek and Limerick Unit 1

Technical Specifications 3/4 10.2.

CHANGE RE(}VESTED FOR CERTIFICATION

~

~

REACTIVITY CONTROL SYSTEMS 3/4. 1.4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIMIZER LIMITING CONDITIONS FOR OPERATION 3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2" ~when THERMAL POWER is less than or equal to 20K of RATED THERMAL POWER, the minimum allowable low-power setpoint.

ACTION:

a 0 With the RWM inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console.

Otherwise, control rod movement is permitted only by actuating the manual scram or by placing the reactor mode switch in the Shutdown position.

b.

The provisions of Specification, 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.1.4.1 The RWM shall be demonstrated OPERABLE:

a o In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />'efore RWM automatic initiation when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.

b.

In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods for the purpose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.

c.

In OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RWM automatic initiation when reducing THERMAL POWER, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.

d.

By demonstrating that the control rod patterns and sequence input to the.

RWM computer are correctly loaded following any loading of the program into the computer.

3/4 1"16

<* Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM before withdrawal of control rods for the purpose of bringing the reactor to criticality.

NINE MILE POINT - UNIT 2

REACTIVITY CONTROL SYSTEMS.

ROD SE UENCE CONTROL SYSTEM LIMITING CONDITIONS FOR OPERATION

3. 1.4.2 The rod sequence control system (RSCS) shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2" "*, when THERMAL POWER is less than or equal to 20K RATED THERMAL POWER, the minimum allowable low-power set-point.

ACTION:

a ~

b.

With the RSCS inoperable, control rod movement shall not be permitted, except by a scram.

With inoperable control rod(s),

OPERABLE control rod movement may continue by bypassing the inoperable control rod(s) in the RSCS provided that:

1.

The position and bypassing of inoperable control rod(s) are verified by a second licensed operator or other technically qualified member of the unit technical staff, and 2.

There are not more than 3 inoperable control rods in any RSCS group.

SURVEILLANCE RE UIREMENTS 4.1.4.2 The RSCS shall be demonstrated OPERABLE by:

Performance of*a self-test:

1.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to each reactor startup, and 2.

Prior to movement of a control rod after rod inhibit mode automatic initiation when reducing THERMAL POWER.

b.

Attempting to select and move an inhibited control rod:

1.

After withdrawal of the first in-sequence control rod for each reactor startup, and 2.

Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after rod inhibit mode automatic initiation when reducing THERMAL POWER.

" See Special Test Exception 3.10.2.

• • Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RSCS prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.

NINE MILE POINT " UNIT 2 3/4 1-17

SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD SE UENCE CONTROL SYSTEM LIMITING CONDITIONS FOR OPERATION red garth r6niviPt'<+) W"

~;~.II S.t,+. ( andky

3. 10.2 The sequence constraints imposed on control rod groups by the rod sequence control system (RSCS) per Specification 3.1.4.2 may be suspended by

~

fbyp 1

f h f11 ig p

i d

a.

Shutdown margin demonstrations, Specification 4.1.1.

b.

Control rod scram, Specification 4.1.3.2.

c.

Control rod friction measurements.

d.

, Startup Test Program with the THERMAL POWER less than 20K of RATED THERMAL POWER.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the requirements of the above specificatson not satisfied, verify that th@RSCS is OPERABLE per Specification ~4-.8-. 3.I.+. I ~d S./0;2,~ r~p~c~ivdy C R,Whl aQ ar Qe, Rsc5 SURVEILLANCE RE UIREMENTS 4.10.2 When the sequence constraints imposed on control rod groups by the RSCS are bypassed, verify:

~/or R M

Q,~

That movement of control rods from 75K ROD DENSITY to power setpoint is limited to the approved control rod sequence during scram and friction tests.

~ c, Conformance with this specification and test procedures by licensed operator or other technically qualified member of

'ical staff.

the RSCS 'low-withdrawal a second the unit tech-soohol roa'roouxoient proson bod pr +is eh'sAIII /5 rsnfwd 8y rosoooua',iueroior or ~~ Mnicn/y/ g~y/>pi'ed'bar og

>Vie unif AcA<<'~i ~~ 8 p~"

+ Mhc'e@.c+or

~ygo[~.'hol rooveueoT af cooers I rod dung shuI'douo iosgiri dseoss)notions is i'~ted

&sees Irsssnbod sstrueuo pg <oudp~on s rs,g, NINE MILE POINT - UNIT 2 3/4 10-2

SPECIAL TEST EXCEPTION 3/4. 10. 3 SHUTOOWN MARGIN OEHONSTRATIONS

~

~

LIMITING CONOITIONS FOR OPERATION 3.10.3 The provisions of Specification 3.9.1, Specification 3.9.3, and

~ Table 1.2 may be suspended to permit the reactor mode switch to be in the Startup position and to allow more than one control rod to be withdrawn for shutdown margin demonstration, provided that at least the following require-ments are satisfied.

a.

The source range monitors are OPERABLE with the RPS circuitry "shorting links" removed per Specification 3.9.2.

b.

The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and is pro-grammed for the shutdown margin demonstration, or conformance with the shutdown margin demonstration procedure is verified by a second licensed operator or other technically qualified member of the unit technical staff.

c.

The continuous rod withdrawal control shall not be used during out-of-sequence movement of the control rods.

d.

No other CORE ALTERATIONS are in progress.

APPLICABILITY:

OPERATIONAL CONDITION 5, during shutdown mar gin demonstrations.

ACTION:

With the requirements of the above specification not satisfied, immediately place the reactor mode swi tch in the Shutdown or Refuel pos ition.

SURVEILLANCE RE UIREMENTS 4.10.3 Within 30 minutes prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of a shutdown margin demonstration, verify that; The source range monitors are OPERABLE per Specification 3.9.2, b.

c The rod worth minimizer is OPERABLE with the required program per Speci-fication 3.1.4. 1 or a second licensed operator or other technically quali-fied member of the unit technical staff is present and verifies compliance with the shutdown demonstration procedures, and No other CORE ALTERATIONS are in progress.

NINE MILE POINT " UNIT 2 3/4 10-3

Changes to Techni cal Speci ficati ons on Other Items

Subject:

Changes to Technical Specifications for items required for certification The requested changes to Technical Specifications are enclosed.

These changes are requested for certification and reflect the Nine Mile Point Unit 2 design.

I 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 1 SAFETY LIMITS THERMAL POWER Low Pressure or Low Flow

2. 1. 1 THERMAL POWER shall not exceed 25K of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10K of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25K of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than lOX of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the require-ments of Specification 6.Q THERMAL POWER Hi h Pressure and Hi h Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with two recirculation loop operation and shall not be less than 1.07 with single recirculation loop operation with the reactor vessel steam dome pressur e gr eater than 785 psig and core flow greater than lOX of rated flow.

APPLICABILITY:

OPERATIONAL CONDITIONS '1 and 2.

ACTION:

With MCPR less than 1.06, with two recirculation loop operation or less than 1.07 with single loop operation, the reactor vessel steam dome pressure greater than 785 psig, and core flow greater than 1(C of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.~

REACTOR COOLANT SYSTEM PRESSURE

2. 1.3 The reactor coolant system pressure, as measured in the reacto~ vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and 4, ACTION:

With the reactor coolant system pressure as measured in the reactor vessel steam dome above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification

6. Q REACTOR VESSEL WATER LEVEL 2.-1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

NINE MILE POINT - UNIT 2 2-1

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B

LIMITING CONDITIONS FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figure 3.2.3-1 times the Kf shown in Figure 3.2.3-2 and adjusted as required for reduced feedwater temperature with:"

< =

('ave

'B A

B wher e:

xA = 0. 86 seconds, control rod,average scram insertion time limit to notch 39 per Specification

3. 1. 3.3, Ng tB = 0.688

+ 1.65

[0.052],

n E

= i=1 ave N.x.

1 1

N.

n

= number of surveillance tests performed to date in cycle N. = number of active control rods measured in the i-th surveillance test, ti = average scram time to notch 39 of all rods measured in the i surveillance test

.th N

= total number of active rods measured in Specification 1

4. l. 3. 2. a.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

"Add 0.03 to the operating MCPR when feedwater temperature is <4004F and

>320 F or add 0.06 to operating MCPR when feedwater temperature

<320 F and

>250 F.

These delta HCPR adjustments are only required for steady state operation when feedwater temperature is reduced.

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

0

fR~Pl. 9P TABLE 4. 3. 6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE RE UIREMENTS TABLE NOTATIONS With THERMAL POWER greater than or equal to 30K or more f RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(a)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b)

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before star tup, if not performed within the previous 7 days.

(c)

Includes reactor manual control multiplexing system input.

NINE MILE POINT - UNIT 2 3/4 3-66

TABLE 3.3.7.5-1 ACCIDENT HONITORING INSTRUHENTATION INSTRUHENT 1.

Reactor Vessel Pressure 2.

Reactor Vessel Water Level REQUIRED NUHBER OF CHANNELS 2'INIHUH CHANNELS OPERABLE APPLICABLE OPERATIONAL CONDITIONS ACTION 1,

2 80 a.

Fuel Zone b.

Wide Range 3.

Suppression Pool Water Level a.

Narrow Range b.

Wide Range 4.

Suppression Pool Water Temperature 5.

Suppression Chamber Pressure

~ ~ Drywell Pressure M '3 Orwell Air Temperature AK 't Drywe11 Oxygen Concentration W lO Orywell Hydrogen Concentration Analyzer and Honitor

~. I>Safety/Relief Valve Position Indicators*

8, 2/quadrant 2/Valve 4, 1/quadrant 1

1/Valve 1,

2 1,

2 1, 2, 3

1, 2, 3

1, 2

1, 2

1, 2

1I 2

1, 2

1, 2

1, 2

80 80 83 83 80 80 80 80 80 80 80 l)2

TABLE 3. 3. 7. 5-1 (Continued)

ACCIDENT HONITORING INSTRUHENTATION INSTRUHENT

~.l2Drywe11 High Range Radiation Honitors

~.l3RHR Heat Exchanger Service Water Radiation Honitor

~.i%Refuel Platform Area Radiation Honitor REQUIRED NUHBER Of CHANNELS 2

HINIHUH CHANNELS OPERABLE

'APPLICABLE OPERATIONAL CONDITIONS ACTION 1, 2, 3

81 82 1/Heat Exchanger 1/Heat Exchanger 1, 2, 3

81 M.ls'eutron F1uxt APRH IRH SRH

~lb Primary Containment Isolation

- Valve Position Indication 2

2 2

1, 2

1, 2 1

1, 2

80 80 80

• Acoustic monitoring and tail pipe temperature

""When handling fuel, or components in the fuel pool or reactor cavity.

tNeutron flux indication is sufficient to meet the OPERABILITY requirement of this specification.

m M

m INSTRUMENT I

CHANNEL CHECK CHANNEL CALIBRATION 1.

Reactor Vessel Pressure 2.

Reactor Vessel Water Level a.

Fuel Zone b.

Wide Range 3.

Suppression Pool Mater Level a.

Narro~ Range b.

Mide Range 4.

Suppression Pool Water Temperature 5.

Suppression Chamber Pressure A. 7 Drywell Pressure W S Orywell Ait Temperature M 't Orywell Oxygen Concentration W i> Orywell Hydrogen Concentration Analyzer and Honitor M.ll Safety/Relief Valve Position Indicators

~.12 Drywell High Range Radiation Honitors 3K lsRHR Keat Exchanger Service 'Water Radiation Honitor

~.l'f Refuel Platfora Area Radiation Ihonitor

~ l5 Neutron, Flux a.

APRH b.

IRH c.

SRH

~ lt Primary Containment Isolation Valve Position Indication R

R R

R R*

R R

R*

R(*A R

Rt R

R R

R R

TABLE 4. 3. 7. 5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS APPLICABLE OPERATIONAL CONDITIONS 1,

2 1,

2 1,

2 1, 2, 3

1, 2, 3

1, 2

1, 2

1, 2

1,- 2 1,

2 1,

2 1,

2 1, 2, 3

1, 2, 3

tt 1, 2 1,

2 1

1, 2

5UpPREus'<o~ Qpg~~~

p < ~<

~Q~Pp~~~~

TABLE 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES -

LEAKAGE PRESSURE MONITORS INSTRUMENT NUMBER P'sX. 1~+

2RHS PPRM

'PsX 1B 2RHS"&Pic~

psxvc<

2RHS RS36c 2CSL"PS108 2RHS"PISlll VALVE NUMBER 2RHS "MOV24A

'RHS*MOV40A 2RHS"MOV22A 2RHS"MOV23A 2RHS"MOV80A 2RHS MOV24B 2RHS*MOV40B 2RHS"MOV22B 2RHS"MOV23B 2RHS*MOV80B 2RHS"MOV104 2RHS*MOV24C 2CSL*MOV104 2RHS"MOV112 2RHS"MOV113 SETPOINT PSIG 47s ~e 475 M 475 Re 475 ae 475 ke 475 &

47S ~e 475 ~e 47S ~6 475 Ie 475 ae 475 K 525 xe 171 te 171 &

TABLE 3.4.3.2-3 HIGH/LOW-PRESSURE INTERFACE INTERLOCKS INSTRUMENT NUMBER 2RHS PPiV5A/76A 2RHS%+75B/76B VALVE NUMBER 2RHSMOV23A 2RHS"MOV23B SETPOINT PSIG aes *12 465 ~

NINE MILE POINT - UNIT 2 3/4 4-16

EMERGENCY CORE COOLING SYSTEMS ECCS - OPERATING SURVEILLANCE RE UIREMENTS 4.5. 1 (Continued) e.

For the ADS by:

1.

At least once per 31 days, performing a CHANNEL FUNCTIONAL TEST of the accumulator backup compressed gas system, low-pressure alarm system.

2.

At least once per 18 months:

a)

Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating

sequence, excluding actual valve actuation.

b)

Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig" and observing that either:

1)

The SRV discharge acoustic monitoring system responds accordingly, or 2)

The control valve or bypass valve'esponds accordingly, or 3)

There is a corresponding change in the measured steam flow, or 4)

The SRV discharge line temperature monitoring system responds accordingly.

c)

Performing a

CHANNEL CALIBRATION of the accumulator backup compressed gas

system, low-pressure'larm

system, and ifyi g 1

p i f1635~p decreasing pressure.

I.0) -l. 0 d)

Performing a'leak rate test for ADS SRV pneumatic operators by pressurizing each ADS accumulator at 178 psig (supply header high pressure alarm) up to its supply header isolation check valve with the SRV in the open position.

Total leakage rate for each SRV shall not exceed 0.5 SCFH for the SRV actuated by either of the ADS solenoids.

" The provisions of Specification 4:0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

\\

NINE'MILE POINT - UNIT 2 3/4 5"5

foal TABLE 3.6. 3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES ISOLATION VALVE NO.

VALVE m

VALVE FUNCTION GROUP ISOLATION SIGNAL(a)

MAXIMUM CLOSING TIHE (SECONDS) 2IAS"SOV164 2IASASOV165 2IAS"SOV166 2IASASOV184 2IASASOV168 2IAS"SOV180 2IASASOV167, 2IASASOV185 2HCSAMOV1 A,B 2HCS"HOV2 A,B 2HCS*MOV3 A,B

. 2HCS*HOV4 A,B(n) 2HCS"MOV5 A,B(n) 2HCS"MOV6 A,B(n) 2CPS"SOV119 2CP SASOV120 2CPS" SOV121(n) 2CPS"SOV122(n) 2CHS*SOV24 A,B,C,D 2CHS*SOV26 A,B,C,D 2CHS"SOV32 A,B 2CHS"SOV33 A,B(n) 2CHS"SOV34 A,B(n) 2CHS"SOV35*A,B 2CHS"SOV60 A,B 2CHS"SOV61 A,B(n) 2CHS"SOV62 A,B 2CHS"SOV63 A,B(n)

ADS Hdr A Nz supply Outside IV ADS Hdr B N~ supply Outside IV IAS Drywell Relief Valve Outside IV IAS Drywell Relief Valve Inside IV Inst. Air to Testable Check Outside IV Inst. Air to Testable Check Inside IV IAS to Test Ck.

8 Vac. Bkrs. Outside IV IAS to Test Ck.

8 Vac. Bkrs. Inside IV Hq Recombiners Sply to Supp.

Chamber Outside IV's Hq Recomb.

Ret.

from Supp.

Chamber Outside IV's Hq Recomb.

Return.from Drywell Outside IV's 8 Hq Recomb.

Suply. to Supp.

Chamber Inside IV's Hq Recomb.

Ret.

from Supp.

Chamber Inside IV's Hq Recomb.

Ret.

from Drywell Inside IV's Containment Purge to Supp.

Chamber Outside IV Containment Purge to Orywell Outside IV Containment Purge to Supp.

Chamber Inside IV Containment Purge to Drywell Inside IV CMS from Drywell Inside 8 Outside IV's CHS from SP Inside 8 Outside IV's CHS to Drywell Outside IV's CHS to Drywell Inside IV's CHS to SP Inside IV's CMS to SP Outside IV's CHS to Drywell Outside IV's CHS to Drywell Inside IV's CMS to Drywell Outside IV's CHS to Drywell Inside IV's B,F;Z,RM B,F,Z,RM B,F,Z,RH B,F,Z,RH B,F,Z,RH B,F,Z,RH B,F,Z,RH B,F,Z,RH B,F,Z,RH B,F,Z,RM B,F,Z,RH B,F,Z,RM B,F,Z,RH B,F,Z,RH B;F,Y,Z,RH B,F,Y,Z,RH B,F,Y,Z,RH B,F,Y,Z,RM B,F,Z,RH B,F,Z,RM B,F,Z,RM B,F,Z,RH B,F,Z,RH B,F,Z,RH B,F,Z,RM B,F,Z,RH B,F,Z,RH B,F,Z,RH 30 30 30 30 30 30 5

5 5

5 5

5 5

5 5

5 5

5

lb pi Q'tV

~

t V

PLANT SYSTEMS PLANT SERVICE WATER SYSTEM PLANT SERVICE WATER SYSTEM - OPERATING SURVEILLANCE RE UIREMENTS 4.7.1.1.1. d (Continued) 3, Each pump runs and maintains service water pump discharge pressure equaT to or greater than 80 psig with a pump flow equal to or greater than 6500 gpm.

4.

The resistance to round is > 28 ohms for each feeder cable that powers the intake escsng heater systems.

4.7.1.1.2 The Intake Deicing Heater System shall be demonstrated OPERABLE:

a.

. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the intake tunnel water temperature is greater than or equal to 39'F, or b.

At least'nce per 7 days by verifying that the current of the heater feeder cables at the motor control centers is 10 amps~

or more (total for three phases) at > 518 volts per divisional heater in each intake struc-

.ture.

I

" For 7 heater elements in operation.

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

PLANT SYSTEMS PLANT SERVICE MATER SYSTEM PLANT SERVICE MATER SYSTEM -

SHUTDOWN SURVEILLANCE RE UIREMENTS 4.7.1.2.1.d (Continued)'

3.

Each pump runs and maintains service water pump discharge pressure equal -to or greater than 80 psig with wach pump flow equal to or greater than 6500 gpm.

4.

The resistanc is 28 ohms or more for each feeder cable that powers the intake deicing heater systems.

I 4.7.1.2.2 The Intake Deicing Heater System shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the intake tunnel water temperature is greater than or equal to 39'F,,or b.

At least once per 7 days by verifying that the current of the heater feeder cables at the motor control centers is 10 amps* or more (total for 3 phases) at > 518 volts 'per divisional heater in each intake Structure.

" For 7 heater elements in operation.

NINE MILE POINT - tjNIT 2 3/4 7-6

SPECIAL TEST EXCEPTIONS 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING LIMITING CONDITIONS FOR OPERATION 3.10.7 During initial core loading within the Startup Test Program the pro-visions of Specification 3/4.9.2 may be suspended provided that at least two source range monitor (SRM) channels with detectors inserted to the normal operating level are OPERABLE with:

a.

One of the required SRM channels continuously indicating" in the control

room, b.

One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant,""

c.

The RPS "shorting links" shall be removed prior to and during fuel

loading, d.

The reactor mode switch is OPERABLE and locked in the Refuel position.

APPLICABILITY:

OPERATIONAL CONDITION 5 ACTION:

With the requirements of the above specification not satisfied, immediately suspended all operations involving initial core loading.

SURVEILLANCE RE UIREMENTS 4.10.7.1 Within one hour prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the initial core loading verify that:

a.

The above required SRM channels are OPERABLE by:

1.

Performance. of a CHANNEL CHECK"""

2.

Confirming that the above required SRM detectors are at the normal operating level and located in the quadrants required by Specification 3. 10.7.

"Up to 16 fuel bundles may,be loaded without a visual indication of count rate.

""The use of spe'cial movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detec-tors are connected to the normal SRM circuits.

• ""May be performed by use of movable neutron source.

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

SPECIAL TEST EXCEPTIONS SPECIAL INSTRUMENTATEON - INITIAL CORE LOADING SURVEILLANCE RE UIREMENTS Continued

4. 10.7. 1 (Continued) b.

The RPS "shorting links" are removed.

c.

The reactor mode switch is locked in the REFUEL position.

4.10.7.2 Perform a CHANNEL FUNCTIONAL TEST for the above required SRM channels within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start and at least once per 7 days 'during initial core loading.

4.10.7.3 For at =least one SRM channel, verify that the count rate is at least 0.7 cps".

a.

Immediately following the loading of the first 16 fuel bundles.

b.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter during initial core loading.

~g 1-1 1

2.

113 1,3 p.

NINE MILE POINT - UNIT 2 3/4 10-8

Changes to Technical Specifications on Other Items

Subject:

Editoral changes to Technical Specifications The requested changes to Technical Specifications are enclosed.

These items are editorial changes and are self explanatory.

TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT APPLICABLE

'INIMUM OPERATIONAL OPERABLE CHANNELS.

CONDITIONS PER TRIP SYSTEM (a)

ACTION 5.

Main Steam Line Isolation Valve-Closure 6.

Main Steam Line Radiation-High 7.

Drywell Pressure

- High 8.

Scram Discharge Volume Water Level - High a.

Transmitter Trip Units b.

Float Switches I'.

Turbine Stop Valve - Closure 10.

Turbine Control Valve Fast

Closure, Valve Trip System Oil Pressure

- Low ll.

Reactor Mode Switch Shutdown Position 12.

Manual Scram 1(e) 1, 2

5(h) 1, 2

5(h) 1, 2

3, 4

5 1,

2 3,

4 5

l4 2@8 2(g) 4(i) 2(J)

~

~

INSTRUMENTATION 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION

'TWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION 3.3.4. 1 The anticipated transient without scram recirculation pump Trip (ATWS-RPT) System instrumentation channels shown in Table 3.3.4. 1-1 shall be OPERABLE with their Trip Setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONDITION 1.

ACTION:

a.

b.

C.

d.

e.

With an ATWS-RPT system instrumentation channel Trip Setpoint less conser-vative the value shown in the Allowable Value column of Tab e 3.3.4.1-2, declare the channel inoperable until the channel is re-stored to OPERABLE status with the channel Trip Setpoint adjusted consis-tent with the Trip Setpoint value.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both Trip Systems, place the inoperable channel(s) in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

With the number of OPERABLE channels two or more less than required by the Minimum Operable Channels per Trip System requirement for one Trip System and:

l.

If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure

channel, place both inoper-able channels in the tripped condition" within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.

If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure

channels, declare the Trip System inoperable.

With one Trip System inoperable, restore the inoperable Trip System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With both Trip Systems inoperable, restore at least one Trip System to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

" The inoperable channels need not be placed in the tripped condition if this would cause the Trip Function to occur.

In this case, the inoperable chan-nels shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or the Trip System shall be declared inoperable.

NINE MILE POINT - UNIT 2 3/4 3-45

SYSTEM/SUBSYSTEM" 1.

RCIC System 2.

RHR System TABLE 3.3.7.4-2 REMOTE SHUTDOWN SYSTEM CONTROLS MINIMUM OPERABLE SYSTEMS/

SYSTEMS/SUBSYSTEMS SUBSYSTEMS 5.'uclear Steam Supply Shutoff System (Isolation Groups 4

8 5 Reset)

Nitrogen Supply to ADS Accumulator Tanks A.

Shutdown Cooling Mode B.

Suppression Pool Cooling Mode 3.

Service Water System A.

Pumps B.

Supply Valves to Division I Division II Diesels 4.

ADS System (Pressure Relief) 6 1/Division 4 Valves/Division 1/Division 1/Division 1/Division 1/Division 2/Division 1/Division 4 Valves/Divisio 1/Division 1/Division

• Includes applicable transfer switches NINE MILE POINT - UNIT 2 3/4 3-79

TABLE 3.6. 3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES

. ISOLATION VALVE NO.

D.

Other

~Sf II li f 2RHS"RV20 A,B,C(o) 2RHS"RV61 A,B,C(o) 2RHS"RV108(o) 2RHS*RV110(o) 2RHS"RV139(o) 2RHS"RV152(o) 2RHS"RV56 A,B(d).

2RHS"SV34 A,B(d) 2RHS*SV62 A,B(d) 2RHS"RVV35 A,B(d).

2CS1" RV105(o) 2CSL"RV123(o) 2RHS"RVV36 A,B(d) 2CCP*RV170(o) 2CCP*RV171(o) 2CSH*RV113(o) 2CSH"RV114(o)

VALVE FUNCTION RHS Rv disch. to SP Outside IVs RHS Rv disch. to SP Outside IVs RHS R

disch. to SP Outside IVs SDC to RHR Pump suction Rv RHR Hdr. Flush to Radwaste RV SDC Supply from RCS RV Inside IV RHS HX shell side RVs RHS HX steam supply Safety valves RHS HX steam supply Safety valves RHS Vacuum Breakers CSL RV Disch. to SP Outside IV CSL RV Oisch. to SP Outside IV RHS Vacuum Breakers CCP RV Discharge Inside IV CCP RV Discharge Inside IV CSH RV Disch. to SP Outside IV CSH RV Disch. to SP Outside IV VALVE ISOLATION GROUP SIGNAL(a)

HAXIHUH CLOSING TIHE (SECONDS)

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT STANDBY GAS TREATMENT SYSTEM LIMITING CONDITIONS FOR OPERATION fiiViLUf+ff 3.6.5.3 Two independent standby gas treatment (SGTS) subsystems shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3, and *.

ACTION:

a.

With one standby gas treatment subsystem inoperable:

l.

In OPERATIONAL CONDITION 1, 2 or 3, suspend all VENTING or PURGING of the drywell and/or suppression chamber"" within 30 minutes, and restore the inoperable subsystem to OPERABLE status within 7 days, 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 />.

In OPERATIONAL CONDITION ", restore the inoperable OPERABLE status within 7 days, or suspend handling fuel in the reactor buildin~>

ORE ALTERATIONS, and with a potential for draini~ the reactor vessel.

~

~

Specsfscatlon 3.0.3 are not applicable.

b.

With both standby gas treatment subsystems inoperable:

subsystem to of irr8diated operations The provisions of 2.

In OPERATIONAL CONDITION 1, 2, or 3, suspend all operations involving VENTING, PURGING, or pressure control of the drywell or suppression chamber and initiate action within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 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 COLO SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

In OPERATIONAL CONDITION *, suspend handling of irradiated fuel in the reactor building, CORE ALTERATIONS or operations with a potential for draining the reactor vessel.

The provisions of Specification 3.0.3.

are not applicable.

" When irradiated fuel is being handled in the reactor building and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

"" The requirement to suspend VENTING or PURGING with one inoperable SGTS sub-system shall not apply to the use of valves 2CPS"AOV108 (14-inch) and 2CPS"AOV110 (14-inch), or 2CPS"AOV109 (12"inch) and 2CPS"AOVlll (12-inch), for primary containment pressure control, provided 2GTS~AOV101 is closed, and its 2"inch bypass line is the'only flow path to the standby gas treatment system.

NINE MILE POINT - UNIT 2 3/4 6-42 AN 35 eS

PLANT SYSTEMS FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM SURVEILLANCE RE UIREMENTS 4.7.7.1.3 The diesel-driven fire pump starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that:

1.

The electrolyte level of each cell is above the plates, 2.

The pilot cell specific gravity, corrected to 77OF and full electrolyte level, is 1.235 or more, 3.

The overall battery voltage is 25.5 volts" or more with the battery on float charge.

b.

At least once per 92 days by verifying that all cell parameters for all battery cells are demonstrated OPERABLE per Specifica-tion 4.7.7.1.3.a and the difference between the pilot cell with the highest specific gravity when compared to the pilot cell with the lowest specific gravity is 0.015 or less.

At least once per 18 months by verifying that:

1.

The batteries, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, and 2.

Battery and terminal connections are clean, tight, and free of corrosion.

" An overall battery voltage of 25.5 volts or more represents 12 pilot cells each earring at least a 2. 13-volt charge.

NINE MILE POINT - UNIT 2 3/4 7-26 AN 8

ELECTRICAL POWERSYSTEMS'f'l E i AC SOURCES AC SOURCES - OPERATING LIMITING CONDITIONS FOR OPERATION 3.8. 1. 1 (Continued)

ACTION:

b.

(Continued) separately for each diesel generator within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."

Restore the in-operable diesel generator to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.

C.

d.

With one offsite circuit of the above required AC sources and diesel gen-erato EDG"1 or EDG"3 of the above required AC electrical power sources inoperable, demonstrate the OPERABILITY of the remaining AC sources by performing Surveillance Requirement 4.8.1.1.1 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 every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.',

If a diesel generator became inoperable from any cause other than preplanned preventive maintenance or.testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generators, separately for each diesel generator, by performing Surveillance Require-ments 4.8.1.1.2. a.4 and 4.8.1.1.2.a.5 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for each diesel generator which has not been successfully tested in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the diesel generators are already operating and loaded."

Restore at least one of the inoperable AC sources to OPERABLE status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-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 />.

Restore at least two offsite circuits and diesel generators EDG"1 and EDG"3 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss 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 />.

With diesel generator EOG"2 of the above required AC electrical power sources inoperable, demonstrate the OPERABILITY of the offsite AC sources by performing Surveillance Requirement 4.8.1.1.1 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 every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

If the diesel generator becomes inoperable as a result of any cause other than preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generators, separately, by performing Surveillance Require-ments 4.8. 1. 1. 2. a.4 and 4.8. 1. 1.2.a. 5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. "

Restore diesel generator EDG"2 to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or declare the HPCS inoperable and take the ACTION required by Specifications 3.5.1 and 3.7.1.1.

'I

" Thi s test i s required to be completed regardless of when the inoperab1 e di ese 1

~

~

~

~

~

~

generator is restored to OPERABLE status.

The provisions of Specification 3.0.2 are not applicable.

NINE MILE POINT - UNIT 2 3/4 8-2

ELECTRICAL POWER SYSTEMS

(

AC SOURCES AC SOURCES - OPERATING LIMITING CONDITIONS FOR OPERATION

3. 8.1. 1 (Continued)

'CTION:

e.

With diesel generator EDG~1 or EDG"3 of the above required AC electrical power sources inoperable, in addition to taking ACTION b or c, as appli-cable, verify within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that all required systems, subsystems,

trains, components, and devices that depend!

on the remaining OPERABLE diesel gen-erator as a source of emergency power are also OPERABLE;otherwise, 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 />.

f.

With both of the above required offsite circuits inoperable, demonstrate the OPERABILITY of three diesel generators, separately, by. performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 separately

for, each diesel generator within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> unless the diesel generators are already operating and loaded; restore at least one of the above required offsite circuits to OPERABLE status within.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-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 />.

With'nly one offsite circuit'restored to OPERABLE -status, restore at least two offsi+e circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from time of initial loss 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 follow-

. ing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A'uccessful test(s)'of diesel generator OPERABILITY per Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5, performed under this ACTION statement for the OPERABLE diesel generators, satisfies the diesel generator test requirements of ACTION statement a.

g.

With diesel generators EDG"1 and EDG"3 of the above required AC electrical power sources inoperable,'emonstrate the OPERABILITY of the remaining AC sources by performing Surveillance Requirement 4.8.1.1.1 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 every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter and Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for diesel generator EDG*2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.*

Restore at least one of the inoperable diesel generators EDG"1 and EDG"3 to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 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 COLO SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Restore both diesel generators EDG"1 and EDG*3 to OPERABLE status within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> s from time of initial loss 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 COLO SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

" This test is required to be 'completed regardless of when the inoperable diesel generator is restored to OPERABLE status.

The provisions of Specification 3.0.2 are not applicable.

NINE MILE POINT - UNIT 2 3/4 8-3

ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION " OPERATING LIMITING CONDITIONS FOR OPERATION 3.8.3. 1 (Continued)

ACTION:

a.

For AC power distribution:

With either Division I oi Division II of the above required AC distribution system not ener gized, reenergize the division within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-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 />.

2.

With Division III of 'the above required AC distribution system not energized, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

b.

For OC power distribution:

With either Division I or Division II of the above required OC distribution system not energized, reenergize the division within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 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 COLO SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

With Division III of the above required OC distribution system not energized, declare the HPCS system inoperable and take the ACTION required by Specification 3.5.1.

SURVEILLANCE RE UIREMENTS 4.8.3.1.1 Each of the above required power distribution system divisions shall be determined energized at least once per 7 days by 'verifying correct supply breaker alignment and by verifying no-bypass inoperability status indicator lights in the control room are lit."

4.8.3.1.2 Each of the above required power distribution switchgear -shall be determined energized at least once per 7 days by verifying the voltage on the panels.

• Which.would indicate a loss of power to one or more of the required MCCs, load

center, or panels.

NINE MILE POINT - UNIT 2 3/4 8-21 AN SS MS

ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION -

SHUTDOWN LIMITING CONDITIONS FOR OPERATION 3.8.3.2 As a minimum, the following power distribution system divisions shall be energized:

a.

For AC power distribution, Division I or Division II, and when the HPCS system is required to be OPERABLE, Division III, with:

1.

Division I consisting of:

a) 4160-volt AC bus b) 600-volt AC load center/MCCs/distribution panels c) 240/120-volt AC and 120-volt AC distribution panels, energized from inverter 2VBA".UPS2A or alternate supply 2.

Division II consisting of:

a) 4160-volt AC bus b) 600-volt AC load center/MCCs/distribution panels c) 240/120-volt AC and 120-volt AC distribution panels, energized from inverter 2VBA"UPS2B or alternate supply 3.

Division III consisting of:

a) 4160-volt AC bus b) 600-volt AC MCCs/distribution panels c) 240/120-volt AC and 208/120-volt AC distribution panels d)

HPCS inverter energized from Division III batteries b.

For DC power distribution, Division I or Division II, and when the HPCS system is required to be OPERABLE, Division III, with:

1.

Division I consisting of 125-volt OC switchgear, MCC, and distribu-tion panels 2.

Division II consisting of 125-volt OC switchgear, MCC, and distribu-tion panel's 3.

Division III consisting of 125-volt DC,distribution panels APPLICABILITY:

OPERATIONAL CONDITIONS 4, 5, and ".

" When handling irradiated fuel in the r~ d'or I iield;q.

NINE MILE POINT - UNIT 2 3/4 8"22 45 85 586

ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION DISTRIBUTION -

SHUTDOWN LIMITING CONDITIONS FOR OPERATION 3.8.3.2.(Continued)

ACTION:

a.

For AC power distribution:

2.

With less than 8ivision I and Division II of the above required AC distribution system energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the reactor building, and operations with a poten-tial for draining the reactor vessel.

With Division III of the above required AC distribution system not energized, declare the HPCS system inoperable and take the ACTION required by Specifications 3.5.2 and 3.5.3.

b.

For DC power distribution:

With less than Division I and Division II of the above required DC distribution system energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the reactor building, and operations with a potential for draining the reactor vessel.

2.

With Division III of the above required DC distribution system not energized, declare the HPCS system inoperable and take the ACTION required by Specifications 3.5.2 and 3. 5.3.

c.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.8.3.2.1 At least the above required power distribution system divisions shall be determined energized at least once per 7 days by verifying correct supp'ly breaker alignment and by verifying no-bypa~ inoperability status indicator lights in the control room are lit."

4.8.3.2.2 Each of the above required power distribution switchgear shall be determined energized at least once per 7 days by verifying the voltage on the panels.

~

~

~

" Which would indicate loss of power to one or more of the required MCCs, load

centers, or panel s.

NINE MILE POINT - UNIT 2.

3/4 8-23

.0

ELECTRICAL POWER SYSTEMS fN)AL ELECTRICAL E UIPMENT PROTECTIVE DEVICES PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITIONS FOR OPERATION 3.8.4.2 All primary containment penetration conductor overcurrent protective devices" shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With one or more of the primary containment penetration conductor overcurrent protective devices" inoperable, declare the affected system or component inoperable and apply the appropriate ACTION statement for the affected system and:

1.

For 13.8-kV circuit breakers, deenergize the 13.8-kV circuits by tripping the associated redundant circuit breaker(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the redundant circuit breaker~go be tripped at least once every 7 days thereafter.

(s~

2.

For 600 volt MCC circuit breakers, remove the inoperable circuit breaker(s) from service by opening the breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the inoperable breaker(s) to be in the open position at least once every 7 days thereafter.

Otherwise, 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 />.

b.

The provisions of Specification 3.0.4 are not applicable to overcurrent devices in 13;8-kV circuits which have their redundant circuit breakers

.tripped or to 600-volt circuits which have the inoperable circuit breaker discorinected.

'URVEILLANCE RE UIREMENTS 4.8.4.2 Each of the primary containment penetration conductor overcurrent pro-tective devices" shall be demonstrated OPERABLE:

a.

At least once per 18 months:

1.

By verifying that the medium voltage 13.8-kV circuit breakers are

-OPERABLE by selecting, on a rotating basis, at least 10X of the circuit breakers of each voltage level and performing:

" Excluded from this specification are those penetration assemblies that are capable of withstanding the maximum current available because of an electrical fault inside containment.

NINE MILE POTINT " UNIT 2 3/4 8-28

g 1.4 CI

~ 1.2 4 1.0 h

0.8 N

M

~ 0.6 au OA g 0.2 m 0.0 0

10 20 30 40 SERVICE LIFE, YEARS

BtTIN~BISE-%~

Bases Figure B3/4.4.6-1 Fast Neutron Fluence (E>l MeV) at b T as a

Function of Service Life at 90K of RATED THERMAL POWER and 9(C Availability NINE MILE POINT - UNIT 2 B3/4 4-7

qadi;

Mine Nile Point Unit 2

PSAR Pene-tration No.

System Desi(!nation Mv supply to IICtuatorS fOr

?CPS@AOV109 GDC or Reg

! I)ide 56

!!o Air/Mm ESP SY I! eE Fluid Size gin)

P SAR Ar'range 14 e ut Pigure11) 1

6. 2-70 S)h-

'43A Location 0!

Va! ve

4. n s l. I Out P f. ~

CI la 1'ut" I'Is Le<I< ""

Pi Oui Xsn!

11 0 '

'I1, NumbeI

,MPC

'CP Sv<<IOV'l33

.II << 'I Type Globe Check Oper-ate!I SOV N/A Elec.

process N/A N/A Actuator Node Valve(>>

Closed closed Closed closed Closed Closed Closed Closed Position Normal Post-Pover

<a)

Shutdovn Accident Pailure Isola-tion Signal (e)

B<<PIYrRN<<Z Reverse flov Closure Time

<e a) 5 N/A TABLE i6. 2-56 (Cont)

Poser Source

<7)

Div II N/A Note Mv supply to actuators for'CPSvAOV107 No 1

6. 2-70 Sh,.

N 38 2 C P S v 8 r!V 1 3 2

?c'vv50 Globe Check SOV N/A Elec.

Process N/A N/A Closed Closed Closed Closed Closed closed Closed Closed B,Pe Y,RN ~ Z Never'se flou M/A Div II

! N/A 2 e 2-9 8 A 2-98B BHB relief va.lve d.ischarge to suppression pool RHR relief valve discharge to suppression pool 56 Yes Mater 3

6.2-70 Outside 207'-6" A:!o < n u )

Sn.

38 56 Yes Mater 3

6.2-70 Outside 89'-;

Sh.

38 2CSLvBV123

?CSLeRV105

?RH I(IIV6 "I 1 2!! I{ S'!I V 1 1 0

? B!!SvHV 139

?!!!S"!IV20A 2CS HvRV '! 19 IIV 113 I, II I uv6!H

!I'!61C RV208 E21-P031 L'2 1-Pp 1 8 E12-P088A E12-F005 H12-P030 L'I2-P 0 2 5 A E? 2-F035

! ? 2-I'0 I I!

.'! 2"'0 HD 1."12"I"088C E!2-!"0258 Relief M/A N/A Valves Relief N/A N/A Valves N/A N/A N/A N/A N/A N/A N/A None None N/A N/A'/A 25 N/A

'5'-9 9 A Hydraulic unit from recirc flov control valve HYV 17A (drain line) 56 No Hy-3/44 6, 2-70 draulic sh.

39 Outside Inside Q ~

Q II 0 I QII

/A No<a)

?"

"OVHBA OVQ?A Globe Globe SOV Elec.

SOV Elec.

N/A N/A Open Open Closed Closed closed.

Closed Closed Closed IMP<<RN2Z B<<P,B'N,Z 5

5 Div I 26 Div II 2-99B Hydraul).c unit to recirc flov control valve HYV 17A (open lir.e) 56 No Hy-1

6. 2-70 draulic Sh.

39 Outside Inside Q I QII PI Qu N/1

!H)<se) 2BCSvSOV67A 2RCSmSPV81A Globe Globe SOV SO" Elec.

L"lec.

N/A M/A Open Closed Closed Closed B,F ~ RN,Z 5

Open Closed I Closed Closed B, P, BN, Z 5

Div I 26

~2e Div II Z-99C Hydraulic unit to recirc flov control valve HYV 17A (pilot line) 56 No Hy-1

6. 2-70 dranlic Sh.

39 Outside Inside p I pu PI pu V/A Po<-1) 2RCSVSOV66A 2 B C S v 8 0 V An II Globe Globe SOV Elec.

SOV Elec.

N/A N/A Open Open Closed Closed

Closed, Closed B,P,RN,Z 5

Closed Closed B,P,RN,Z 5

Div I 26 Div II TI APERTURE CARD Amendment 26 15 of 244 Nay 1986 gg~ go g'ada 7-8/

mme N).le Po1nt Bn1t 2 FSAR TABLE. 6. 2-56 (Cont) 1.on fl 2-67 z-68 2.-6 g 2-70 2-71 Z-72 S 7ste III Desi

>><In a' O> le 8" v 8 I;i.em F'id Sime

>rr anI e

'. II1 ourn<>>

->>n Locat'.on V>>.v, L acth Of Ins Pj

.on 0 ut.

0 P <1 Illarg CO II'nti Nu ~her.

-OU34L NSGUOSB Type G lohe G l.ohe Oper ator SOV SOU L'lec Elec.

N/A N/A Actuator Node Primary Secondary Normal

< a)

Open Open Post-Accident Shntdovn Closed Closed Open Open Valve<~)

Pos).t1on Pover Failure Closed Closed Is ola-tion Signal

<m)

B~F,RN B,F,RN Clomure T1me

<a a)

,Z 5

,Z 5

Pov er Source

<1)

Div II Div II Notes Z-73 RHS relief valve di.

bar'ge

<0 s u l;pr e s s j.o r.

pool No f'.ater 6.?-70 33 Out

.6"

>> oIf 1 0 ni? 0 E12-F036 RV N/A N/A 1?-F025C N/A N/A N/A N/A None N/A N/A g 5 2-74 z-75 2-76

?,-77 Z-78 z-Vg

'Z-80 Z-81 2-82 Flanged spare Capped spare Capped spare

, capped spare Capped spare Capped spare Snent fuel 56 pool cooling capped spa.re Capped spare No 1/2 1/?

1/2 Pater 1 1/:

"-70

".h nuts...

C C

2SFCNV20.l

') 8("t

?DI Glohe Glohe Nanual Nanual Nanual Nanual.

N/A

,N/A TI APERTURE CARD LC LC Closed Closed Closed N/A Closed Closed Closed N/A N/A N/A N/A Il/A Amen)ment '26 13 of 24