Revision as of 06:06, 6 July 2018

Text

Proposed Tech Specs Removing Obsolete Info to Provide Consistency Between Units 1 & 2 to Provide Consistency with Std Tech Specs,To Provide Clarification & to Correct Typos
	ML17335A362
Person / Time
Site:	Cook
Issue date:	12/03/1998
From:	; INDIANA MICHIGAN POWER CO.
To:	;
Shared Package
ML17335A360	List: ML17335A359; ML17335A362;
References
	NUDOCS 9812080050
	Download: ML17335A362 (142)
	v • d • e

ATTACHMENT 2A TO AEP:NRC:0433Q TECHNICAL SPECIFICATIONS PAGES'ARKED TO SHOW PROPOSED CHANGES UNIT 1 REVISED PAGES 3/4 0-3 3/4 3-2la 3/4 4-38 3/4 4-40 3/4 7-15 3/4 9-1 3/4 9-13 5-6 5-7b 6-4 9812080050 981203 PDR ADOCK 05000815 P PDR

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3 4.0 APPLICABILITY SURVEILLANCE RE UIREMENTS 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:

ASME Boiler And Pressure Vessel Code and applicable Addenda Required frequencies for terminology for inservice performing inservice inspec-inspection and testing criteria tion and testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Yearly or annually At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 366 daysc.The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d.Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e., Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

4.0.6 endments$00, 107 and 108 granted extensions or cer a n su eillanck require to be p rformed n or before July 31, 1987, and til the end of e Cycle-10 refu ling outage.or these speci c surve lances der thi section, the sP cified ime interva requir d by Spe ificatio 4.0.2 w ll be termin d with the new i tiation date es blished y the s eilla ce date during the Unit 1 987 refuelin outa e.4.0.7 en e e ann surveys anc to be perfo ed on o before pril 1, 1989, until the end of the Cyc 10-11 fueling utage.For th e speci c su eillan es under this ext sion, t p speci ed ti interv s req ired b Specifi ation 4.2 will bq dete ned wx h the n q ini ation establis d by the urveillance date durin the Un'1 89 refueling'out te COOK NUCLEAR PLANT-UNIT 1 3/4 0-3 AMENDMENT NO.ggg<</gal<<144 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 6.MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS MINIMUM TOTAL NO.OF CHANNELS CHANNELS APPLICABLE CHANNELS TO TRIP OPERABLE MODES ACTION a.Steam Generator Water Level-Low-Low 3/Stm.Gen.2/Stm.Gen.2/Stm.Gen.any Stm.Gen.1,2,3 14 b.4 kv Bus Loss of Voltage Pump Start Valve Actuation (Both trains)c.Safety Injection d.Loss of Main Feedwater Pumps 7.TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS 3/Bus 2/Bus 2/bus (Tl I A-Train B;T11D-Train A)2/bus on (Tl IA 8c T11B or 2/busses Tl lc&T11D), 2/Bus 1,2,3 1, 2, 3 1,2 14 18 18 a.Steam Generator 3/Stm.Gen.Water Level-Low-Low b.Reactor Coolant Pump 4-1/Bus Bus Undervoltage 8.LOSS OF POWER a.4 kv Bus Loss of 3/Bus Voltage 2/Stm.Gen.any 2 Stm.Gen.2/Bus 2/Stm.Gen.2/Bus 1, 2, 3 1,2,3 1,2,3,4 14 19 14 b.4 kv Bus Degraded 3/Bus Voltage (Ttl 8 TraInlj at t)-Tr~'~h)2/Bus 2/Bus I, 2, 3, 4 (Ttth-TrRI'~Sj (rttA-7~:~8>

Yttb-frai~h) sub-T~t~h) 14 COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-21a AMENDMENT$C,~, 153 4.4.$2.1 Both Reactor Vessel head vent paths shall be deaanstrated OPI',RAB least once per 18 months by: 1.Verifying the ccmaaa manual isolation valve in the Reactor vessel head vent is sealed in the open position.2.Cycling each of the remotely operated valves in esoh path through at least one complete cycle of full travel tree the Control RocNa while in Bodes 5 or 6.3~Verifying floe through both of the Reactor Vessel head vent paths dur ing vaat~operation, awhile in Bodes 5 or 6.AeQ.lance equiraaen to demons ate the oper ility of eac eactor V el head ve path vQ,l e performed he next time e unit enters H ES 5 or 6 Ilying th issuance of his Technical spec1r1 tion, and or the app prtate pl prccednrea p ee been vrit ten.D.C COOK-UNIT l 3/4 4-38 4.4.$g.2 Both suriser s~eaa space veat paths shall be deneastrated 0?at least oace pec 18 cenths byc f.Verifying the comma aaaual isolation valvo ia the Pressurizer st04$space vent is sealed ia the open yoaitioao Cyoliag each of the reaetely operated valves in each yath through at least oae ooaplete cycle of full travel froa tho Coatrol Rooe~e in Hades 5 or 6.3.Verifying flam through both of the Pressurizer steaa sycLce rent yaths 4uriag veatiag operation,~le in}fedos 5 oc'.illanoe req ants to 4 nstrate t~operabi y of each Pr urer ste space veat th Wll b perfo the next t the t eaters ES 5 oc'f loviag the ssuaaoe this Techaiy Speed.fkcati9, aad after e appropr te Plant rocedures hag@been e'it tea./D~C o COOK 3/4 4-40 Aaaadmeat ceo.98 C

3.7.3.1 OOL G V TE 0 0 ERATIO a.At least two independent component cooling water loops shall be OPERABLE.ML b.At least wnce component cooling water flowpath in support of Unit 2 shutdown functions shall be available.

Specification 3.7.3.1.b-At all times when Unit 2 is in MODES 1, 2, 3, or 4.~CT~IO V&bh Specification 3.7.3.1.a is applicable:

Vith only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .When Specification 3.7.3.1.b is applicable:

With no flowpath to Unit 2 available,'return at least one flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 2 and return at least one flow path to available status within the next 60 days, or have Unit 2 in HOT STANDBY within the next 12,hours and HOT SHUTDOWN.within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least two component cooling.water loops shall be demonstrated OPERABLE'.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equip-ment that is not locked, sealed, or otherwise secured in position, is in its correct position.b.At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety In]ection test signal.c.By verifying pump performance pursuant to Specification 4.0.5.d.At least once per 18 months during shutdown, by verifying that thehcross-tie valves can cycle full travel.Pollowing cycling, the valves will be verified to be in their closed positions.

COOK NUCLEAR PLANT-UNIT 1 3/4 7-15 AMENDMENT NO.447, 434, 444, 164 r I I G 0 ORO CONCE LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbol'ted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: a.Either a K ff of 0.95 or less, which includes a 1%Jc/k.eff conservative allowance for uncertainties, or b.A boron concentration of greater than or equal to 2400 ppm, which includes a 50 ppm conservative allowance for uncertainties.

~CT~O With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes~+and initiate and continue boration at greater than or equal to 10 gpm of 20,000 ppm boric acid solution or its equivalent until K fis reduced eff to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2400 ppm, whichever is the more restrictive.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLPSCE RE UIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall'oe determined prior to: a.Removing or unbolting the reactor vessel head, and b.Withdrawal of any full length control rod in excess of 3 feet from'its fully inserted position.4.9.1.2 The boron concentration of, the reactor coolant system and the~~~it~Mx i 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .0ACe.pe,r*The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.*~For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

D.C.COOK-UNIT 1 3/4 9-1 AMENDMENT NO.

~~c'~l 3,9.12 The spent fuel storage pool exhaust ventilation system shall be OPERABLE.Whenever irradiated fuel is in the storage pool.a~With no fuel storage pool exhaust yentilati.on system OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one spent fuel storage pool exhaust ventilation system is restored to OPERABLE status.*b.The provisions of Specificati'ons 3.0.3 and 3.0.4 are not applicable.

'4C 4.9.12 The above required fuel storage pool ventilation system shall be t demonstraied OPERABLE: a.At least once per 31 days by initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.At least once per 18 months or (1)after any structural.

maintenance on the HEPA filter or charcoal adsorber housings, or (2)following painting, fire or chemical release in any ventilation zone communicating with the system, by: 1.Deleted.2.Verifying that the charcoal adsorbers remove>99%of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation system at a flow rate of 30,000 cfm+10%.5ov oor y e ovxfli(h lovi14n c~(aN8~+7*The crane bay roll-up door and the runua&g-room-rcrH~~oer may be opened under administrative control during movement of fuel within the storage pool or crane operation with loads over the storage pool,~Shared system with D.C.COOK-UNIT 2.~~This does not include the main load block.For purposes of this specification, a deenergi"ed main load block need not be considered a load.D.C.COOK-UNIT 1 3/4 9-13 Amendment No.1~4 I

5.0 DESIGN FEATURES 5.6 FUELSTORA E Continued Region 1 is designed to accommodate new fuel with a maximum nominal enrichment of 4.95 wt%U-235, or spent fuel regardless of the discharge fuel burnup.2.Region 2 is designed to accommodate fuel of 4.95%initial nominal enrichment burned to at least 50,000 MWD/MtU, or fuel of other enrichments with equivalent reactivity.

3.Region 3 is designed to accommodate fuel of 4.95%initial nominal enrichment burned to at least 38,000 MWD/MtU, or fuel of other enrichments with equivalent reactivity.

The equivalent reactivity criteria for Region 2 and Region 3 is defined via the following equations,'nt+raphicalbpdepieted-i~re-5

~Minimum Assembly Average Burnup in MWD/MTU=-22,670+22,220 E-2,260 E+149 E3 Minimum Assembly Average Burnup in MWD/MTU=-26,745+18,746 E-1,631 E+98.4 E3 Where E=Initial Peak Enrichment 5.6.1.2: Fuel stored in the spent fuel storage racks shall have a maximum nominal fuel assembly enrichment as follows: Description Maximum Nominal Fuel Assembly Enrichment Wt.%U-235 1)Westinghouse 2)Exxon/ANF 3).Westinghouse 4)Exxon/ANF 15 x 15 STD 15 x 15 OFA 15 x 15 17 x 17 STD 17 x 17 OFA 17 x 17 V5 17 x 17 4.95 4.95'.95 4.95 COOK NUCLEAR PLANT-UNIT 1 Page 5-6 AMENDMENT SV, 436, 4N, 44&, 213 eg ons I r QXC4.Q~~~COO 1>eeoc I r~'CC'CCC:: SCHMO C L I OOOO r r r SXO OO Ac)I"-I oUR.P 9QMiN I JH""O ABLE 3URN'DCM+H l 2.0~S.O!NGl4.-IRIPIMEHT.

aU-235 CWGK i'~

6.0 ADMINISTRATIVE

CONTROLS**""Q**'" 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1)the Plant Radiation Protection Manager, who shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (2)the Shift Technical Advisor, who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents and, (3)the Operations Superintendent, who must holMr-htt~eld-a-Senier-Operatm-license as specified in Section 6.2.2P..e va.tig)t.4 6.4 TRAININGt;<;6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.5.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION 6.5.1.1 The PNSRC shall function to advise the Site Vice President/Plant Manager, or designee, on all matters related to nuclear safety.COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Department Superintendents, or supervisory personnel reporting directly to the Site Vice President/Plant Manager, Assistant Plant Managers or Department Superintendents.

The membership shall represent the functional areas of the plant, including, but not limited tn Operations, Technical Support, Licensing, Maintenance and Radiation Protection The PNSRC membership shall consist of at least one individual from each of the areas designated.

All members, including the Chairman and his ahemates, the members and their alternates, shall be designated by the Site Vice President/Plant Manager.PNSRC members and alternates shall meet or exceed the minimum qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions.

The nuclear power plant operations individual shall meet the qualifications of Section 4.2.2 of ANSI N18.1-1971 except for the requirement to hold a current Senior Operator License.The operations individual must hold or have held a Senior Operator License or have been certified for equivalent senior operator knowledge at Cook Nuclear Plant or a similar reactor.The maintenance individual shall meet the qualifications of Section 4.2.3 of ANSI N18.1-1971.

COOK NUCLEAR PLANT-UNIT 1 Page 64 AMENDMENT 49, 4B)48K, 454)486)4Q)212 ATTACHMENT 2B TO AEP:NRC:0433Q TECHNICAL SPECIFICATIONS PAGES MARKED TO SHOW PROPOSED CHANGES UNIT 2 REVISED PAGES 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 0-3 0-4 3-11 3-20 3-30 3-31 3-34 3-44d 3-47 4-13 4-14 4-33 4-35 4-37 5-4 5"8 6-12 6-14 6-47 7-12 7-13 7-16a 7-20 8-9 8-13 8-15 9"12 5-6 5-8 6-4 3 4A APPLICABILITY SURVEILIANCE RE UIEUUKNTS 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 xequired'y the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follow's ia these Technical Specifications:

ASME Boiler And Pressure Vessel Code and applicable-Addenda terminology for inservice ins ection and testin criteria Required frequencies for performing inservice inspec-tion and testin activities

'o d.Weekly At least once per 7 days Monthly At least once per 31 days~rterly or every 3 months At least once per 92 days Semiannually or every 6 months , At least once per 184 days Yearly or annually At least once per 366 days The provisions of Specification 4.0.2 axe applicable to the above'required frequencies for performing insexvice inspection and testing activities.

Performance of the above inservice inspection and testing mctivities shall be in addition to other specified Surveillance Requirements.

e~4.0.6 Nothing in the ASME Boiler and Pressure Vessel Code shall be con-stxued to supersede the requirements of any Technical Specification.

'ba, en g an e e ens ons or cer n surve ances required to be per ormed gn or bef re Mar h 31, 1 86, unti the end of the Cyc~e 5-6 efueliilg outage For'se sp cific s eillanc)under this ectio the s$cified time in exvals equired y Speci catio 4.0.2 ill be determi ed with the ne nitia)on date establis ed b the urveil ance da e durin the Unit 2 1984 refu outa e.4.0.7 Amendments 97 and 99 granted extensions for certa n re uired to e perform d on or be re July~1, 1988, until the end of th Cycle 6-refueling utage.Fo these h ecific s eillan es unde this sec on, the s ecified t e inte ls requir d by Speci cation 4..2 will be determine with th new init ation te establi+ed by the surveilla ce date d ng the it 2 19 refueling outage.COOK NUCLEAR PLANT-UNIT 2 3/4 0-3 AMENDMENT NO.7 g, g7, 1 31 1't) 3 40 LXC SURVEILLANCE 4.0.8 By specific re erence to s sect on, ose surve ances which must be p rformed on or before hu t 13, 1994, and are designed as-month r 36-mo surve lances or re ed as ou ge-related s eillan s unde the pro sions o Specifi ation 4..5)may be del ed unt the en of the cl'e 9-10 efueli outage.For these spec ic s eillanc s under this se tion, speci ied time inte ls re red by pecifica ion 4.0.will b determ ed with the ne initiat on date stablish d by the surveill ce dat during th'e Unit 2 1994 efuelin outa e.4.0.9 By specific re erence to s sect on, ose surve ances whi.ch mus be performed on before eptemb r 7, 19 4, and a desi ted as 1 month h~eillan es may e dela d unti just pr or to re eload n the tJqit 2 Cy e 9-10 efueli outa COOK NUCLEAR PLANT-UNIT 2 3/4 0-4 AMENDMENT NO.~, 166 n O T B.E 3-CTOR I SYST ST E 0 S V C U FUNCTIONAL UNI 1.Manual Reactor Trip A.Shunt Trip Function B.Undervoltage Trip Function 2.Power Range, Neutron Flux 3.Power Range, Neutron Flux, High Positive Rate 4.Power Range, Neutron Flux, High Negative Rate 5.Intermediate Range, Neutron Flux CHANNEL~GEE N.A.N.A.N.A.N.A.CHANNEL C 0 N.A.N.A.(20)IM(I)and Q(608)R(6)R(6)'(6,8)CHANNEL FUNCTIONAL

~AT.~S/U(1)(10)S/U(1)(10)M and S/U(l)S/U(l)MODES IN WllICH SURVEILLANCE R U R 3*, 4*, 5*3*, 4*, 5*1, 2 and*1, 2 1, 2 1, 2 and*a W o 6.Source Range, Neutron Flux 7.Overtemperature AT 8.Overpower AT 9.Pressurizer Pressure--Low 10.Pressurizer Pressure--High ll.Pressurizer Water Level--High 12.Loss of Flow-Single Loop R(6,14)R(9)g R(9)e" Re RP aQ R(8)M(14)and S/U(l)2(7), 3(7), 4 and 5 1, 2 1, 2 1, 2 1,~2 1, 2 r>t0 TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT MINIMUM TOTAL NO.CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION 7.TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS t fl a.Steam Generator Water Level--Low-Low 3/Stm.Gen.2/Stm.2/Stm.1, 2, 3 Gen.any Gen.2 Stm.Gen.14*b.Reactor Coolant Pump Bus Undervoltage 8.LOSS OF POWER I 4-1/Bus 1, 2, 3 19*a.4 kV Bus Loss of Voltage 3/Bus 2/Bus 2/Bus 1, 2, 3, 4 14*b.4 kV Bus Degraded Voltage 9~MANUAL 3/Bus (Qi A-(nfl'i 721 l)VIE+Qj 2/Bus 2/Bus 1, 2, 3, 4 14*(T7lb TmlnBj (T2lA Tm'+Si Tz, I D-Twi<<)T't<D-<~'<4)a.Safety Injection (ECCS)Feedwater Isolation Reactor Trip (SI)Containment Isolation-Phase"A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System 2/train 1/train 2/train 1, 2, 3, 4 18 b.Containment Spray Containment Isolation-Phase"B" Containment Purge and Exhaust Isolation Containment Air Recirculation Fan 1/train 1/train 1/tra'in 1, 2, 3, 4 18 c.Containment Isolation-Phase"A" Containment Purge and Exhaust Isolation 1/train 1/train 1/train 1, 2, 3, 4 18 d.Steam Line Isolation 2/s team 2/s team line (1 line (1 per train)per train)2/opera-1,'2, 3 ting steam line (1 per train)20 COOK NUCLEAR PIANT-UNIT 2 3/4 3-20 AMENDMENT NO g7 j f gg~137' 3/4..LIMI'HNG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTfONAL UNIT l.SAFETY INIECIION, TURBINE TRIP, FEEDWATER ISOLATION.

AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS CHANNEL CfiANNEL CHANNEL FUNCTIONAL CHECK CALI RAIIDH IEEI'RIP ACTUATING DEVICE OPERATIONAL MODES IN WHICH SURVEILLANCE SQUIRED a.Manual Initiation b.Automatic Actuation Logic c.Containm<<nt Pressure-High d.Pressurixcr Prcssurc-Low c.DifFerential Pressure Bctwccn Steam Lines-High f.Stcam Lue Pressure-Low 2.CONTAINMENT SPRAY a.Manual Initiation b.Automatic Actuation Logic c.Concunmcnt Pressure-High.High 3.CONTAINMENT ISOLATION a.Phase'A Isolation N.A.N.A.N.A.N.A.M(3)N.A.N.A.N.A.N.A.Sce Functional Unit 9 M(2)N.A.M(3)N.A.Scc Functional Unit 9 M(2)N.A.I, 2, 3, 4 1,2,3 I, 2, 3 I, 2, 3 1,2,3 1, 2, 3, 4 1,2,3 I)Manual 2)From Safety Injection Automatic Actuation Logic'.Phase B'solation N.A.N.A.Scc Functional Unit 9 M(2)N.A.1, 2, 3, 4 I)Manual 2)Automatic Actuation Logic 3)Containmcnt Pressure-f{igh.High N.A.S Scc Functional Unit 9 M(2)M(3)N.A.N.A.1,2,3,4 1, 2', 3 COOK NUCLEAR PLANT-UNIT 2 Pago 3/4 3-30 AMENDMENT 34, 434, 4K, 158 C J 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT c.Purge and Exhaust Isolation CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK CALIBRATION TEST l TRIP ACTUATING DEVICE OPERATIONAL TEST MODES IN WHICH SURVEILLANCE

~RE UlRED I)Manual 2)Containment Radioactivity

-High Scc Functional Unit 9-N.A.1,2,3,4 4.STEAM LINE ISOLATION a.Manual-Sec Functional Unit 9--b.Automatic Actuation Logic c.Containmcnt Pressurc-High-High d.Steam Flow in Two Steam Lines-High Coincident with T,,-Low-Low c.Steam Linc'Prcssurc-Low S.TURBINE TRIP AND FEEDWATER ISOLATION a.Steam Generator Water Level-High-High N.A.N.A.M(2)M(3)N.A.N.A.N.A.N.A.N.A.I, 2, 3 I, 2, 3 I, 2, 3 1,2,3 1,2,3 6.MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.Stcam Generator iVater Level-Low-Low b.4 kV Bus Loss of Voltage c.Safety Injection d.Loss of Main Feed Pumps N.A.N.A.N.A.N.A.M(2)N.A.N.A.I, 2, 3 I, 2, 3 I, 2, 3 1,2"%teart)vhion~~tnicaLSpcciiieatiea-4A)A Ltpplicable.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-31 AMENDMENT aI,&, 434, 434, 4W, 459, 16S P k0 3/4.3.3 KOH~~OEIHC GiSTM~LTION LXKKTZSC COHDZTZOS FOE QPKLLTIQS 3.3.3.1 Tha r a~~~mcraiarz~~~~mantas~mn

~mn~ahcncn Tahoe 3.~ahaU.be QPKLL'.KZ~~~/~sespo~W~the)spec'~~~4 CXEZLZZT: ha abc'Tabb<3.~.JCXZOE: a.V'A a~~non'~~channel akaxa/a~aespoins czcaa~tha vaLaa shee w Table 3.M, adgnos~aetpo~to Mahdi tha Xfad.t W~4 boers or de~a tha channel fmoperabIa.

bo%~A one oT ncce t~~~celt?ftoT~

ch@nL'e ezabla, rake the~H ebcnca fa Tabl.a 3.~.c.The pmvta~of Spe~caeLana 3.0.3 cad 3.0.4 are not apylf cabLa.4.3.3.1 Each zmLfaahan zoon" metal c1nnuaak ahaLl be as she fx'eqnencMs abner ia TabLa 4.3 3.~i~eh Q ca1~H4ng-osage D, C.CQOK~2 3/4 3 34 A$EtldQolTC)f0~43 ri O O c<ri Xuasauumr:

ir 1.Steam Generators 1 and 4 Level V 0 0 0 G V~G V E S¹LE~0 41 Cabinet l and LSI Cabinet 4 ClhNNEL~CWCQ CllANNEL C 0 2, Steam Generators 2 and 3 Lovel LSI Cabinet 2 and LSI Cabinet 4 3.Steam Generators 1 and 4 Prossure LSI Cabinet 4 and LSI Cabinet 5 Steam Generators 2 and 3 Pressure LSI Cabinet 4 and LSI Cabinet 6 5, Reactor Coolant Loop 4 Temperature (Cold)6.Reactor Coolant Loop 4 Temperature (llot)LSI Cabinet 4 and LSI Cabinet 5 LSI Cabinet 4 and LSI Cabinet 5 Ry)7, Reactor Coolant Loop 2 Temperature (Cold)LSI Cabinet 4 and LSI Cabinet 6 8., Roactor Coolant Loop 2 Temperature (llot)LSI Cabinet 4 and LSI Cabinet 6 eg 0 9.Pressurizer Lovel.10.Raactor Coolant System Pressure ll.Charging Cross-Flow Between Units LSI Cabinet 3 LSI Cabinet 3 gorridor Kiev.587'/a R'2.Source Range Neutron Detector (N-23)LSI Cabinet 4 n/a*Cbarging Gross-1'low between Units is an instrpment common to'both Unit 1 and 2, Tbis surveillance vill only bo conducted on an interval consistent Mith Unit 1 refueling.

wu hl I r>

BL 3-0 OS-C Ol TORING S E AT 0 0 V U E E CJNNHEL~CEEC 1.2~3~4, 5.6.7~8.9.10'1.12~13.14.15.16.17~1B.Containment Pressure Reactor Coolant Outlet Temperature

-Tgp)(Hide Range)Reactor coolant Inlet Temperature

-T<<,p (Hide Range)Reactor Coolant Pressure-Hide Range Pressurizer Water Level Steam Line Pressure Steam Gei>orator Hater Level-Narrow Range BHST Water Level Boric Acid Tank Solution Level Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator-Limit Switches PORV Block Valve Position Indicator-Limit Switches Safety Valve Position Indicator-Acoustic Monitor Incore Thermocouples (Core Exit Thermocouples)

Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

Containment Sump Level Containment Hater Level M M M M M M M M M M M M M M M M(2)M M R+R R R R R R(1)%'(3)P R Partial range channel calibration for sensor to be performed below P-12 in MODE 3.(2)(3)Hith one train of Reactor Vessel Level Indication inoperable, Subcooling Margin Indication and Core Exit Thermocouples may be used to perform a CHANNEL CHECK to verify the remaining Reactor Vessel Indication train OPERABLE.Completion of channel cal'ibration for sensors to be.performed below P-12 in MODE 3.COOK NUCLEAR PLANT-UNIT 2 3/4 3-47 AMENDMENT NO.'9&E 9S,

~>

CD n CD CD 7C 1ST SAMPLE.1NSPECTlON TABLE 4.4-2 STEAM GENERATOR TUBE tNSPECTION 2ND SAMPLE lNSPECTION 3AD SAMPLE INSPECTION Saintito Sire A cauli Acticin ltcituircd Result Acriun tt'oquircd A cauli Aclion Acquucd A minunum ol S Tubes tier S.G.C-1 C-2 C-3 Nuiie Ptua dclcclive lubca and lnspecl edititionat 2S lubes tn llus S.G.tnstiect all tuties rn ttua S.G..l)lug Ue locllve tubes ailil tnsticct 2S lubes in cacti otlicr S.G.Promtll nultticallon to NAC pursuant to specilication 6 9.1 NIA C-1 C-2 C-3 Atl alber S Giseic C l Same S.G.s C-2 but no additional S.G.aie C-3 Additional S.G.ia C-3 NIA Nnne Phag'detective tubes aiid initiccl aitititianat 4S tubes ui ltws S.G.Pciloim aclioii lor C-3 result ol lull aamtite None Pcilorm ac(ian tbr C-2 result ol second tnspoct all lubea in each S.G.and pkry d elective tubes.Prompt nolilicatlon lo NRC pursuant lo specilicalian G.Q.I N/A NIA C-1 C-2 NIA NIA NIA NIA NIA None NIA NIA NIA NIA NIA Ptug detective luliea Pcrloim action tor C-3 result ol lust samtite 3 N<Shore H la tlie number ol stcam ttaneratora tn lhe unit, and it ls lhe number ol steam tteneiatora

<<specter iluitnQ Jn tnrtioctlon C

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLVPP SYSTEM LLQCAGE L~QCAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System, leakage detection syseems shall be OPERABLE: a.One of the containmene atmosphere particulate radioactivity monitoring'hannels (ERS-2301 or ERS-2401), b.The containment sump level and flow monitoring system, and c.Either ehe containmene humidity monitor or one of the containmene atmosphere gaseous radioactivity monitoring channels (ERS-2305 or ERS-2405).

APPLICABILITY:

MODES 1, 2, 3 and 4 ACTION'ith only two of the above required leakage detection systems OPERABLE, operation may coneinue for up to.30 days provided grab samples of the containmene atmosphere are obtained and analyzed at lease once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when~~~~ehe required gaseous and/or particulate radioactivity monieoring channels are inoperable; otherwise, be in at lease HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN wiehin ehe following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonserated OPERABLE by: a~Containment atmosphere particulate and gaseous (if being used)monitoring system-perfonnance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.Containmene sump level and flow monitoring syseem-performance of CHANNEL CALIBRATION ae least once per 18 months, c~Containmene humidiey monitor (if being used)-performance of CHANNEL CALIBRATION ae lease once per 18"months.

COOK NUCL<<R PLANT-UNIT 2 3/4 4-14 AMENDMENT NO.W,~, 159 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE<QUIRL<MI<

NTS ('/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued With PORVs and block valves not in thc same line inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the valves to OPERABLE status or close and dc-energize the associated block valve and place the associated PORV in manual control in each respective line.Apply the portions of ACTION c or d above, relating to the OPERATIONAL MODE, as appropriate for two or three lines unavailable.

h.The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.11.1 In addition to the requirements of Specification 4.0.5, each PORV shall bc demonstrated OPERABLE: At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and At least once per 18 months by operating the PORV through one complete cycle of full travel during MODES 3 or 4, and c.At least once per 18 months by operating solenoid air control valves and check valves in PORV control systems through one complete cycle of full travel, and At least once per 18 months by performing a CHANNEL CALIBRATION of thc actuation instrumentation&

4.4.11.2 Each block valve shall bc demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, c, or d in Specification 3.4.11.4.4.11.3 Deleted.f-'Feehnieal-Speei6eatie licable.COOI(NUCLEAR PLANT-UNIT 2 Page 3/4 4-33 AMENDMENT 4W,~, 4W, 444, 196 i t

'~4~~4.4.12.1 Both Reactor Vessel head vent paths shall be deaanstrated OPE least once per 18 months by: Verifying the common manual isolation valve in the Reactor vessel head vent is sealed in the open position.2.Cycling each of the remotely operated valves in each path through at least one complete cycle of fulL travel from the Control Room while in Nodes 5 or 6~3.Verifying floe through both of the Reactor Vessel head vent paths during venting operation, awhile in Nodes 5 or 6.Surveillan e requi nts to d nstrate the, operabilit of each eactor Ves el head ve t path be perfoH the next ime the u t ters NOD 5 or 6 fol ing the suance of Technic Sp fication, and after he appropr te Plant p cedures ha e been m'i ten.D.C.COOK-terr 2 3/4 4-35 amencunent Wo.65 r>

4.4.12.2 Beth Pr surizer steam space vent paths shall be demonstrated OPBKLBL at least once per 18 months by: 1~Verifying the common manual isolation valve in the Pressurizer steam space vent is sealed in the open position.Cyoling each of the reaetely operated valves in each path through at least one complete cycle of full travel from the Control Room while in Modes 5 or 6.3.Verifying flow through both of the Pressurizer steam space vent paths during venting operation, while in Modes 5 or 6.r I Surveill+ce requireme to demo trate the pdrability of h Pressurizer steam sp e vent pa will be formed the n time He unit en ers MODES 5 or 6 fol ing the i uance of t echnic Speoifgcation, an after th appropria Plant proc es have en writ t~~0.C.COOK-UHZT 2 3/4 4-37 Amenchnent No.65 EMERGENCY CORE COOLING SYSTEMS~~SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: a.At 1'east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the following valves are in the indicated positions with power to the valve operators removed: Valve Number Valve Function Valve Position a.IMO-390 b.IMO-315 c.IMO-325 0 d.Q-262*e.IN-263*f.IMO-261*g.ICM-305*h.ICM-306*a.RWST to RHR b.Low head SI to Hot Leg c.Low head SI to Hot Leg d.Mini flow line e.Mini flow line f.SI Suction g.Sump Line, h.Sump Line a.Open b.Closed c.Closed d.Open e.Open f.Open g.Closed h.Closed b.At least once per 31 days by verifying that each valve (manual, power operated or automatic}

in the flow path that is not locked, sealed, or~other se secured in position, is in its correct position.c.By a visual inspection which verifies that no loose debris (rags, tx'ash, clothing, etc.}is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.

This visual inspection shall be performed:

1." For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2.Of the areas affe'cted within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

ese valves must change position during the switchover from injection to recirculation flow following LOCA.COOK NUCLEAR PLANT-UNIT 2 3/4 5-4 AMENDMENT N0.7$,$3$

ee GENCY COR COOL G S S SURVe.ELLANCE R Cc.5.3el The ECCS subsystem shall be demonstrated OPERABLE per the applicable

~~Sn~eillence Reqnizenence e=4.5.2.Oe.4.5.3.2 All charging pumps and safety in]ection pumps, except tne above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their elecMcal po~er supply circuits, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months vhenever the temperature of one or more of"the RCS cold legs is less than or equal to 152 F as determined at least once per hour~hen any RCS cold leg temperature is between 152'F and 200 F.v COOK HUCLZAR PLANT-UNXT 2 3/c&5-8 CO SYS S URVEILLANCE E REMENTS Co tinued C~At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure--High-High test signal+.d.At least once per 5 years by verifying a water flow rate of at least 20 gpm (greater than or equal to 20 gpm)but not to exceed 50 gpm (less than or equal to 50 gpm)from the spray additive tank test line to each containment spray system with the spray pump operating on recirculation with a pump discharge pressure greater than or equal to 255 psig.chnM~peci-re-appkkcab1e.

COOK NUCLEAR PIANT-UNET 2~3/4 6-12 AMENDMENT NO.4Sg Mf 48k)15

CONTAINM NT SYSTEMS SURVE LANCE R UIREMENTS Continued 4.6.3.1.2 Each containment isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOHH or REFUELING NODE at least once per IG months by:I a.Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.b.Verifying that on a Phase B containment isolation test signal,.each Phase B isolation valve actuates to its isolation position.c.Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation, position.4.6.3.1.3 The isolation time of each power operated or automatic containment isolation valve'shall be determined to be within its limit when tested pur suant to Specification 4.0.5~e-pmv4~-ien~Meeh

'OOK NUCLEAR PLANT-UNIT 2 3/4 6-14 AMENOMENT NO.97, 43k 458, 165 CO A DW ARRI S G COND+0 ON 3.6.5.9 The divider barrier seal shall be OPERABLE.With the divider barrier seal inoperable, restore the seal to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200'F.VE C RE 4.6.5.9 The divider barrier seal shall be determined OPERABLE at least once per 18 months during shutdown by: a.Removing two divider barrier seal test coupons and verifying that the physical properties of the test coupons are within the acceptable range of values shown in Table 3.6-2.b.Visually inspecting at least 95 percent of the seal's'.entire length and: 2.Verifying that the seal and seal mounting bolts are pro-perly installed, and Verifying that the seal material shows no visual evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearances.

COOK NUCLEAR PLANT-UNIT 2 3/4 6-47 AVOND~NO.M, P

LANT SYSTEMS 3 4 7 3 COMPONENT COOL NG WA ER S S TING COND TIO 0 OPERATION 3.7.3.1 a.At least two independent component cooling water loops shall be OPERABLE.b.At least one component cooling water flow path in support of Unit 1 shutdown functions shall be available.

Specification 3.7.3.1.b.

-At all times when Unit 1 is in MODES 1, 2, 3, or 4.~CT ON: When Specification 3.7.3.1.a.is applicable'.

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .When Specification 3.7.P.l.b is applicable:

3o With no flowpath&Unit 1 available, return at least one flowpath to available status ithin 7 days, or provide equivalent shutdown capabil'ity in Unit 1 and return at least e flow path to available-status within-the next 60 days, or have-Unit 1 in HOT ANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE: a.At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.b.At least once per 18 months during shutdown, by verifying that each automatic valve servicing sagety related eguipment actuates to its correct position on a Safety Infection test signal.g C'.by Verifying Pumt feepm'<~Pung'~%SPCt4C.Wk~

'I,Od5.At least once per 18 months during shutdown, verify that the unit cross-tie valves can cycle full travel.Following cycling, the valves will be verified to be in their closed positions.

OK NUCLEAR PLANT-UNIT 2 3/4 7-12 AMENDMENT NO.M, 4k@, 158

~~0 I 1

~C 3 4.7.4 ESSENTIAL SERVICE WATER SvSTEYt LIMITING CONDITION FOR OPERATION 3.7.4.1 a.A least two independent essential service water loops shall be OP~MI>.b.At least one essential service water flowpath associated with support of Unit 1 shutdown functions shall be available.

APPLICABILITY:

Specification 3.7.4.1.a.

-MODES 1, 2, 3, and 4.Specification 3.7.4.1.b.

-At all times when Unit 1 is in MODES 1, 2, 3, or 4.ACTION'hen Specification 3.7.4.1.a is applicable:

With only one essential service~ater loop OPERABLE;restore at.least two loops to.OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 's or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .When Specification 3.7.4.l.b is applicable:

With no essential service water flow path available in support of Unit 1 shutdown functions, return at least one flow path to available status within 7 days or provide equivalent shutdown capability in Unit 1 and return the equipment to service within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIRENENTS 4.7.4.1 At least two essential service water loops shall be demonstrated OPERABLE: a e At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.b.At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test sitnalQff~~~-~~~~c a COOK NUCL:"4Z.PI MT-UNIT 2 3/4 7-13

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued e.At least once per 18 months by: 1.Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.2.a.Verifying that on a Safety Injection Signal from Unit 1, the system automatically operates in the pressurization/cleanup mode.&b.Verifying that on a Safety Injection Signal from Unit 2, the system automatically operates in the pressurization/cleanup mode.3.Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/16 inch W.G.relative to the outside atmosphere at a system flow rate of 6000 cfm'plus or minus 10%.After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99%of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99%of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.plieable.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 7-16a AMENDMENT&, 434, 488, 202 3/4.7 7 SNUBBERS IHITINC CONDITION FOR OP ATION 3.7.7.1 All safety-related snubbers shall be OP~~"LZ.systems required OPERABLE in those MODES).~CT la Pith one or more snubbex's inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months x'eplace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.7.1.c on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system.SURVE L NC RE UIR NTS~C 4.7.7.1 Each snubber shall be demonstrated OP~~LE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.a Visual ns ectioSnubbers are categorized as inaccessible or accessible during reactor operation.

Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 3.7-9.The visual inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 3.7-9 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Amendment No.~.b.Visual Ins ection Acce tance Criteria Visual inspections shall verify (1)that there are no visible indications of damage or impaired OPERABILITY, (2)attachments to the foundation or supporting structure are secure, and (3)in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not frozen up.'Snubbers which appear inoperable as'result of visual inspections shall be classified as unacceptable and may be reclassified as acceptable for the purpose of establishing the next visual inspection interval," providing that (1)the cause of the rejection is clearly established and remedied for that COOK NUCLEAR PLANT-UNIT 2 3/4 7-20 AMENDMENT NO.492,~Ž-,~159 ELECTRICAL POUTER SYSTEHS SHUTDOWN LIHITINC COND TIQN FOR OPERA ION 3.8.1.2 As a minimum, the following A.C.electrical power sources snail'oe OPERABI-'ne circuit between the offsite transmission network and the onsite Class 1E distribution system, and b.One diesel generator with: 1.A day fuel tank containing a mintuaun of 70 gallons of fuel, 2.A fuel'storage system containing a minimum indicated volume of 46,000 gallons of fuel, and 3.A fuel transfer pump.ACTION Pith less than the above minimum required A.C.electrical power"sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive-reactivity changes+until the minimum required A.C.electrical power sources are restored to OPENABLE status.SURVEIL'LANCE RE UI EHENTS 4.8.1.2 The above required A.C.electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2.a.5.

f*For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RUST is greater than the minimum required by Specification 3.1.2.7.b.2.

The COOK NUCL:"M PLANT-UNIT 2 3/4 8-9 AHENDHENT NO.4M, Ma.159

~p 3/4.0 LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8'LECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS Continued t b.At least once per 92 days by verifying that: 1.The voltage of each connected cell is greater than or equal to 2.13 volts under float charge.The specific gravity, corrected to/7'F, and full electrolyte level (fluid't the bottom of the maximum level indication mark), of each connected cell is greater than or equal to 1.200 and has not decreased more than 0.03 from the value observed during the previous test, and 3.The electrolyte level of each connected cell is between the top of the minimum level indication mark and the bottom of the maximum level indication mark.At least once per 1&months by verifying that: 1.The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, The cell-to~ll and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material, f 3.The battery charger will supply at least 140 amperes at greater than or equal to 250 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .At least once per 18 months, perform a battery service test during shutdown (MODES 5 or 6), by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status the actual or simulated emergency loads for the design duty cycle which is based on the composite load profile.The composite load profile envelopes both the LOCA/LOOP and Station Blackout profiles and provides the basis for the times listed in Table 4.8-2.The battery charger will be disconnected throughout the test.The battery terminal voltage shall be maintained greater than or equal to 210 volts throughout this test~At least once per 60 months, conduct a performance test of battery capacity during shutdown (MODES 5 or 6), by verifying that the battery capacity is at least 80%of the manufacturer's rating.When this test is performed in place of a battery service test, a modified performance test shall be conducted.

Annual performance tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85%of the service life expected for the application.

Degradation is indicated when the battery capacity drops morc than 10%from its capacity on the previous performance test, or is below 90%of the manufacturer's rating.If the bauery has reached 85%of service life, delivers a capacity of 100%or greater of the manufacturer's rated capacity, and has shown no signs of degradation, performance testing at two year intervals is acceptable until the battery shows signs of degradation.

Mte-provisions-ef-Specif cati a licable COOK NUCLEAR PLANT-UNIT 2 Page 3/4 8-13 AMENDME<NT 443, 439, 4@i, 183 3/4.0 LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8 ELECTRICAL POWER SYSTEMS D.C.DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C.electrical equipment and bus shall be energized and OPERABLE: 1-250-volt D.C.bus, and 1-250-volt battery bank and charger associated with the above D.C.bus.APPLICABILITY:

MODES 5 and 6.ACTION: With less than the above complement of D.C.equipment and bus OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .SURVEILLANCE RE UIREMENTS 4.8.2.4.1 The above required 250-volt D.C.bus shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 250-volt battery bank and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.~

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 8-15 e-epplieabl

+242 X2.cLfowho4-C~wtte~~+dtargcr.

ttirtnncnt AMENDMENT 443, 4', 183 3.9.12 The spent fuel storage pool exhaust ventilation system shall be OPERABLE.whenever irradiated fuel is in the storage pool.Pith no fuel storage pool exhaust ventilation system OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one spent fuel storage pool exhaust ventilation system is restored to OPERABLE status.*b.The provisions of Specifications 3.0.3 and 3.0'are not applicable.

S'RV VC R U ITS 4.9.12 The above required fuel storage pool ventilation system shall be demonstrated OPERABLE: a.At least once per 31 days by initiating flow through the HEPA filter and charcoal adsorber train and verifying-hat the train operates for at least 15 minutes.b.At least once per 18 months or (1)after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2)following painting, fire or chemical release in any ventilation zone communicating with the system, by: 1.Deleted., C.COOK-UNIT 2 3/4 9-12 Amendment No 111 2.Verifying that the charcoal adsorbers remove)99%of a halogenated hydrocarbon

-,refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation system at a flaw rate of 30,000 cfm+10%.gillie QALac 4%clqqr og~clvxl l I+The crane bay roll-up door and the may be opened under administrative control.during movement of fuel within the storage pool or crane.operation with loads over the storage pool.Shared system with D.C.COOK-UNIT 1.This does not include the main load block.For purposes of this specification, a deenergized main load block need not be considered a load.

5.0 DESIGN FEATURES 5.6 FUEL STORAGE Continued CRITICALITY

-SPENT FUEL Continued The equivalent reactivity criteria for Region 2 and Region 3 is defined via the following equations'an~mphieai~ieted=in=Rgtt~~

For Re ion 2 Stora e Minimum Assembly Average Burnup in MWD/MTU=-22,670+22,220 E-2,260 E+149 E For Re ion 3 Stora e Minimum Assembly Average Burnup in MWD/MTU=-26,745+18,746 E-1,631 E2+98.4 E3 Where E=Initial Peak Enrichment 5.6.1.2 Fuel stored in the spent fuel storage raCks shall have a nominal fuel assembly enrichment as follows: Description Maximum Nominal Fuel Assembly Enrichment Wt.%U-235 1)Westinghouse 2)Exxon/ANF 3)Westinghouse 4)Exxon/ANF 15 x 15 STD 15 x 15 OFA 15 x 15 17 x 17 STD 17 x 17 OFA 17 x 17 V5 17 x 17 4.95 4.95 4.95 4.95 COOK NUCLEAR PLANT-UNIT 2 Page 54 AMENDMENT 44, 424, 447, 4', 198

'3 r'9 4~C~Z CCC r u p o r p xi~cc BUrc%lP L"=MAN 2 Los I (tl~E iA~~f.4.14 LO LS LO A+A 44.~IN~i.BlRlCHMKHT,%U-235 5-8

6.0 ADMINISTRATIVE

CONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facility staff sh a ll meet or exceed!he minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (I)the Plant Radiation Protection Manager, who shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (2)the Shift Technical Advisor, who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents and, (3)the Operations Superintendent, who must hetthuntmuohct

'.crater Lt specified in Section 6.2.2P.6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION 6.5.1.1 The PNSRC shall function to advise the Site Vice President/Plant Manager, or designe", on all matters related to nuclear safety.COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Department Superintendents, or supervisory personnel reporting directly to the Site Vice President/Plant Manager, Assistant Plant Managers or Department Superintendents.

The membership shall represent the functional areas of the plant, including, but not limited to Operations, Tcchnical Support, Licensing, Maintenance and Radiation Protection.

The PNSRC membership shall consist of at least one individual from each of the areas designated.

All members, including the Chairman and his alternates, the members and their alternates, shall be designated by the Site Vice President/Plant Manager.PNSRC members and alternates shall meet or exceed the minimum qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions.

The nuclear power plant operations individual shall meet the qualifications of Section 4.2.2 of ANSI N18.1-1971 except for the requirement to hold a current Senior Operator License.The operations individual must hold or have held a Senior Operator License or have been certified for equivalent senior opeiator knowledge at Cook Nuclear Plant or a similar reactor.The maintenance individual shall meet the qualifications of Section 4.2.3 of ANSI N18.1-1971.

COOK NUCLEAR PLANT-UNIT 2 Page 6-4 AMENDMENT 34, 447)438, 4VR, 47$, 197

ATTACHMENT 3A TO AEP:NRC:0433Q PROPOSED TECHNICAL SPECIFICATIONS PAGES REVISED PAGES UNIT 1 3/4 0"3 3/4 3-2la 3/4 4-38 3/4 4-40 3/4 7-15 3/4 9-1 3/4 9-13 5-6 5-7b 6-4 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY SURVEILLANCE REQUIREMENTS 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:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice ins ction and testin criteria Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Yearly or annually Required frequencies for performing inservice inspection and testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 366 days The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

Performance of the above'inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

4.0.6 4.0.7 Deleted.Deleted.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 0-3 AMENDMENT 400, 43k, 444 1 U~f<t 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMIi22'I'ATION ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 6.MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS C a.Steam Generator Water 3/Stm.Gen.Level-Low-Low 14 1,2,3 2/Stm.Gen.2/Stm.Gen.any Stm.Gen.MINIMUM TOTAL NO.OF CHANNELS CHANNELS APPLICABLE CHANNELS TO TRIP OPERABLE MODES ACTION b.4 kv Bus Loss of Voltage 3/Bus Pump Start Valve Actuation (Both trains)c.Safety Injection 2/bus (T11A-Train B;T11D-Train A)2/bus on (T11A&T11B or 2/busses T11C&T11D)1,2,3 1, 2, 3 14 18'.Loss of Main Feedwater Pumps 7.TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS'2 1,2 18 a.Steam Generator Water 3/Stm.Gen.Level-Low-Low b.Reactor Coolant Pump 4-1/Bus Bus Undervoltage 8.LOSS OF POWER 2/Stm.Gen.2/Stm.Gen.any 2 Stm.Gen.I, 2, 3 1,2,3 14 19 a.4 kv Bus Loss of Voltage 1,2,3,4 14 b.4 kv Bus Degraded Voltage 3/Bus (TI IA-Train B;TI ID-Train A)2/Bus 2/Bus 1, 2, 3, 4 (T11A-Train B;(T11A-Train B;T11D-Train A)T11D-Train A)14COOK NUCLEAR PLANT-UNIT 1 Page 3/4 3-21n AMENDMENT$9, 48$, 4' I

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT VENT SYSTEM REACTOR VESSEL HEAD VENTS SURVEILLANCE RE UIREMENTS 4.4.12.1 Both Reactor Vessel head vent paths shall be demonstrated OPERABLE at least once per t8 months by: Verifying the common manual isolation valve in the Reactor vessel head vent is sealed in the open position.Cycling each of the remotely operated valves in each path through at least one complete cycle of full travel from the Control Room while in Modes 5 or 6.Verifying flow through both of the Reactor Vessel head vent paths during venting operation, while in Modes 5 or 6.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 4-38 AMENDMENT 98 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4,4 REACTOR COOLANT SYSTEM REACTOR COOLANT VENT SYSTEM PRESSURIZER STEAM SPACE VENTS SURVEILLANCE RE UIREMENTS 4.4.12.2 Both Pressnrlacr steam space vent paths shall be demonstrated OPERABLE at least once per t8 months by: 1.Verifying the common manual isolation valve in the Pressurizer steam space vent is sealed in the open position.Cycling each of the remotely operated valves in each path through at least one complete cycle of full travel from the Control Room while in Modes 5 or 6.3.Verifying flow through both of the Pressurizer steam space vent paths during venting operation, while in Modes 5 or 6.COOK NUCLEAR PLANT-UNIT I Page 3/4 440 AMENDMENT 98 3/4 LIMITING CONDITIONS I OR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.7 PLANT SYSTEMS PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 APPLICABILITY:

ACTION: At least two independent component cooling water loops shall be OPERABLE.At least one component cooling water flowpath in support of Unit 2 shutdown functions shall be available.

Specification 3.7.3.1.a-MODES 1, 2, 3 and 4.Specification3.7.3.1.b-At all times when Unit 2 is in MODES 1, 2, 3, or 4.When Speciiflication3.7.3.1.a is applicable:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .When Specification3.7.3.1.b is applicable:

With no flowpath to Unit 2 available, return at least one flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 2 and return at least one flow path to available status within the next 60 days, or have Unit 2 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirementsof Specification3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.By verifying pump performancepursuant to Speciftcation4.0.5.

At least once per id months during shutdown, by verifying that the unit cross-tie valves can cycle full travel.Following cycling, the valves will be verified to be in their closed positions.

COOK'NUCLEAR PLANT-UNIT 1 Page 3/4 7-15 dhggbtDhtgbyg gttg, ggh, ddd, ddd r>0 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met: Either a K,of 0.95 or less, which includes a 1%Ak/k conservative allowance for uncertainties, or A boron concentration of greater than or equal to 2400 ppm, whLh includes a 50 ppm conservative allowance for uncertainties.

APPLICABILITY:

ACTION: MODE 6 With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes" and initiate and continue boration at greater than or equal to 10 gpm of 20,000 ppm boric acid solution or its equivalent until K,z is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 2400 ppm, whichever is the more restrictive.

The provisions of Specification3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to: a.Removing or unbolting the reactor vessel head, and b.Withdrawal of any full length control rod in excess of 3 feet from its fully inserted position.4.9.1.2 The boron concentration of the reactor coolant system and the refueling, canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .'The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.10 For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

COOK NUCLEAR PLANT-UNIT 1 Page 3/4 9-1 AMENDMENTS%

II~~I C 4 k A t" t'I ,K 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMIPlTS 3/4.9 REFUELING OPERATIONS REFUELING OPERATIONS STORAGE POOL VENTILATION SYSTEMh*LIMITING CONDITION FOR OPERATION 3.9.12 The spent fuel storage pool exhaust ventilation system shall be OPERABLE.APPLICABILITY:

Whenever irradiated fuel is in the storage pool.ACTION: a.With no fuel storage pool exhaust ventilation system OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool+until at least one spent fuel storage pool exhaust ventilation system is restored to OPERABLE status.'h b.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.12 The above required fuel storage pool ventilation system shall be demonstrated OPERABLE: At least once per 31 days by initiating fiow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.At least once per 18 months or (1)after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2)following painting, fire or chemical release in any ventilation zone communicating with the system, by: 1.Deleted.2.Verifying that the charcoal adsorbers remove 2 99%of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation system at a flow rate of 30,000 cfm k 10%.srthe crane bay roti-up door and the south door oy the auxiliary building crane bay may be opened under administrative control during movement of fuel within the storage pool or crane operation with loads over the storage pool.'pd'Shared system with D.C.COOK-UNIT 2.+This does not include the main load block.For purposes of this specification, a dewnergized main load block need not be considered a load.COOK NUCLEAR PLANT-UNIT 1 Page 3/4 9-13 u't ll I'I'r a s~>>'+'.,>>.r 5.0 DESIGN FEATURES 5.6 FUEL STORAGE Continued 1.Region 1 is designed to accommodate new fuel with a maximum nominal enrichment of 4.95 wt%U-235, or spent fuel regardless of the discharge fuel burnup.Region 2 is designed to accommodate fuel of 4.95%initial nominal enrichment burned to at least 50,000 MWD/MtU, or fuel of other enrichments with equivalent reactivity.

Region 3 is designed to accommodate fuel of 4.95%initial nominal enrichment burned to at least 38,000 MWD/MtU, or fuel of other enrichments with equivalent reactivity.

The equivalent reacdvtqr criteria for Region 2 and Region 3 is defined via the following equations:

For Re ion2Stora e Minimum Assembly Average Burnup in MWD/MTU=22670+22220 E-2260'+149 E For Re ion 3 Stora e Minimum Assembly Average Burnup in MWD/MTU=-26,745+18,746 E-1,631 E~+98.4 E3 Where E=Initial Peak Enrichment 5.6.1.2: I Fuel stored in the spent fuel storage racks shall have a maximum nominal fuel assembly enrichment as follows: Maximum Nominal Fuel Description Assembly Enrichment Wt.%U-235 1)Westinghouse 2)Exxon/ANF 3)Westinghouse 4)Exxon/ANF 15 x 15 STD 15 x 15 OFA 15x15 17 x 17 STD 17x 17OFA 17 x 17 VS 17 x 17 4.95 4.95 4.95 4.95 COOK NUCLEAR PLANT-UNIT 1 Page 54 AMENDMENT 5V, 486, 448, 469, SH t~'~~~I'~<<')i~4v l (Lh 4'4 III l, q yg Figure 5.6-3 intentionally deleted.COOK NUCLEAR PLANT-UNlT 1 Page 5-7b AMENDMENT Bi9.

6.0 ADMINISTRATIVE

CONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1)the Plant Radiation Protection Manager, who shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (2)the Shift Technical Advisor, who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents and, (3)the Operations Superintendent, who must be qualified as specified in Section 6.2.2.g.6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITI'EE NSRC FUNCTION 6.5.1.1 The PNSRC shall function to advise the Site Vice President/Plant Manager, or designee, on all matters related to nuclear safety.COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Department Superintendents, or supervisory personnel reporting directly to the Site Vice President/Plant Manager, Assistant Plant Managers or Department Superintendents.

The membership shall represent the functional areas of the plant, including, but not limited to Operations, Tcchnical Support, Licensing, Maintenance and Radiation Protection.

The PNSRC membership shall consist of at least one individual from each of the areas designated.

All members, including the Chairman and his alternates, the members and their alternates, shall be designated by the Site Vice President/Plant Manager.PNSRC members and alternates shall meet or exceed the minimum qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions.

The nuclear power plant operations individual shall meet the qualifications of Section 4.2.2 of ANSI N18.1-1971 except for the requirement to hold a current Senior Operator License.The operations individual must hold or have held a Senior Operator License or have been certified for equivalent senior operator knowledge.

at Cook Nuclear Plant or a similar reactor.The maintenance individual shall meet the qualifications of Section 4.2.3 of ANSI N18.1-1971.

COOK NUCLEAR PLANT-UNIT 1 Page 64 AMENDMENT 49, 6S, 4', 484)486, 493, 212

ATTACHMENT 3B TO AEP:NRC:0433Q PROPOSED TECHNICAL SPECIFICATIONS PAGES REVISED PAGES UNIT 2 , 3/4 0-3 3/4 0-4 3/4 3-11 3/4 3-20 3/4 3-30 3/4 3-31 3/4 3-34 3/4 3-44d 3/4 3-47 3/4 4-13 3/4 4-14 3/4 4-33 3/4 4-35 3/4,4-37 3/4 5-4 3/4 5"8 3/4 6-12 3/4 6-14 3/4 6-47 3/4 7-12 3/4 7-13 3/4 7-16a 3/4 7-20 3/4 8-9 3/4 8-13 3/4 8-15 3/4 9-12 5-6 5-8 6-4 3/4 LIMITING CONDITIONS FOR OPERATION AND SIRVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY SURVEILLANCE RE UIREMENTS 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:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice inspection and testing criteria Weekly Monthly Quarterly or every 3 months Semiannually or every 6 monoths Yearly or annually Required frequencies for performing inservice inspection and testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 366 days c.The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

4.0.6 Deleted.4.0.7 Deleted.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 0-3 AMENDMENT VS, QV, 48k 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.0 APPLICABILITY SURVEILLANCE RE UIREMENTS 4.0.8 Deleted.4.0.9 Deleted.COOK NUCLEAR PLANT-UNIT2 1 Page 3/4 S4~wuMHmm, m TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 1.Manual Reactor Trip A.Shunt Trip Function B.Undervoltage Trip Function 2.Power Range, Neutron Flux 3.Power Range, Neutron Flux, High Positive Rate 4.Power Range, Neutron Flux, High Negative Rate 5.Intermediate Range, Neutron Flux 6.Source Range, Neutron Flux 7.Overtemperature b,T 8.Overpower hT 9.Pressurizer Pressure-Low 10.Pressurizer Pressure-High 11.Pressurizer Water Level-High 12.Loss of Flow-Single Loop CHANNEL CHECK N.A.N.A.S N.A.CHANNEL CALIBRATION N.A.N.A.D(2,8), M(3,8)and Q(6,8)R(6)R(6)R(6,8)R(6,14)R(9)R(9)R(8)CHANNEL FUNCTIONAL TEST S/U(1)(10)

S/U(1)(10)

M and S/U(1)S/U(1)M(14)Gild S/U(1)M M MODE IN WHICH SURVEILLANCE RE UIRED I, 2, 3, 4, 5 1,2, 3,4',5 1,2and 1,2 1,2 1,2, and 2(7), 3(7), 4 and 5 1,2 1,2 1,2 1,2 1,2 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 7.TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS MINIMUM TOTAL NO.CHANNELS CHANNELS OF CHANNELS TO TRIP OPERABLE APPLICABLE MODES ACTION a.Steam Generator Water 3/Stm.Gen.Level-Low-Low 2/Stm.Gen.2/Stm.Gen.any 2 Stm.Gen.1,2, 3 14'.Reactor Coolant Pump 4-1/Bus Bus Undervoltage 8.LOSS OF'POWER 1,2, 3 19'.4 kV Bus Loss of Voltage b.4 kV Bus Degraded Voltage 9.MANUAL a.Safety Injection (ECCS)Feedwater Isolation Reactor Trip (SI)Containment Isolation-Phase"A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System b.Containment Spray Containment Isolation-Phase"B" Containment Purge and Exhaust Isolation Containment Air Recirculation Fan c.Containment Isolation-Phase"A" Containment Purge and Exhaust Isolation d.Steam Line Isolation 3/Bus 3/Bus P21A-Train 8;T21D-Tnin A)2/train 1/train 1/train 2/steam line (1 per train)2/Bus (T21A-Tnin 8;TZID-Train A)1/train 1/train 1/train 2/steam line (1 per train)2/Bus 2/Bus (TRIA-Tnin B;T21D-Tnin A)2/train 1/train 1/train 2/operating steam line (1 per train)1,2,3,4 14',2,3,4 14 1,2,3,4 18 1,2,3,4 18 1,2,3,4 18 1,2,3 20 COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-20 AMENDMENT VV;4%, 437 3/4 LIMlTING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCFIONAL UNIT l.SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS TRIP ACTUATING DEVICE OPERATIONAL MODES IN CHANNEL WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST~RE(EUIRED a.Manual Initiation b.Automatic Actuation Logic c.Containment Pressure-High d.Pressurizer Pressure-Low e.Differential Pressure Between Steam Lines-High f.Steam Linc Pressure-Low N.A.M(3)N.A.N.A.N.A.N.A.Sec Functional Unit 9 M(2)N.A.I, 2, 3, 4 I, 2, 3 I, 2, 3 I, 2, 3 I, 2, 3 2.CONTAINMENT SPRAY a.Manual Initiation b.Automatic Actuation Logic c.Containment Pressurc-High-High 3.CONTAINMENT ISOLATION a.Phase A Isolation I)Manual 2)From Safety Injection Automatic Actuation Logic b.Phase B Isolation I)Manual 2)Automatic Actuation Logic 3)Containment Pressure-High-High N.A.N.A.N.A.N.A.N.A.See Functional Unit 9 M(2)M(3)Sec Functional Unit 9 M(2)See Functional Unit 9 M(2)M(3)N.A.N.A.N.A.N.A.N.A.I, 2, 3, 4 I, 2, 3 I, 2, 3, 4 I, 2, 3, 4 I, 2, 3 COOK NUCLEAR PLANT-UNlT 2 Page 3/4 3-30 AMENDMENT 84, 9, 43V, 488 g~t b 0 II 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3 INSTRUMENTATION TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT c.Purge and Exhaust Isolation CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCfIONAL TEST TRIP ACTUATING DEVICE OPERATIONAL MODES IN WHICH SURVEILLANCE RHtHUIRED I)Manual 2)Containment Radioactivity

-High See Functional Unit 9 e N.A.I, 2, 3, 4 4.STEAM LINE ISOLATION a.Manual b.Automatic Actuation Logic c.Containment Pressure-High-High d.Steam How in Two Steam Lines-High Coincident with T~-Low-Low c.Steam Line Pressure-Low N.A.N.A.See Functional Unit 9 M(2)M(3)M N.A.N.A.N.A.N.A.I, 2, 3 1,2,3 1,2,3 1,2,3 5.TURBINE TRIP AND FEEDWATER ISOLATION a.Steam Generator Water Lcvcl-High-High M N.A.1,2,3 6.MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.Steam Generator Water Level-Low-Low b.4 kV Bus Loss of Voltage S c.Safety Injection N.A.d.Loss of Main Feed Pumps N.A.R N.A.N.A.M M(2)R N.A.N.A.N.A.N.A.1,2,3 1,2,3 I, 2, 3 1,2 COOK NUCLEAR PLANT-UNIT2 Page 3/4 3-31 AMENDMENT%,%, 43k, 434, 4&V,

~)~/,~

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4,3 INFIRUMENTATION INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.APPLICABILITY:

ACTION: As shown in Table 3.3-6.b.With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.c.The provisions of Specifications 3.0.3 and 3.0.4 are not applicable SURVEILLANCE RE UIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-3.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 3-34 TABLE 4.3-6A APPENDIX R REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS 0 0 R C O R I R tA 4J C4 INSTRUMENT 1.Steam Generators 1 and 4 Level 2.Steam Generators 2 and 3 Level 3.Steam Generators 1 and 4 Pressure 4.Steam Generators 2 and 3 Pressure 5.Reactor Coolant Loop 4 Temperature (Cold)6.Reactor Coolant Loop 4 Temperature (Hot)7.Reactor Coolant Loop 2 Temperature (Cold)Reactor Coolant Loop 2 Temperature (Hot)9.Pressurizer Level 10.Reactor Coolant System Pressure 11.Charging Cross-Flow Between Units 12.Source Range Neutron Detector (N-23)LOCATION LSI Cabinet 1 and LSI Cabinet 4 LSI Cabinet 2 and LSI Cabinet 4 LSI Cabinet 4 and LSI Cabinet 5 LSI Cabinet 4 and LSI Cabinet 6 LSI Cabinet 4 and LSI Cabinet 5 LSI Cabinet 4 and LSI Cabinet 5 LSI Cabinet 4 and LSI Cabinet 6 LSI Cabinet 4 and LSI Cabinet 6 , LSI Cabinet 3 LSI Cabinet 3 Corridor Elev.587'SI Cabinet 4 CHANNEL CHECK M CHANNEL CALIBRATION ,R*Charging Cross-Flow between Units is an instrument common to both Unit I and 2.This surveillance will only be conducted on an interval consistent with Unit I refueling.

TABLE 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION 1.2.3.4.5.6.7.8.9.10.11.12.13.14.15.16.17.18.Containment Pressure Reactor Coolant Outlet Temperature

-Tnor (Wide Range)Reactor Coolant Inlet Temperature

-Tco~(Wide Range)Reactor Coolant Pressure-Wide Range Pressurizer Water Level Steam Line Pressure Steam Generator Water Level-Narrow Range RWST Water Level Boric Acid Tank Solution Level Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator-Limit Switches PORV Block Valve Position Indicator-Limit Switches Safety Valve Position Indicator-Acoustic Monitor Incore Thermocouples (Core Exit Thermocouples)

Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

Containment Sump Level Containment Water Level M M M M M M M M M M M M M M M Mu~M M R R R R R R R R R R R R R R Rtu RP)R R tnPartial range channel calibration for sensor to be performed below P-12 in MODE 3."Kith one train of Reactor Vessel Level Indication inoperable, Subcooling Margin Indication and Core Exit Thermocouples may be used to perform a CHANNEL CHECK to verify the remaining Reactor Vessel Indication train OPERABLE.+Completion of channel calibration for sensors to be performed below P-12 in MODE 3.

-0 0 1ST SAMPLE INSPECTION TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size A minimum of S Tubes per S.G.Result C-1 C-2 Action Required None Plug defective tubes and inspect additional 2S tubes in this S.G.Result N/A C-1 C-2 Action Required N/A None Plug defective tubes and inspect additional 4S tubes in this S.G.Result N/A N/A C-1 Action Required N/A N/A None C-3 Inspect all tubes in this S.G., plug defective tubes and inspect 2S tubes in each other S.G.C-3 All other S.G.s are C 1 Perform action for C-3 result of first sample None C-2 C-3 N/A N/A Plug defective tubes Perform action for C-3 result of first sample N/A N/A Prompt notification to NRC pursuant to specification 6.9.1 Some S.G.s Perform action for C-2 C-2 but no result of second sample additional S.G.are C-3.N/A N/A Additional S.G.is C-3 Inspect all tubes in each S.G.and plug defective tubes.Prompt notification to NRC pursuant to specification 6.9.1.'/A N/A S=3(N+n)%Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection.

I I h 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE: a1 APPLICABILITY:

One of the containment atmosphere particulate radioactivity monitoring channels (ERS-2301 or ERS-2401), The containment sump level and flow monitoring system, and Either the containment humidity monitor or one of the containment atmosphere gaseous radioactivity monitoring channels (ERS-2305 or ERS-2405).

MODES 1, 2, 3 and 4 ACTION: With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity monitoring channels are inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by: Containment atmosphere particulate and gaseous (if being used)monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3.Containment sump level and flow monitoring system-perfo'rmance of CHANNEL CALIBRATION at least once per 18 months.C.Containment humidity monitor (if being used)-performance of CHANNEL CALIBRATION at least once per 18 months.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 4-14 AMENDMENT V8, 43k, 459 II 1>i~C5 P I 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued With PORVs and block valves not in the same line inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the valves to OPERABLE status or close and de-energize the associated block valve and place the associated PORV in manual control in each respective line.Apply the portions of ACTION c or d above, relating to the OPERATIONAL MODE, as appropriate for two or three lines unavailable.

h.The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.1 F 1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE: a.At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and b.At least once per 18 months by operating the PORV through one complete cycle of full travel during MODES 3 or 4, and c.At least once per 18 months by operating solenoid air control valves and check valves in PORV control systems through one complete cycle of full travel, and d.At least once per 18 months by performing a CHANNEL CALIBRATION of the actuation instrumentation.

4.4.1 1.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION b, c, or d in Specification 3.4.11.4.4.1 1.3 Deleted.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 4-33 AMENDMENT 45k)458, 459, 46k, 496

~~w 8'I ll I 0 0 ,t~+il~p"/'q I'r', ii'I'9 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT VENT SYSTEM"'EACTOR VESSEL HEAD VENTS SURVEILLANCE RE UIREMENTS 4.4.12.1 Both Reactor Vessel head vent paths shall be demonstrated OPERABLE at least once per 18 months by: Verifying the common manual isolation valve in the Reactor vessel head vent is sealed in the open position.Cycling each of the remotely operated valves in each path through at least one complete cycle of full travel from the Control Room while in Modes 5 or 6.Verifying flow through both of the Reactor Vessel head vent paths during venting operation, while in Modes 5 or 6.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 4-35 AMENDMENT 65

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT VENT SYSTEM PRESSURIZER STEAM SPACE VENTS SURVEILLANCE RE UIREMENTS 4.4.12.2 Both Pressurizer steam space vent paths shall be demonstrated OPERABLE at least once per 18 months by: 1.Verifying the common manual isolation valve in the Pressurizer steam space vent is sealed in the open position.2.Cycling each of the remotely operated valves in each path through at least one complete cycle of full travel from the Control Room while in Modes 5 or 6.3.Verifying flow through both of the Pressurizer steam space vent paths during venting operation, while in Modes 5 or 6.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 4-37 AMENDMHNT dd

~')F+F~<<g/<<I, F<<b~<<"<<," P<<FY 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4,5 EMERGENCY CORE COOLING SYSTEMS (ECCS)SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: ao At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the following valves are in the indicated positions with power to the valve operators removed: Valve Number Valve Function Valve Position a.IM0-390 b.IMO-315 c.IM0-325 d.IM0-262'.IM0-263 f.~IM0-261'.ICM-305'l.

ICM-306'.RWST to RHR b.Low head SI to Hot Leg c.Low head SI to Hot Leg d.Mini flow line e.Mini flow line f.SI Suction g.Sump Line h.Sump Line a.Open b.Closed c.Closed d.Open e.Open f.Open g.Closed h.Closed b.At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.c.By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.)is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.

This visual inspection shall be performed:

1.For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2.Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

These valves must change position during the switchover from injection to recirculation fiow following LOCA.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 54~mnmm vs, m

't A U"~a v',>I'~~

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)SURVEILLANCE RE UIREMENTS 4.5.3.1, The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance

!Requirements of 4.5.2.4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the RCS cold legs is less than or equal to 15ZF as determined at least once per hour when any RCS cold leg temperature is between 15ZF and 200'F.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 54

.w a 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure-High-High test signal.At least once per 5 years by verifying a water flow rate of at least 20 gpm (greater thanor equal to 20 gpm)but not to exceed 50 gpm (less than or equal to 50 gpm)from the spray additive tank test line to each containment spray system with the spray pump operating on recirculation with a pump discharge pressure greater than or equal to 255psig.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 6-12 p~a (~s I I*V~~f C 3/4 LIMrIING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.6.3.1.2 Each containment isolation valve specified shall hc demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by: Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation position.4.6.3.1.3 The isolation time of each power operated or automatic containment isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 6-14 AMENDMENT PP, age, Sgg, Sdg e~4 gIf I I'v e g~C j Jf~g I t'I If~~0'll i'Jf II, If ll 1 J 3/4 LIMH'ING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6 CONTAINMENT SYSTEMS DIVIDER BARRIER SEAL LIMITING CONDITION FOR OPERATION 3.6.5.9 The divider barrier seal shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3 and 4.ACTION: With the divider barrier seal inoperable, restore the seal to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200'F.SURVEILLANCE RE UIREMENTS 4.6.5.9 The divider barrier seal shall be determined OPERABLE at least once per 18 months during shutdown by: Removing two divider barrier seal test coupons and verifying that the physical properties of the test coupons are within the acceptable range of values shown in Table 3.6-2.Visually inspecting at least 95 percent of the seal's entire length and: Verifying that the seal and seal mounting bolts are properly installed, and Verifying that the seal material shows no visual evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearances.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 6-47 AMENDMENT V8, 4R, 459 ll ti l>t~Hyi'~I, gp)I.Al I s~r 1" i<<e' 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS i 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1 al At least two independent component cooling water loops shall be OPERABLE.At least one component cooling water flow path in support of Unit 1 shutdown functions shall be available.

APPLICABILITY:

ACTION: Specification 3.7.3.1.a.

-MODES 1, 2, 3, 4.Specification 3.7.3.1.b.

-At all times when Unit 1 is in MODES 1, 2, 3, or 4.When Specification 3.7.3.1.a is applicable:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .When Specification 3.7.3.1.b is applicable:

With no flowpath to Unit 1 available, return at least one flowpath to available status within 7 days, or provide I equivalent shutdown capability in Unit 1 and return at least one flow path to available status within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.b.At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.I c.By verifying pump performance pursuant to Specification 4.0.5.d.At least once per 18 months during shutdown, verify that the unit cross-tie valves can cycle full travel.Following cycling, the valves will be verified to be in their closed positions.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 7-12 AMENDMENT%, 44G, 43k, 458

~~'~, 4'j t(~'g 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS 3/4.7.4 ESSENTIAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 a.At least two independent essential service water loops shall be OPERABLE.b.At least one essential service water flowpath associated with support of Unit 1 shutdown functions shall be available.

APPLICABILITY:

Specification 3.7.4.1.a.

-MODES 1, 2, 3, and 4.Specification 3.7.4.1.b.

-At all times when Unit 1 is in MODES 1, 2, 3, or 4.ACTION: When Specification 3.7.4.1.a is applicable:

With only one essential service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .When Specification 3.7.4.1.b is applicable:

With no essential service water flow path available in support of Unit 1 shutdown functions, return at least one fiow path to available status within 7 days or provide equivalent shutdown capability in Unit 1 and return the equipment to service within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.4.1 At least two essential service water loops shall be demonstrated OPERABLE: At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 7-13 AMENDMENT%, 446, 48k, 458, 459

~t H C/nH 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued e.At least once per 18 months by: 1.Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.2.a.Verifying that on a Safety Injection Signal from Unit 1, the system automatically operates in the pressurization/cleanup mode.b.Verifying that on a Safety Injection Signal from Unit 2, the system automatically operates in the pressurization/cleanup mode.3.Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/16 inch W.G.relative to the outside atmosphere at a system flow rate of 6000 cfm plus or minus 10%.After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99%of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.i After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99%of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 7-16a AMENDMENT W, m, m, am o s~d k'l>>g 4 6~0 I'e ri" 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7 PLANT SYSTEMS 3/4.7.7 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.7.1 All safety-related snubbers shall be OPERABLE.APPLICABILITY:

MODES 1, 2, 3 and 4.(MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES).ACTION: With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.7.1.c on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system.SURVEILLANCE RE UIREMENTS 4.7.7.1 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.Visual Ins ction Snubbers are categorized as inaccessible or accessible during reactor operation.

Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 3.7-9.The visual inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 3.7-9 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Amendment No.156.Visual Ins ction Acce tance Criteria Visual inspections shall verify (1)that there are no visible indications of damage or impaired OPERABILITY, (2)attachments to the foundation or supporting structure are secure, and (3)in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not frozen up.Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified as acceptable for the purpose of establishing the next visual inspection interval, providing that (1)the cause of the rejection is clearly established and remedied for thatCOOK NUCLEAR PLANT-UNIT 2 Page 3/4 7-20 AMENDMENT ue, m, 1, m r r IA h, l,a 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8 ELECTRICAL POWER SYSTEMS ELECTRICAL POWER SYSTEMS SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C.electrical power sources shall be OPERABLE: One circuit between the offsite transmission network and the onsite Class 1E distribution system, and One diesel generator with: 1.A day fuel tank containing a minimum of 70 gallons of fuel.2.A fuel storage system containing a minimum indicated volume of 46,000 gallons of fuel, and APPLICABILITY:

3.A fuel transfer pump.MODES 5 and 6.ACI'ION: Will less than the above minimum required A.C.electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes*until the minimum requiredA.C.

electrical power sources are restored to OPERABLE status.SURVEILLANCE RE UIREMENTS 4.8.1.2 The above required A.C.electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1~1.2 except for!requirement 4.8.1~1.2.a.5.For purpose of this specification, addition of'water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 8-9 AMENDMENT ARAN, 444 459 3/4.0 LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8 ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS Continued At least once per 92 days by verifying that: The voltage of each connected cell is greater than or equal to 2.13 volts under float charge.The specific gravity, corrected to 77'F, and full electrolyte level (fluid at the bottom of the maximum level indication mark), of each connected cell is greater than or equal to 1.200 and has not decreased more than 0.03 from the value observed during the previous test, and The electrolyte level of each connected cell is between the top of the minimum level indication mark and the bottom of the maximum level indication mark.C.At least once per 18 months by verifying that: 1.The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration, The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with antieorrosion material,3.The battery charger will supply at least 140 amperes at greater than or equal to 250 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .At least once per 18 months, perform a battery service test during shutdown (MODES 5 or 6), by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status the actual or simulated emergency loads for the design duty cycle which is based on the composite load profile.The composite load profile envelopes both the LOCA/LOOP and Station Blackout profiles and provides the basis for the times listed in Table 4.8-2.The battery charger will be disconnected throughout the test.The battery terminal voltage shall be maintained greater than or equal to 210 volts throughout this test.e.At least once per 60 months, conduct a performance test of battery capacity during shutdown (MODES 5 or 6), by verifying that the battery capacity is at least 80%of the manufacturer's rating.When this test is performed in place of a battery service test, a modified performance test shall be conducted.

Annual performance tests of battery capacity shall be given to any battery that shows signs of degradation or has reached 85%of the service life expected for the application.

Degradation is indicated when the battery capacity drops more than 10%from its capacity on the previous performance test, or is below 90%of the manufacturer's rating.If the battery has reached 85%of service life, delivers a capacity of 100%or greater of the manufacturer's rated capacity, and has shown no signs of degradation, performance testing at two year intervals is acceptable until the battery shows signs of degradation.

COOK NUCLEAR PLANT-UNIT 2 Page 3/4 8-13 AMENDMENT 448, 489, 466, 46 II~~

3/4.0 LIMI'rING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8 ELECTRICAL POWER SYSTEMS D.C.DISTRIBUTION

-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.4 As a minimum, the following D.C.electrical equipment and bus shall be energized and OPERABLE: 1-250-volt D.C.bus, and 1-250-volt battery bank and charger associated with the above D.C.bus.APPLICABILITY:

MODES 5 and 6.ACTION: With less than the above complement of D.C.equipment and bus OPERABLE, establish CONTAINMENT INTEGRITY within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .SURVEILLANCE RE UIREMENTS 4.8.2.4.1 The above required 250-volt D.C.bus shall be determined OPERABLE and energized at least once per 7 days by verifying correct breaker alignment and indicated power availability.

4.8.2.4.2 The above required 250-volt battery bank and charger shall be demonstrated OPERABLE per Surveillance Requirement 4.8.2.3.2.

ICOOK NUCLEAR PLANT-UNIT 2 Page 3/4 8-15 AMENDMENTS, ae, m Iv,~t 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9 REFUELING OPERATIONS STORAGE POOL VENTILATION SYSTEM" LIMITING CONDITION FOR OPERATION 3.9.12 The spent fuel storage pool exhaust ventilation system shall be OPERABLE.APPLICABILITY:

Whenever irradiated fuel is in the storage pool.ACTION: a.With no fuel storage pool exhaust ventilation system OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storagepool+

until at least one spent fuel storage pool exhaust ventilation system is restored to OPERABLE status'.The provisions of Speciflcations3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.12 The above required fuel storage pool ventilation system shall be demonstrated OPERABLE: a.At least once per 31 days by initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 15 minutes.At least once per IS months or (1)after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2)following painting, fire or chemical release in any ventilation zone communicating with the system, by: 1.Deleted.2.Verifying that the charcoal adsorbers remove 2 99%of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation system at a flow rate of 30,000 cfmk 10%.'The crane bay roii-up door and the south door of the auxiliary building crane bay may be opened under administrative control during movement of fuel within the storage pool or crane operation with loads over the storage pool."Shared system with D.C.Cook-Unit 1.+This docs not include the main load block.For purposes of this specification, adeenergized main load block need not be considered a load.COOK NUCLEAR PLANT-UNIT 2 Page 3/4 9-12 AMENDMENT 444 5.0 DESIGN FEATURES 5.6 FUELSTORAGE Continued CRITICALITY-SPENT FUEL Continued The equivalent reactivity criteria for Region 2 and Region 3 is defined via the following equations:

I For Re ion2Stora e Minimum Assembly Average Burnup in MWD/MTU=-22,670+22,220E-2,2608

+149 E3 For Re ion 3 Stora e Minimum Assembly Average Burnup in MWD/MTU=-26,745+18,746 E-1,631 E'98.4 Ei Where E=Initial Peak Enrichment 5.6.1.2 Maximum Nominal Fuel Assembly Enrichment Wt.%U-235 Fuel stored in the spent fuel storage racks shaH have a nominal fuel assembly enrichment as follows: Description 1)Westinghouse 2)Exxon/ANF 3)Westinghouse 4)Exxon/AN F 15 x 15 STD 15 x 15 OFA 15x 15 17 x 17 STD 17 x 17 OFA 17x17V5 17x 17 4.95 4.95 4.95 4.95COOK NUCLEAR PLANT-UNIT 2 Page 54 AMENDMENT 4k, 49k, 447, 483, 498 5.0 DESIGN FEATURES Figure 5.6-3 intentionally deleted.COOK NUCLEAR PLANT-UNIT 2 Page 5-8

~'I~Q~I 0 II eaI Ilk f i i P'I'I]la l

<I i 6.0 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facili staff sh ty all meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1)the Plant Radiation Protection Manager, who shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (2)the Shift Technical Advisor, who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents and, (3)the Operations Superintendent, who must be qualified as specified in Section 6.2.2.g.6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 10 CFR Part 55.6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION The PNSRC shall function to advise the Site Vice President/Plant Manager, or designee, on all matters related to nuclear safety.COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Department Superintendents, or supervisory personnel reporting directly to the Site Vice President/Plant Manager, Assistant Plant Managers or Department Superintendents.

The membership shall represent the functional areas of the plant, including, but not limited to Operations, Technical Support, Licensing, Maintenance and Radiation Protection.

The PNSRC membership shall consist of at least one individual from each of the areas designated.

All members, including the Chairman and his alternates, the members and their alternates, shall be designated by the Site Vice President/Plant Manager.PNSRC members and alternates shall meet or excccd the minimum qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions.

The nuclear power plant operations individual shall meet the qualifications of Section 4.2.2 of ANSI N18.1-1971 except for the requirement to hold a current Senior Operator License.The operations individual must hold or have held a Senior Operator License or have been certified for equivalent senior operator knowledge at Cook Nuclear Plant or a similar reactor.The maintenance individual shall meet the qualifications of Section 4.2.3 of ANSI N18.1-1971.

COOK NUCLEAR PLANT-UNIT 2 Page 6-4 AMENDMENT 34, 447, 438, 4', 4VS, 493 ATTACHMENT 4 TO AEP:NRC:0433Q 4 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Attachment 4 to AEP:NRC:0433Q Page 1 Evaluation of Si nificant Hazards Consideration The Licensee has evaluated this proposed amendment and determined that it involves no significant hazards consideration.

According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: 1.involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated; 2.create the possibility of a new or different kind of accident from any previously analyzed;or 3.involve a significant reduction in a margin of safety.The Licensee proposes to make administrative changes to several technical specifications (T/S)for Donald C.Cook Nuclear Plant unit 1 and unit 2.The proposed changes include: (1)revising boron sampling requirements in mode 6;(2)deleting a reference to obsolete equipment in a footnote;(3)deleting a redundant figure;(4)correcting a reference to another requirement; (5)deleting obsolete notes;(6)adding to surveillance requirements; (7)clarifying instrumentation configuration; and (8)correcting typographical errors.These changes are proposed to remove obsolete information, provide consistency between unit 1 and unit 2, provide consistency with the Standard Technical Specifications, provide clarification, and correct typographical errors.The determination that the criteria set forth in 10 CFR 50.92 are met for this amendment request is indicated below.1.Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change for boron sampling requirements in mode 6 does not affect the probability of a fuel handling accident.The unlikely event of a fuel assembly being misloaded is independent of the sampling frequency for fuel pool boron concentration.

It has no impact on the event initiator, which is a human error while positioning a fuel assembly.The change has no impact on the*assumptions for a fuel handling accident.The boron concentration requirement is not changed;there is sufficient boron in the fuel storage pool to maintain k,qq below 0.95 to preclude an inadvertent criticality.

Therefore, the consequences of the accident will be mitigated as previously evaluated.

The 72-hour maximum interval between samples is maintained.

Operating experience has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate.Removing the additional limitation of sampling at least three times per week would allow the sample to be collected two or three times per week, consistent with the maximum 72-hour interval.This is acceptable because boron concentration changes occur slowly due to the large 1

Attachment 4 to AEP:NRC:0433Q Page 2 volume of water in the system and relatively small volumes of dilution sources.The consequences are not increased because there are no changes to the spent fuel, shielding (water), or systems used to mitigate the consequences of an accident.Additionally, there is no change in the types or significant increase in the amounts of any effluents released offsite.Deleting the redundant figure for equivalent reactivity criteria for regions in the spent fuel storage racks does not impact the storage requirements because the equations provide equivalent requirements.

The unlikely event of a fuel assembly being misloaded is independent of the characteristics of the spent fuel in the pool.It has no impact on the event initiator, which is a human error while positioning a fuel assembly.The change has no impact the assumptions for a fuel handling accident because the fuel storage requirements are not changed.The consequences of an accident are not increased because the fuel storage requirements are not changed and no other changes are made to systems that mitigate the consequences of an accident.The proposed changes to correct a reference to another requirement, delete obsolete notes, revise the name of drumming room roll-up door, and correct typographical errors are considered administrative.

The reference leads to a section that no longer exists;the proposed change corrects the error.The notes permitted exceptions to requirements, and they are no longer required.The normal requirements have applied since the provisions expired.Deleting them eliminates extraneous information.

The revised description of the door reflects the current use of the installed door.Correcting the typographical errors improves readability.

The corrections are not intended to change the meaning.These changes do not affect accidents described in the UFSAR.Adding new surveillance requirements to test the unit 2 pump performance pursuant to T/S 4.0.5 does not affect accident initiators or precursors.

The change reflects ASME code requirements.

Including the requirements in the corresponding section provides assurance that the pumps will operate as assumed in the accident analyses.As such, the probability and'onsequences of previously evaluated accidents is unchanged.

The proposed change to the description of instrumentation configuration is considered administrative because the configuration had been reviewed and approved by the NRC Staff, as documented in the Safety Evaluation Report for amendment 39 for DPR-58 and amendment 22 for DPR-74.There are no changes to the actual plant configuration.

The change is intended to describe the installed equipment more clearly.The change does not affect the probability and consequences of previously evaluated accidents because the equipment is installed and operated as described in the correspondence related to the previous amendments.

Based on this review, it is concluded that the proposed changes do not involve a significant increase in the L.

Attachment 4 to AEP:NRC:0433Q Page 3 probability of occurrence or consequences of an accident previously evaluated.

2.Does the change create the possibility of a new or different kind of accident from any accident previously evaluated' The proposed changes remove obsolete information, provide consistency between unit 1 and unit 2, provide consistency with the Standard Technical Specifications, provide clarification, and correct typographical errors.These changes are considered administrative because they do not affect the design or operation of any system, structure, or component in the plant~The accident analysis assumptions and results are unchanged.

No new failures or interactions have been created.Based on this review, it is concluded that the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3.Does the change involve a significant reduction in a margin of safety?The proposed changes are considered administrative in nature.They do not affect any safety limits or T/S parameter limits.The proposed changes do not introduce new equipment, equipment modifications, or new or different modes of plant operation.

These changes do not affect the operational characteristics of any equipment or systems.Based on this review, it is concluded that no reduction in the margin of safety will occur as a result of the changes.'n summary, based upon the above evaluation, the Licensee has concluded that these changes involve no significant hazards'onsideration.

ATTACHMENT 5 TO AEP:NRC:0433Q ENVIRONMENTAL ASSESSMENT

Attachment 5 to AEP:NRC:0433Q Page 1 Environmental Assessment The Licensee has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21.The Licensee has determined that this license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9)

.This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria.(i)The amendment involves no significant hazards consideration.

As demonstrated in attachment 4, this proposed amendment does not involve any significant hazards consideration.(ii)There is no significant change.in the types or significant increase in the amounts of any effluent that may be released offsite.As documented in attachment 4, there will be no change in the types or significant increase in the amounts of any effluents released offsite.(iii)There is no significant increase in individual or cumulative occupational radiation exposure.The proposed changes will not result in changes in the operation or configuration of the facility.There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant.Therefore,'here will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

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