Text

Proposed Tech Specs Re Final Policy Statement Improvements
	ML20087G382
Person / Time
Site:	Callaway
Issue date:	03/29/1995
From:	; UNION ELECTRIC CO.
To:	;
Shared Package
ML20087G370	List: ML20087G382; ULNRC-03184, Application for Amend to License NPF-30,responding to Concerns Raised by NRC Staff During Review of Final Policy Statement on TS Improvements;
References
	NUDOCS 9504040191
	Download: ML20087G382 (73)
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. Attachment 1 Page 9 of 11 Sheet 11.1 ', Although.not designed with the'same high. range, further diversity is available from the containment atmosphere radiation monitors (GT-RE-0031 and -0032) which display at the digital radiation monitoring panel SP067.

Section 22 of NUREG 0830 specifically accepted the response to NUREG-0737 Item II.F.1 Attachment 3. Additional discussion is found in FSAR Section 18.2.12. For RVLIS, diversity is provided by the 46 core. exit thermocouples, pressurizer. level indication (BB-LI-0459A,.-0460A,-and - ;

7 0461), and RCS subcooling monitor indication (BB-TI-1390A and B). Additional discussion is found in FSAR Table 7A-3 Data.

Sheet 1.4. If these alternate methods'are used, new Action c does not require a plant shutdown, rather a Special Report is submitted within 14 days per Specification 6.9.2. The report provided to the NRC would discuss the preplanned alternate methods used, outline the cause of the inoperability, and provide a schedule for restoring the normal PAM channels.

New Action d applies when two hydrogen monitor channels are inoperable, requiring the restoration of one hydrogen monitor channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time is reasonable based on the backup-capability of the Post Accident Sampling. System to monitor the hydrogen concentration for evaluation of core damage and to provide information for operator decisions. Also, it is unlikely that a LOCA (which would cause core damage) would occur during this time. Consistent with the new STS, NUREG-1431, 000 2 4 4 A. o del-t si 3.3.3.6 contains the a opriate adions and surveillances. ;

-PORV and FORV bicch valvc pao uieu indicaterr ha"e been dulcted frcr Tcchnical Specificati^" 'A Enne nf pccitivu ludicacivu &cqui.es that the Actione seeeciated with _

LCG 3.4.4 Le enceted; cheteivuc, where ic no nccd to 21re _

ha c thesc audiuuLvuo uudei LCG 3.3.3.0. It ic further noted -

tFct thace indicatero ate avu Type A vuristic; at O'11?"2y -

ner arm Lhcy RG 1.57 Category 2. avuthly channci chccks for-these indiuuturs have oeen acaea as dx a.*.s.3 and Cn 4.1 ; 1.

Determination o No Un viewed Safetv_ Ouestion The proposed changes to the Technical Specifications do not involve an unreviewed safety question because the operation of Callaway Plant in accordance with these proposed changes would not:

(1) Involve an increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety.

previously evaluated in the FSAR. Overall protection system performance will remain within the bounds of the accident analyses documented in FSAR Chapter 15, WCAP-10961-P, and WCAP-11883 since

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i ULNRC-03184 1

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ATTACHMENT THREE REPLACEMENT PAGES FOR ATTA'IMENT 4 TO ULNRC-3023 l

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t INDEX

_ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE '

INSTRUMENTATION (Continued) l ,

Was te Gas Ho l dup Sys tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2/ ? :- 7';)Eu 3/4.3.4 TURB INE OVERS PEED PROTECTION. . . . . . . . . . . . . . . . . . .0--7 . . .C. b. E. .M. 7. E. 2/

b 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION S tartup and Power Operation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Col d Shutdown - Loops Fil l ed. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-5.

Col d Shutdown - Loops Not Fi11 ed. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 3/4.4.2 SAFETY VALVES S hu t d own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2/ 4 0 pe ra ti ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/ 4 4-8 3/4.4.3 PRESSURIZER.............................................. 3/4 4-9 3/4.4.4 RELIEF VALVES.....................I...................... 3/4 4-10 3/4.4.5 STEAM GENERATOP.S. . . . . . .[d.8b'h'. .M M

. ..I.. . . . . . . . . . . . . . . -2/ 4 Table 4.4-1 BEIpSPECT,EQ n !

MINIMUM NUMBER OF STEAM GENERATOR DURING INSERVICE INSPECTION..... 6 .. -e Table 4.4-2 STEAM GENERATOR TUBE INSPECTION. O . . .[.d4b!' .#A S* .. 3/'

3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE Leakage Detection Systems................................

3/4 4-18 ;

Ope ra ti ona l L ea kag e. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-19 i

CALLAWAY - UNIT 1 VII Anendment No. 50 j 1

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. BASES SECTION PAGE-REACTOR COOLANT SYSTEM (Continued)

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3/4.4.5 STEAM GENERATORS.... b.4f.fh ................... J :/4 4, M al.,

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3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.7 CHEMISTRY................................................ " O/4 4 ; ;>fiLdPZE2) l 3/4.4.8 SPECIFIC ACTIVITY........................................ B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................. B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.......................... B 3/4 4-10 FIGURE B 3/4.4-1 CAST NCUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE........................ B 3/4 4-12 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF RT '

NDT FOR REACTOR VESSEL STEELS EXP03ED TO IRRADIATION AT 550*F........................... B 3/4 4-13 3/4.4.10 STRUCTURAL INTEGRITY..................................... " " '" " " bdF4dITMEb :

3/4.4.11 REACTOR COO LANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . -0 :/4 4 17 k"//"72E6

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l 3/4.5 EMERGENCY CORE COOLING SYSTEMS i

j 3/4.5.1 ACCUMULATORS............................................. B 3/4 5-1 i 3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS..................... B 3/4 5-1 l 3/4.5.5 REFUELING WATER STORAGE TANK............................. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS . I 1

3/4.6.1 PRIMARY CONTAINMENT................................ ..... B 3/4 6-1 l 3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS..................... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 6-4 -

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3/4.6.4 COMBUSTIBLE GAS CONTROL........... ...................... B 3/4 6-4 t

l CALLAWAY - UNIT 1 XV --

i DEFINITIONS '

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CONTAINMENT INTEGRITY l

1.7 CGNTAINMENT INTEGRITY shall exist when;

a. All penetrations required to be closed during accident conditions are either:

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) ClosM by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 tf Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with .the requirements of Specification

/;.r.}ed;,, Na Aare.r s$

e., The containme " ' '

3 6 1- ?"A *nl We /towu'bfee Qees hicclfon f 6././.d adore The ~ . hanism associated with each penetration (e.g., welds, Ed.-e-- bellows, or 0-rings) is OPERABLEgM

h. S/ruchenl indefri T.r a r.rar-ed via '/l e f rem ru m de rcriA* l

j ' ') CONTROLLEO LEAXAGE /,, f , ,f V\ -A f .A ,,./g,,A gg,r e, A 1.8 CONTROLLED LEAKAGE shall be tnat seal water flow from the reactor coolant pump seals.

CORE ALTEPATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTEPATION shall not preclude completion of movement of a component to a safe cons'ervative position.

CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit specific document that provides core operating limits for the current operating reload cycle.

The cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Plant operation wif.hin these operating limits is addressed in individual specifications.

DOSE EOUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /

l gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132,1-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-iaEa4, " Calculation of Distance Factors for Power and Test Reactor r tes."

CALLAWAY - UNIT I l-2 Amendment No. JE,35, 58

r REACTIVITY CONTROL' SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued)

2. The rod is declared inoperable and the remainder of the rods

' in the group with the inoperable rod are alignea to within-

+ 12 steps of the inoperable rod while maintaining the rod sequence and insertion-limits of Specification 3.1.3.6. l The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

(gyu& -fJa,, y qu / -h /.3*4 J The rod is declared inoperable and the' SHUTDOWN RARGINV :

ent c' Sp::i'i:: tier 2.1 1 .; ::ti:fied. POWER OPERATION may then continue provided that; a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) Th SH"TDC'J" ""'C!" ::;;irc cr.t Of Speci'i;;tirr. :.1 1 i: det: H ed at 1:::t :n : per 12 h:er:;

Q-e)- A power distribution map ,is obtained from the movable incore #Jetectors and Fg (Z)'and FaH are verified to be

within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and c)-dt The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER.

ACTION d - Restore the inoperable rods to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during t;...e intervals when the rod position deviation monitor is inoperable, then verify the group positions at least ,

1 once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one i direction at least once per 31 days. )

4. l.3. /. 3 .rNSEKr 1 CALLAWAY - UNIT 1 3/4 1-15 Amendment No.NY,58 i

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'4.1.3.1.3 Prior to reactor criticality, the rod drop time of the individual full-length shutdown and control' rods from the fully-withdrawn position shall be demonstrated to be I less than or equal to 2.7 seconds from the beginning of decay.of stationary gripper coil voltage to dashpot entry with T vg2551*F and all reactor coolant pumps operating:

a. For all rods following each removal of the reactor vessel head, and

(. b. For specifically affected individual rods i following any maintenance on or modification-l to the Control Rod Drive System which could l_ affect the drop time of those specific rods, i

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REACTOR COOLANT SYtTEM l 3/4.4.4 RELIEF VALVES \w LIMITING CONDITION FOR OPERATION I

3.4.4 30th' power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

a. With one or both PORV(s) inoperable because of excessive seat leakage, l within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERA 8LE status or close !

the associated block valve (s)with power maintaineri tn the hinck valva (s); ;

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 HOT l SHilTDOWN within the followino 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . !

b. With one PORV 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 either restore the PORY to OPERA 8LE status, or close its associated block valve and remove power from the block ;

valve; restore the PORV to OPERABLE status within the following '

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT !

SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With both PORV(s) inoperable due to causes other than excessive seat t.

leakage, within I hour either restore at least one PORY to OPERABLE status or close its ' associated block valve ' and remove power from

~ the block valve and be in HOT STAN0BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN within the fo110 win 9 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one or both block valves inoperable, within I hour restore the l block valve (s) to OPERABLE status or place its associated PORV(s) in manual control. Restore at least one block valve to OPERABLE status within the next hour if both valves are inoperable; restore any remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. The prowistons of Specification 3.0.4 are not applicable.

SURVE1LLANCE RE0VIREHENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each ruwv shall be demonstrated OPERABLE at least once per 18 months by performance of a CilANNEL CALIBRATION of the actuation instrumentation.

4.4.4.2 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'o'rder to meet the requirements of Atil0N b. or c. in Specification 3.4.4. l' (Vd~Cb)

With all RCS cold leg temperatures above 368'F.

CALLAWAY - UNIT 1 3/4 4-10 Amendment No.83

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4.4.4.3 Both PORV p ' tion indicators s be demonstrated OPERABLE ,

at least once per 31 days erfo ce of a CHANNEL CHECK unless the 5 associated block valve is in the ed position.

4.4.4.4 Both PORV b valve position in ' ators shall be demonstrated

_ OPERABLE at le once per 31 days by perfo ce of a CHANNEL .-

CHECK unie e block valve is verified in the clos osition and power is :

removed.

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l3/4.4.5 SIEAM GENERATORS *

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t,IMITING CONDITION FOR OPERATION I '

, 3.4.5 Each steam generator shall be OPERABLE.

j APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one or more steam generators inoperable, restore t inoperable steam generator (s) to OPERABLE status prior to increasing T g above 200*F.

SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstr ted OPERABLE by performance of the following augmented inser,vice inspection ogram and the requirements.of Specification 4.0.5.

4.4.5.1 Steam Generator Sample Selection nd Inspection b Each steam generator shall be determined OPERABLE during shut n by selecting and inspecting at -

.least the minimum number of steam gener ors specified in Table 4.4-1.

4.4.5.2- Steam Generator Tube Sample election and Inspection - The steam generator tube minimum sample size, nspection result classification, and the corresponding action required shal be as specified in Table 4.4-2. The inservice inspection of steam gen ator tubes shall be performed at the fre- ;

quencies specified in Specificat on 4.4.5.3 and the inspected tubes shall be- '

verified acceptable per the ac ptance criteria of Specification 4.4.5.4. The tubes selected for each inse ce inspection shall include at least 3% of the-total number of tubes in all team generators; the tubes selected for these inspections shall be selec d on a random basis except: Y

a. Where experien in similar plants with similar water chemistry indicates cri cal areas to be inspected, then at least 50% of the

' tubes inspec d shall be from these critical areas; I

b. The first ampleoftubesdelectedforeachinserviceinspection (subsequ t to the preservice inspection) of each steam generator

j. shall 1 lude:

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REACTOR C00l' ANT SYSTEM-

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SURVEILLANCEREQUIREMENTS(Continued)-

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1) All' nonplugged tubes that previously had'detecta e' wall #

penetrations (greater than 20%),

2) Tubes in those areas where experience has in cated potential -

problems, and

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3) -A tube inspection (pursuant to Specificat on 4.4.5.4a.8)'sha11 be performed on each selected tube. If ny selected tube does not permit the passage of the eddy cur nt probe for a tube-inspection, this shall be recorded an an adjacent tube shall be selected and subjected to a tube nspection.

c. The tubes selected as the second and t rd samples (if required by Table 4.4-2) during each inservice in pection may be subjected to a partial tube inspection provided:

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1) The tubes selected for these amples include the tubes from '

s-those areas of the tube she array where tubes with imperfections were previou ly found, and

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The inspections include hose , i imperfections were pre ously' found. portions of the tubes where ,

The results of each sample inspecti n shall be classified into one of the

1 t following three categories: '

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_ Category Inspection Results 5'

C-1 Less than 5% of the total tubes inspected are .

degraded tubes and none of the inspected tubes are defective. -

C-2 One or more tubes, but not more than 1% of the t

total tubes inspecter' are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the, inspected i..

tubes are defective, N te: In all inspection's, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations -

to be included in the above percentage calculations.-

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RfACTORC001ANLSYSTEM .REVISI ' M]

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SURVE!LLANCE REQUIREMENTS-(Continued)

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3 4.4.5.3 Inspection Frequencies - The above required inservice i pections of 7

steam generator tubes shall be performed at the.following frequ ncies:

a. The first inservice inspection shall be performed ter 6 Effective , :

_s Full Power Months but withi'n 24 calendar months o initial criticality. "i Subsequent inservice inspections shall be perfo d at intervals of not less than 12 nor more than'24* calendar mon s after the previous- .

inspection. If two consecutive inspections, at including the ;

'preservice inspection, result in all-inspec on results falling into the C-1 category or if two consecutive ins ctions demonstrate that-previously observed, degradation has not c tinued and no additional degradation has occurred, the inspection nterval may be extended to ,

a maximum of.once per 40 adnths; :

b.

If the results of the inservice insp tion of a steam generator conducted in accordance with Table .4-2 at 40-month intervals fall in Category C-3, the inspection fr quency shall be increased to at -

least once per 20 months. The i rease in inspection frequency shall apply until the subsequen inspections satisfy the criteria of y . .

4 Specification 4.4.5.3a.; the i erval may then be extended to a maximum of once per 40 months and

c. Additional, unscheduled in rvice inspections shall be performed on' <

each steam specified generator in Table 4.4-in aduring cordance with the first sample inspection '

r the following'conditio : the shutdown subsequent'to any of

1) Reactor-to-seco from tube-to- beary tubes leaks (not including leaks originating-sheet welds) in excess of the Ilmits of

-X Specificatic 3.4.6.2, or -

2) A seismic ccurrence greater than the Operating Basis E'rthquake,.

a or

3) A loss of-coolant accident requiring actuation of the Engineered '

Safe Features, or

4) A ain steam line or feedwater line break. .

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ALLAWAY - UNIT 1 3/4 4-13 I :

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REACTOR C00LANTISYST REVistog I' p SURVEILLANCE REQUIREMENTS (Continued)

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4.4.5.4 Acceptance Criteria _ a . k .

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a. As used in this. specification:

1) Imperfection means an exception to the dimensio , finish or contour of a tube from that required by fabric ion drawings or specifications. Eddy-current testing indicat) ns below 20% of the nominal tube wall thickness, if detecta Te, may be considered as imperfections; .

2)

Degradation means a ser.vice-induced era ing, wastage, wear or general corrosion occurring on either tube; side or outside of a 3)

Degraded Tube means a tube containi g imperfections greater #

than or equal to 20% of the nomin 1 wall thickness caused by degradation;

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% Degradation means the perce age of the tube wall thickness affected or removed by degra ation; 5)

Defect means an imperfect n of such severity that it exceeds tne plugging limit. At e containing a defect is defective; v'

6) Plugoino limit means e imperfection depth at or beyond which h the tube shall be re. Ved from service and'is equal to 48% .

of the nominal tube all thickness; ,

7) -

Unserviceable contains a defe des ribes the condition of a tube if it leaks or rity in the ev large enough to affect its structural integ-t of an Operating Basis Earthquake, a loss-of-coolant accid t, or a steam line or feedwater line break. as specified in pecification 4.4.5.3c. , above; O,

' .' Tube Inspe tion means an inspection of the steam generator tube from^the Sint of entry (hot leg side) completely around the U-bend the top support of the cold leg; and )

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1 CAL AY - UNIT 1 3/4 4-14 ,

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i cM k rnzw a r r), REVIS y y REACTOR COOLANT SYSTEM

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SURVEILLANCE REQUIREMENTS (Continued)

9) Preservice Inspection means an inspection of the ull length of each tube in eact} steam generator performed by pddy current techniques prior to service to establish a bassline condition of the tubing. This inspection shall be perf reed prior to initial POWEC OPERATION using the equipment nd techniques ,

expected to,,be used during subsequent inse vice inspections.

b. The steam generator shall be determined OPER LE after completing the corresponding actions (plug all tubes e ceeding the plugging limit and all tubes containing through-wal cracks) required by Table 4.4-2. -

4.4.5.5 Reports a.

Within 15 days following the comple on of each inservice inspection Y

.of steam generator tubes, the numb of tubes plugged in each steam generator shall be reported to th Commission in a Special Report pursuant,to Specification 6.9.2;

-. b. The complete results of the s am generator. tube inservice inspection shall be submitted to the Con)41ssion in a Special Report pursuant to

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Specification 6.9.2 within )2 months following the completion of the inspection. This Special eport shall include:

1) Number and extent tubes inspected.

2) Location and per ent of wall-thickness penetration for each indication of a imperfection, and A,

3) Identificati of tubes plugged. *

c. Results of stea C-3. shall be eported generator tube inspections, which fall into Category to Specifica in a Special Report to the Commission pursuant plant opera on. on 6.9.2 within 30 days and prior to resumption of gations co This report shall provide a description of investi-correctiv ucted to determine cause of the tube degradation and measures taken to prevent recurrence.

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CA LAWAY - UNIT 1 3/4 4-15

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, , . TABLE 4.4-1 E MINIMUM NUMBER OF STEAM CENER ATORS TO BE .-

l INSPECTED DURfNG INSERVICE INSPECTION +

Preserkspecten No Yes

No. of Steam toes per Unit ,

Two Three Four Two Three Foue First inservice Inspection All One Two Two Second & Subsequent Inservice Ichpections One' One l One2 One 3 x -

h TABLE NOTATIONS U l. The inservice inspection may be limited to Meern generator on a rotating schedu$ enc 3 N % of tie tubes '

a

fwhere N is the number of stearn generators in IQtl if the results of the first or previous inspections indicate that alt steam generators are performing in a like enanner. re_cte that tender some circuenstances. she operating conditions in h

O one or more steam generalors may be found to be enore e than those in other steam generators. Under such circum 05 stances the sampft sequence shall be modified to inspect the severe conditions. .~

2. The offer steam generator not in1pected during the first inservice in ' shall be inspected. The third and subsequent inspections should follow the instructions described in 1 'above.

1 Each of the other two steam generators not inspected during the first inservice ' ons shall be inspected during the second and thisd inspections. The fourth and subsequent inspections shaft fonow the in tions described in I above.

, o s

2

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% TABLE 4.4-2 ,

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STEAM GENERATOR TUBE INSPECTION E

. N IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION E Sam Site Result Action Required Result Action Required Result Action Required

~ A minimum C-1 None N. A. ' N. A. N. A. N. A.

S Tubes per \

S. G. -

C-2 Pidg defective tubes C-1 None N. A. N. A.

inspect additional N 25 Plug defective tubes C-1 None en this S. G. C and inspect additional 4S tubes in this S. G.

C-2 Plug defective tubes k -

Perform action for -

C-3 C-3 result of first sample

. . . s Perform action for 'g

-l o ,

C- C-3 result of first N. A. N. A.

) qmple .

p C-3 Inspect all tubes in . Allother . . .

this S. G., plug de-

~

- S. G.s are ne

" N. A. N. A.

fective tubes and . C-1 . .

inspect 2S tubes is.

- S me S. G.s N. A.

each other S. G. Perform action ik N. A.

C-2 but no C12 result of second addstsonal ,,,p ,,

Notification to NRC S. G. are pursuant to 550.72 C-3 lb)(2) of 10 CFR Additional inspect all tubes in

Part 50 S. G. is C-3 each S. G. and plug defective tubes.

Notification to NRC N. A. N. A. -

pursuant to $50.72 *

.4 - 5 (b)(2) of 10 CFR ITI Part 50 -

$ a S = 3 h Where N is the number of steam generators in the unit, and n is the number of steam generators inspected -

during an inspection

$ g'g n E v' 3

>a Uf.

6 l

3/4.6 CONTAINMENT SYSTEMS

? 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within

- I hour 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 SNUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

. a. At least once per 31 days by verifying that all penetrations

not I capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by manual valves, blind flanges, or deactivated automatic valves secured in tneir closed positions, except as provided in Table 3.6-1 of Specification 3.6.3; I

b. By verifying that each containment a r lock i n compliance with the requirements of SpecificatJon 3.6.1 3; eet-

c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure not less tha 48.1 p verifying that when the measured leakage rate for thIs,e seals is added the leakage ra es determined pursuant to Specification 4.6.1.@. for all other T e a d C penetrations, the combined leakage ra islesthan0.6,Lj

> ZNSEAT A Ex alves flang.c eactivated automatic valves which are ocated inside the containment and are locked, sealed or otherwise secured l in the closed position. These' penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

3/4 6-1 Amendment No. 73, 62 CALLAWAY - UNIT 1 l

-j 1

< . INSERT A

d. By performing' containment leakage' rate testing, except for containment air locks, in accordance with-10 CFR 50, Appendix J, as modified.by approved exemptions; and

e. By . verifying containment structural integrity in = '

accordance with die Containment Tendon Surveillance.

Program of Specification 6.8.5.c.

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,,_ _. a . D C0'NTAINMENT SYSTEMS.

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SURVEILLANCE REQUIREMENTS-i 4.6.1.7.1 Each 36-inch containment shutdown purge supply and exhaust isolation 'i valve (s)* shall be verified blank flanged and closed at least once per 31 days.. ,

t 4.6.1.7.2 Each 36-inch containment shutdown purge supply and exh'aust isolation i valve and its associated blank flange shall be leak tested at least once per !

24 months and following each reinstallation of the blank flange when pressurized :

to P me d leakage rate for these l valv$s,and48.1flanges, psig, and verifying including that wskage stem s added o the leakage rates. r determined pursuant to Specificati n 4.6.1 . for al other Type 8 and C penetrations, the combined leakag rate is ess than 0.60 L,. ,

.I .

4.6.1.7.3 The cumulative time that a 18-in tainment mini-purge supply and exhaust isolation valves have been n during a calendar year shall be determined at least once per 7 days.

4.6.1.7.4 At least once per 3 months each 18-inch containment mini purge -

supply and exhaust isolation valve with resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than-0.05 L, when pressurized to P,.

3

)

~

I 4

l.

l-

Except valves and flanges which are located inside containment. These valves shall be verified to be closed with their blank flanges installed prior to entry into MODE 4 following each COLD SHUTOOWN.

l I

CALLAWAY - UNIT 1 3/4 6-12 Amendment No. 13 lQl :

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REVIS j ftFACTOR__ COOLANT SYSTEM

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%j ' BASES

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Usw Cwsvs# * 's o i n. s EsQu $ i

"......is (si ina .istien LI U.; at; .~; sCT.s. stei 14;;

on ure that the structural integrity of this portion of the RCS will be main- s"

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d. The program for inservice inspection of steam generator tubes is based n a modification of Regulatory Guide 1.83, Revision 1. Inservice nspect n of steam generator tubing !s essential in order to maintain s veil-ance of e conditions of the tubes in the event that there is eviden of nechanical amage or progressive degradation due to design, manufact ng orrors, or i ervice conditions that lead to corrosion. Inservice ispection ,

of steam gener or tubing also prov. ides a means of characterizing he nature

.ind cause of any ube degradation so that corrective measures c be taken.

Unscheduled in rvice inspections are performed on each team generator

'ollowing: (1) react to secondary tube leaks; (2) a sei ic occurrence ureater than the Operat g Basis Earthquake; and (3) a 1 s-of-coolant accident equiring actuation of th Eng*neered Safety Features, ich for this Specification is defined t be a break greater than t t equivalent to the

- everance of a 1" inside dia ter pipe, or, for a m n steamline or feedline, a break greater than that equi lent to a steam g erator safety valve

'alling open; to ensure that ste ' generator tub s retain sufficient integrity s 'or continued operation. Transien s less seve than these do not require J9 nspections because the resulting s esses a well within the stress criteria Q',. , '

ostablished by Regulatory Guide 1.121, whi unplugged s' team generator tubes oust be' capable of withstanding.

c oolant The plant is expected to be oper ed 1 a manner such that the secondary will be maintained within t se chem try limits found to result in n egligible corrosion of the steam enerator tu

s. If the secondary coolant c

hemistry is not maintained with'n these limits, localized corrosion may ikely result in stress corros n cracking. The e tent of cracking during plant operation would be lim ed by the limitation steam generator tube eakage between the Reactor Coolant System and the 5e ndary Coolant System i

reactor-to-secondary lea age = 500 gallons per day'per team generator).

Cracks o

having a reactor o-secondary leakage less than th's limit during peration will have a adequate margin of safety to withst d the loads imposed during normal opera on and by postulated accidents. Operat g plants have demonstrated that eactor-to-secondary leakage of 500 gallons r day per s

team generator an readily be detected by radiation monitors o team generator hinwdown. Lea ga in excess of this limit will require plant shut own anc; an unscheduled spection, during which the leaking tubes will be loca d and p luggetf.

Wa age-type defects are unlikely with proper chemistry treatment o the

! econ ry coolant. However, even if a defect should develop in service, d.,

vil be found during scheduled inservice steam generator tube examinations.

P I gging will be required for all tubes with imperfections exceeding the

W ng ' Mt of M cf t": t'M; ner'n:1 _::!' thi'ne:: . Ete g:ncr:ter CALLAWAY - UNIT 1 8 3/4 4-3

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I REVISloy j -

REACTOR COOLANT SYSTEM T

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C'.". r."_?. !C_"._; ' C; 7 t i ne d ) _k_M M #

'.,im i . n; e ' ai;; re t i;bba t.,' 5 i-- e n: t r:t d th; : 4 !'ity t; gg .

e eliab ct degradation that has penetrated 20% of the original a1 +

hickness. Resu om WCAP-10043 have been used to establ ugging limit.

Whenever the results o eam generator g inservice inspection

'all into Category C-3, these results ~ reported to the Commission pursuant to Specification 6.9.2 to resum of plant operation. Such cases will be considere

e Commission on a" case- basis and may esult in a re ent for analysis, laboratory examinations, additional

e ddy-c - inspection, and revision of the Technical Specifications, i ,

ry.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS Leakage Detection Systems required by this specification are -

provided to monitor and detect leakage" from the reactor coolant pressure --

boundary. These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakaga Detection 4, Systems." May 1973.

kN e

q 3/4.4.6.2 OPERATIONAL LEAKAGE

.y ,

i. PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may g be indicative of an impending gross failure of the pressure boundary. Therefore, g%v%?f

. s. . the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

.m

[ Industry experience has shown that while a limited amount'of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less t'an 1 gpm. This threshold value is sufficiently low to ens.ure early detection of additional leakage.

The total steam generator tube leakage limit of 1 gpm for all steam generators not isolated from the RCS ensures that the dosage contribution from the tube leakage will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of either a steam generator tube rupture or steam line break. The 1 gpm limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakcqe limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture or under LOCA conditions.

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interf ere with the deter. tion nf UNIDENTIFIED LEAKAGE Dy the Leakage Detection Systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total fiow f rom the reactor cnolant pump seals exceeds 8' gpm per RC pump at e nominal RCS pre ...ure of ??35 psig. This limitation ensures adequate performance of the RC pump .p.ih.. [' .

CAIiAWAY - UNil 1 B 3/4 4-4 '

i

. ADMINISTRATIVE CONTROLS 1: :

t PROCEDURES AND PROGRAMS (Continued) ,

l f.9,-

Radiolocical Environmental Monitorino Procram (Continued)

3) Participation in a'Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy ;

of the measurenents of radioactive materials in environmental i

, sample matrices are perfonned as part of the quality assurance ;

j g grog am for environmental monitoring.

6.9- REPORTING REQUIREMENTS ROUTINE REPORTS

~

6.9.1 In addition to the applicable reporting requirements of Title 10. Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted.

STARTUp REPORT

.6.9.1.1 A sumary report of plant startup and power escalation testing shall be

.. submitted following: (1) receipt of an Operating License (2) amendment to .the License involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that 'may have significantly alterad the nuclear.

3, thermal, or hydraulic perfonnance of the plant. .

~

- 6.9.1.2 The Startup Report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in-license conditions based on other comitments shall be included in this report.

6.9.1.3 Startup Reports shall.be submitted within: (1)90daysfollowing completion of the Startup Test Program, (2) 90 days following resumption or comencement of comercial power operation, .or (3) 9 months following initial.

criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of comercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality, i CALLAWAY - UNIT 1 6-19 Anendment No. 27, 50

n V.

INSERT 9 (page 1 of 2)

The'following programs,. relocated from the Technical Specifications to FSAR Chapter 16, shall be implemented and maintained:

a. Exolosive Gas and Storace Tank Radioactivity Monitorino Procram This program provides controls for potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM, the quantity of radioactivity contained in gas storage tanks, and the quantity.of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

1. The limits for concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM and a surveillance program to ensure the limits are maintained.

2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 20.5 rem to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY in the event of an uncontrolled release of the tanks' contents, consistent with Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases due to. Waste Gas System Leak or Failure," in NUREG-0800, July 1981.

3. A surveillance program to ensure that the quantity of radioactivity contained in the following outdoor liquid radwaste tanks, that i are not surrounded by liners, dikes, or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area ,

drains connected to the liquid radwaste system, ;

is less than the amount that would result in' [

concentrations less than the limits of 10 CFR ,

Part 20.1 - 20.602, Appendix B (redesignated at 56FR23391, May 21, 1991) at the nearest potable -

water supply and the nearest surface water supply in an UNRESTRICTED AREA, in the event of ;

an uncontrolled release of the tanks' contents: '

a. Reactor Makeup Water Storage Tank, i

b. Refueling Water Storage Tank,

c. Condensate Storage Tank, and D

L INSERT 9 (page 2 of 2)

d. Outside temporary tanks, excluding demineralizer vessels and the liner being used to solidify radioactive waste.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

b. Reactor Coolant Pump Flywheel InsDection Procram Each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, dated August 1975.

c. Containment Tendon Surveillance Procram This program provides controls for monitoring tendon performance, including the effectiveness of the tendon corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial plant operation as well as periodic testing thereafter. The Containment Tendon Surveillance Program, and its inspection frequencies and acceptance criteria, shall be in accordance with the Callaway position on proposed Revision 3 of Regulatory Guide 1.35 dated April 1979.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Containment Tendon Surveillance Program inspection frequencies.

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TECHNICAL SPECIFICATION RETYPED PAGES

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INDEX l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE INSTRUMENTATION (Continued)

Wa ste G a s H old up Syste m .. ..... ....... .. . . . . ...... ... .... . . . ... . . . . . .. ... .. ... Deleted 3/4.3.4 TURBINE OVERSPEED PROTECTION ........................................ Deleted 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION S ta rtup a nd Powe r O pe ration .. .. . ..... .... .. ... .. . ... . . .. .... . . ... . . . .... . . . . 3/4 4- 1 H o t S t a nd by . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-2 H ot S h ut d o w n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-3 Cold Shutd own - Loo ps Fille d . . . . .... .... ......... . ... . . . . . .... .... . .. . .. . . ... 3/4 4-5 Cold Shutdown - Loops Not Filled ........................................... 3/4 4-6 ,

3/4.4.2 SAFETY VALVES S h ut d o w n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . De l e t e d O pe r a t i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4- 8 3/4.4.3 PR E S S U R IZ E R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4-9 3/4.4.4 R E LI E F VA LV E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 4- 10 3/4.4.5 STEAM G EN ER ATO RS . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .~. 3/4 4- 1 1 Table 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION................. 3/4 4-16 Table 4.4-2 STEAM GENERATOR TUBE INSPECTION ........................ 3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Le a ka g e Detec tion Syste m s . .. . ... .. . . . .. . .. .... .. . ..... . . ...... ... ... ... . . . .. 3 /4 4- 18 0 pe ra tional Le a ka g e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4- 19 t

CALLAWAY - UNIT 1 Vil Amendment No. 50

f INDEX BASES

PAGE SECTION REACTOR COOLANT SYSTEM (Continued) 3/4.4.5 STEAM G EN ERATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .................................B 3/4 4-4 3/4.4.7 CH EMI STRY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Delet ed I

i 3/4.4.8 S PE CI FI C A CTIVITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS ..........................................B 3/4 4-6 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS.................................B 3/4 4-10 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E> 1MeV) AS A FUNCTION - ,

OF FULL POWER SERVICE LIFE ................................B 3/4 4-12 l l

I FIGURE B 3/4.4 2 EFFECT OF FLUENCE AND COPPER ON SHIFT OF

, RTNDT FOR REACTOR VESSEL STEELS EXPOSED l TO IRRADIATION AT 5 50 F..................................... B 3/4 4-13 1 3 /4.4.10 STR UCTU RAL INTEG RITY. .... . .. . . ... ....... . .... . . . . .. . . . .. ... . . ... . . ... . . . ... Deleted 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ..................................... Deleted 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 A C C U M U LATO R S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 /4 5 - 1 3/4.5.2, 3/4.5.3, and 3/4.5.4 ECCS SUBSYSTEMS................................ B 3/4 5-1 3/4.5.5 REFUELING WATER STORAGE TANK....................................... B 3/4 5-3 l

l 3/4.6 CONTAINMENT SYSTEMS

3/4.6.1 PRIM A RY CONTAIN M ENT.. ..... .. . . .. . .. . . . .. ... .. .. . . . ...... . .... .. . . ..... . ... B 3 /4 6- 1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS........................B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ....................................... B 3/4 6-4 3/4.6.4 COM BU STIBLE G AS CONTR O L . ..... . . . .. . .. .. . .. . . .. . . .. .. . . . .. . .. ... . . . . . .. B 3/4 6-4 CALLAWAY - UNIT 1 XV

y .. _. _. __ _ _ _ .

}

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DEFINITIONS

)

CONTAINMENT INTEGRITY 1.7. CONTAINMENT INTEGRITY shall exist when:

a. - All penetrations required to be closed during accident conditions are either:

4 .

1) Capable of being closed by an OPERABLE containment automatic isolation ,

valve system, or '

2) . Closed by manual valves, blind flanges, or deactivated automatic valves 1 secured in their closed positions, except as provkled in Table 3.6-1 of Specification 3.6.3.

]

b. All equipment hatches are closed and sealed.

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3.

d. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

e. . The containment leakage rates are determined per Specification 4.6.1.1.d and are ,

within the limits listed in the Bases of Specification 3.6.1.1, and

f. Structural integrity is assured via the program described in Specification 6.8.5.c. !

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel. . Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT l 1.10 The CORE OPERATING LIMITS REPORT (COLR)is the unit specific document that l provides core operating limits for the current operating reload cycle. The cycle specific core operating limits shall be determined for each reload cycle in accordar.ce with Specification 6.9.1.g.

Plant operation within these operating limits is addressed in individual specifications. ,

i DOSE EQUIVALENT I-131

i 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which j alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." i CALLAWAY- UNIT 1 1-2 Amendment No. 15,35,58 i

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+

REACTIVITY CONTROL SYSTEMS n

= LIMITING CONDITION FOR OPERATION ACTION (Continued)

2. The rod is declared inoperable and the remainder of the rods in the group with .

the inoperable rod are aligned to within 12 steps of the inoperable rod while q maintaining the rod sequence and insertion limits of Specification 3.1.3.6.

The THERMAL POWER level shall be restricted pursuant to Specification l

, 3.1.3.6 during subsequent operation, or i

E

3. The rod is declared inoperable and the SHUTDOWN MARGIN is greater than or w equal to 1.3% Ak/k. POWER OPERATION may then continue provided that: .,

l a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) A power distribution map is obtained from the movable incore detectors and Fo(Z) and F$ are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and c) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following

hours the High Neutron Flux Trip Setpoint is reduced to less than or equa i to 85% of RATED THERMAL POWER.

ACTION d - Restore the inoperable rods to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS l 'l 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except ~ !

during time intervals when the rod position deviation monitor is inoperable, then verify the I group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . l 4.1.3.1.2 Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

4.1.3.1.3 Prior to reactor criticality, the rod drop time of the individual full-length shutdown and control rods from the fully withdrawn position shall be demonstrated to be -l l

less than or equal to 2.7 seconds from the beginning of decay of stationary gripper coil !

voltage to dashpot entry with T,2551'F and all reactor coolant pumps operating: j

a. For all rods following each removal of the reactor vessel head, and

b. For specifically affected individual rods following any maintenance on or '

modification to the Control Rod Drive System which could affect the drop time of those specific rods.

CALLAWAY - UNIT 1 3/4 1-15 Amendment No. SA,58 !

L__ _u______________________._______________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ _ .

s.

32 3/4.6 CONTAINMENT SYSTEMS [

3/4.6.i PRIMARY CONTAINMENT .

CONTAINMENT INTEGRITY ,

1 r

LIMITING CONDITION FOR OPERATION 1

3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

l APPLICABILITY: MODES 1,2,3 and 4. j ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY 'within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRlTY shall be demonstrated:

l

a. At least once per 31 days by verifying that all penetrations

not capable of ;

~

being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by manual valves, blind flanges, or deactivated automatic valves' secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3; !

b. By verifying that each containment air lock is in compliance with the ,

requirements of Specification 3.6.1.3;

c. After each closing of each penetration subject to Type B testing, except the containment air locks,if opened following a Type A of B test, by leak rate testing the seal with gas at a pressure not less than P,,48.1 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.1.d for all other Type B and C penetrations, the combined leakage rate is less than 0.60 L,; )

d. By performing containment leakage rate testing, except for containment air locks,in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions; and

e. By verifying containment structural integrity in accordance with the Containment Tendon Surveillance Program of Specification 6.8.5.c.

Except valves, blind flanges, and deactivated automatic valves which are located inside '!

the containment and are locked, sealed or otherwise secured in the closed position.

l These penetrations shall be verified closed during each COLD SHUTDOWN except that

such verification need not be performed more often than once per 92 days.

CALLAWAY JNIT 1 3/4 6-1 Amendment No. YS,62 4

___n_,__ _ _ _ _ _ -_.__ __ _ _ _ _ . . - _

j l

4 CONTAINMENT SYSTEMS i SURVEILLANCE REQUIREMENTS ?

4.6.1.7.1 Each 36-inch containment shutdown purge supply and exhaust isolation ;

valve (s)" shall be verified blank flanged and closed at least once per 31 days. -

4.6.1.7.2- Each 36-inch containment shutdown purge supply and exhaust isolation -

valve and its associated blank flange shall be leak tested at least once per 24 months and following each reinstallation of the blank flange when pressurized to P,,48.1 ;

psig, and verifying that when the measured leakage rate for these valves and flanges, including stem leakage, is added to the leakage rates determined pursuant to Specification 4.6.1.1.d for all other Type B and C penetrations, the combined leakage !

rate is less than O.60 L, .

4.6.1.7.3 The cumulative time that all 18-inch containment mini-purge supply and ;

exhaust isolation valves have been open during a calendar year shall be determined at - !

least once per 7 days. !

~4.6.1.7.4 At least once per 3 months each 18-inch containment mini-purge supply.-

and exhaust isolation valve with resilient material seals shall be demonstrated .

OPERABLE by verifying that the measured leakage rate is less than 0.05 L, when pressurized to P, .

l a

J i

Except valves and flanges which are located inside containment. These valves shall be verified to be closed with their blank flanges installed prior to entry into MODE 4 .>

following each COLD SHUTDOWN. .;

4 4- l CALLAWAY - UNIT 1 3/4 6-12 Amendment No.13

r x

}

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) -

Radiolooical Environmental Monitorina Proaram (Continued) -

f.-

' 3) - Participation in an Interlaboratory Comparison Program to ensure that . 'I independent checks on the precision and accuracy of the measurements ;

of radioactive materials in environmental sample matrices are performed I as part of the quality assurance program for environmental monitoring. q l

6.8.5 The following programs, relocated from the Technical Specifications to FSAR Chapter 16, shall be implemented and maintained-

a. Exolosive Gas and Storaae Tank Radioactivity Monitorina Proaram This program provides controls for potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM, the quantity of radioactivity ,

contained in gas storage tanks, and the quantity of radioactivity contained in s unprotected outdoor liquid storage tanks. i The program shall include:

1. The limits for concentrations of hydrogen and oxygen in the WASTE GAS :

HOLDUP SYSTEM and a surveillance program to ensure the limits are q maintained.

2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of it 0.5 rem to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY in the event of an uncontrolled release of the tanks' contents, consistent with Branch Technical Position -

ETSB 11-5, " Postulated Radioactive Releases due to Waste Gas System Leak or Failure," in NUREG-0800, July 1981.

3. A surveillance program to ensure that the quantity of radioactivity- ;

contained in the following outdoor liquid radwaste tanks, that are not ;

surrounded by liners, dikes, or walls capable of holding the tanks' -!

contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste system,is less than the amount that ;

would result in concentrations less than the limits of 10 CFR Part 20.1 -

20.602, Appendix B (redesignated at 56FR23391, May 21,1991) at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA,in the event of an uncontrolled release of the ;

tanks' contents:

a. Reactor Makeup Water Storage Tank,

b. Refueling Water Storage Tank, CALLAWAY - UNIT 1 6-19 Amendment No. 2F, 50 l

ro .

i

[ l

c. Condensate Storage Tank / and . .

~ d. _ Outside temporary tanks, excluding dem'neralizer vessels and the liner being used to solidify radioactive waste.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the j

' Explosive Gas'and Storage Tank Radioactivity Monitoring Program '

l surveillance frequencies.

b. Reactor Coolant Pumo Fivwheel Insnection Proaram : ].

. . i Each reactor coolant pump flywheel shall be inspected per the . .

recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14,- '

Revision 1, dated August 1975. :

c. Containment Tendon Surveillance Proaram l This program provides controls for monitoring tendon performance, including the effectiveness of the tendon corrosion protection medium, to ensure l containment structuralintegrity. The program shallinclude baseline measurements prior to initial plant operation as well as periodic testing !

thereafter. The Containment Tendon Surveillance Program, and its inspection

~

l' frequencies and acceptance criteria, shall be in accordance with the Callaway position on proposed Revision 3 of Regulatory Guide 1.35 dated April 1979. {.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Containment Tendon Surveillance Program inspection frequencies. ,

6.9 REPORTING REQUIREMENTS :

o 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional i Administrator of the NRC Regional Office unless otherwise noted.

STARTUP REPORT !

i 6.9.1.1 A summary report of plant startup and power escalation testing shall be !

submitted following: (1) receipt of an Operating License, (2) amendment to the . 1 License involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) ;

modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The Startup Report shall address each of the tests identified in the FSAR and ;

shallinclude a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this ;

report. ;

I CALLAWAY - UNIT 1 6-19a Amendment No. 27,50 i l

6.9.1.3 Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of ,

commercita power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Otartup Test Program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all -l three events have been completed.

ANNUAL REPORTS

)

6.9.1.4 Annual Reports covering the activities of the unit c:: oescribed below for the l

- previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality.

r

+

t t

t CALLAWAY - UNIT 1 6-19b Amendment No. 27,50

i ULNRC-0 3184 ATTACHMENT FOUR REPLACEMENT PAGES FOR ATTACHMENT 5 TO ULNRC-3023

l

-l TABLE 1 l Summary of Criteria Application Results Instmmentation Tech STS Rev. 5 Technical Specification NRC Callaway Note Spec Number Title . Results Results Number 3.3.1 3.3.1 Reactor Trip System Retain Retain l Instrumentation -i 3.3.2 3.3.2 Eng. Safety Feature Retain Retain -l Actuation System .

Instrumentation 3.3.3.1 3.3.3.1 Radiation Monitoring Retain Retain Instrumentation 3.3.3.2 3.3.3.2 Movable Incore Detectors Relocate Relocate 3.3.3.3 3.3.3.3 Seismic Instrumentation Relocate Relocate 3.3.3.4 3.3.3.4 Meteorological Relocate Relocate Instrumentation 3.3.3.5 3.3.3.5 Remote Shutdown Retain Retain Instrumentation 3.3.3.6 3.3.3.6 Accident Monitoring Retain Retain 5 Instrumentation 3.3.3.8 3.3.3.9 Loose Parts Detection Relocate Relocate System 3.3.3.10 Explosive Gas Monitoring Not Relocate 6 Instrumentation Reviewed ,

C :

3.3.4 3.3.4 Turbine Overspeed Relocate Relocate - l Protection j

I l

l i

4

TABLEI Summary of Criteria Application Results Reactor Coolant System ;

Tech STS Rev. 5 Technical Specification NRC Callaway Note l Spec Number Title Results Results Number ;

3.4.1.1 3.4.1.1 Reactor Coolant Loops Retain Retain -

and Coolant Circulation 3.4.1.2 3.4.1.2 RCS Hot Standby Retain Retain 3.4.1.3 3.4.1.3 RCS Hot Shutdown Retain Retain 3.4.1.4.1 3.4;1:4 1 Cold Shutdown Loops Retain Retain Filled 3.4.1.4.2 3.4.1.4.2 Cold Shutdown Loops Retain Retain Not Filled 3.4.2.1 3.4.2.1 Safety Valves -Shutdown Relocate Relocate 3.4.2.2 3.4.2.2 Safety Valves -Operating Retain Retain 3.4.3 3.4.3 Pressurizer Retain Retain Retain _ j '

3.4.4 3.4.4 ReliefValves RAain 3.4.5 3.4.5 Steam Generators /~ R i ; :. ; n;:- 2

-e-3.4.6.1 3.4.6.1 Leakage Detection Systems

[ Retain [/8 tain /'efufn ri 3.4.6.2 3.4.6.2 Operational Leakage ( Retain Retain 3.4.7 3.4.7 Chemistry ( Relocath Relocate 9, 3.4.8 3.4.8 Specific Activity Retain Retain 3.4.9.1 3.4.9.1 Pressurefremperature Retain Retain Limits 3.4.9.2 3.4.9.2 Pressurizer Relocate Relocate Pressure / Temperature 3.4.9.3 3.4.9.3 Overpressure Protection Retain Retain System 3.4.10 3.4.10 Stmetural Integrity Relocate Relocate 10 3.4.11 3.4. I 1 RCS Vents Relocate Relocate '

- - . - - - - - . - - - - - , - .-----,--a -

TABLEI Summary of Criteria Application Results Containment Systems Tech STS Rev. 5 Technical Specification NRC Callaway Note Spec Number Title Results Results Number C 3.6.1.1 3.6.1.1 Containment Integrity Retain Retain / 12,/S 3.6.l.2 3.6.1.2 Containment Leakage See Note 12' 12 Note 12 (# -

3.6.1.3 3.6.1.3 Containment Airlocks Retain Retain 3.6.1.4 3.6.1.5 Internal Pressure Retain Retain 3.6.1.5 3.6.1.6 Air Temperature Retain Retain 3.6.1.6 3.6 1.7 Contain. Vessel Structural Relocate Relocate 13 Integrity 3.6.1.7 3.6.1.8 Containment Ventilation Retain Retain 14 System 3.6.2.1 3.6.2.1 Containment Spray Retain Retain System 6.2.2 3.6.2.2 Spray Additive System Retain Retain J 6.2.3 Containment Cooling Retain Retain System 3.6.3 3.6.3 Containment Isolation Retain Retain Valves 3.6.4.1 3.6 4.1 Hydrogen Analyzers Retain Delete 15 3.6.4.2 3.6.4.2 Hydrogen Control System Retain Retain

.. - .. .- . . - . . - . - . . . . - - - -~ ..

TABLE 1 Summary of Criteria Application Results Plant Systems Tech STS Rev. 5 Technical Specification NRC Callaway Note -

Spec Number Title Results Results Number 3.7.1.1 3.7.1.1 Safety Valves Retain Retain -

3.7.1.2 3.7.1.2 Auxiliary Feedwater Retain Retain :

System .*

3.7.1.3 3.7.1.3 Condensate Storage Tank Retain Retain

. 3.7.1.4 3.7.1.4 Specific Activity Retain Retain -

3.7.1.5 3.7.1.5 Main Steam Isolation Retain Retain Valves 3.7 l.6 Main Feedwater Isolation Not Retain Valves Reviewed 3.7.1.7 Steam Generator Not Retain Atmospheric Steam Dump Reviewed Valves 3.7.2 3.7.2 Steam Generator Relocate Relocate Pressure / Temperature Limits 3.7.3 3.7.3 Component Cooling Retain Retain Water .

3.7.4 3.7.4 Essential Senice Water Retain Retain {

System ;

3.7.5 3.7.5 Ultimate Heat Sink Retain Retain 3.7,6 Control Room Emerg. Retain Retain Ventilation Syrtem 3.7.7 3.7.8 Emerg. Exhaust System - - Retain Retain e ,

~

Auxiliary Building /

3.7.8 3.7.9 Snubbers Relocate Relocate [ -M-3.7.9 3.7.10 Scaled Source Relocate Relocate Contamination 3.7.12 3.7.13 Area Temperature Relocate Relocat( @

Monitoring / >

i i

b 8

w- , , , . , .n---- , - - , , - - ,--~- - - . . + . - . , -. - - = - - .

TABLE 1 Summary of Criteria Application Results Radioactive Efiluents Tech STS Rev. 5 Technical Specification NRC Callaway Note Spec Number Title Results Results Number ,

C 3.I1.1.4 3.1 ! . l .4 Liquid Ifoldup Tanks Relocate Relocate / -Be 4 3.I1.2.5 3.11.2.5 Explosive Gas Mixture Relocate Relocate ( 6 3.I1.2.6 3.I1.2.6 Gas Storage Tanks Relocate Relocate / 4

5, and 6 (except when the RV head is removed).

This is an operating restriction of the reactor vessel cold overpressure analysis.

This SR will be retained under LCO 3.5.4, ECCS Subsystems - Tavg $200 F for Modes 5 and 6.

s The footnote to 3.1.2.3 is deleted because it is redundant to the footnote for Specification 3.5.4. SR 4.5.3.2 addresses Mode 4.

3. The NRC review of LCO 3.1.3.2 and LCO 3.1.3.3 concluded that they could be relocated.

However, if an associated SR is necessary to

' meet the operability requirements for a retained LCO, the SR should be relocated to the retained LCO. Our evaluation found that LCO 3.1.3.2 is associated with a transient analysis initial condition _and supports LCO 3.1.3.1. As such, LCO 3.1.3.2 will be retained as is. The surveillance associated with LCO 3.1.3.3 is not required for any retained LCO and, therefore, SR 4.1.3.3 will be relocated.

4. The NRC review of this LCO concluded that it could be relocated. However, if an associated SR is necessary to meet the operability requirements for a retained LCO,.the SR should be relocated to the retained LCO. SR 4.1.3 i *cquir^ h cnc ~ *ab '.

control rods under LCO 3.1.3.1 and will be }

retained under that LCO,w1 u uLe red J.cy time-

-li-4* gi"cr in ncu "CA" Scction 1C.1.2.2.

This is consistent with STS.

A N

5. The Regulatory Guide l.97, Rev. 2, Type A variables identified in FSAR Appendix 7A are retained. The neutron flux (Gamma-Metrics) and RVLIS instrumentation will be added. The non-Type A variables are identified and evaluated on the screening form. The relocated instruments are:

_f V Containment Pressure - Extended Range PZR Safety Valve Position Indication

[#AV dt/8dn M/c,rdr Unit Vent High Range Noble Gas Monitor. ,

f4AVllock Ulve ~~*'

/,ff g,, gfy.j.,y 0? Pl ^ct; Val ? Pccitica Indicateys a{_ , "euccu jnd PC' ueleceu uivm acuuu.uul speu2&.caticr-

-3.3.3.7 '"d Tcnthly chOn"c1 chcch; hiic L6en I

-cdd;d tc LCO 2.4.4 ac diccucccd in the Enfet"-

Z u2uu-.um, Aemcch~^-- 2 A __

^

y and Nor9015nlc $clinc/fg},

6 This specification will be relocated and an Explosive GasVMonitoring Program ctcteront-will be incorporated into new Section 6.8.5.

7. -Thio oyecificction will hc relccated md ?

-Turh_uu Overspeed stucevuivu Reliabil m -

-Progccr ctctcment '-cill 'O inco Se/s h [r cccted into-

-ncw CccLwu 6.5.5.

l 8. -Thio opecliiuacivu ill hc r; loc &-ud and a

- S c cain Genetauvt LLm Surecilicncc Program etctcm:nt 1.e.:. willLei[c-ludcdinncuCcction

.be,/eh

9. Thic cpccificction till 50 rciccatcd and c-T w l

-Prir' q wator bc inc1 9 cham 4" nc': 4stry Progra.

Ccction C.C. ctatc$ cnte/Me[l f

A- % #

10. The LCO will be relocated and the associated SR regarding RCP flywheel integrity will be retained in new Section 6.8.5 as a programmatic requirement.

11. This LCO is intended to prevent loss of the decay heat removal function in Mode 5 and Mode 6 with vessel head installed by allowing SI pumps to be operable when the water level is below the vessel flange. The LCO will be retained. Consideration was given to incorporating the restrictions on pump operation into LCO 3.4.9.3, Overpressure Protection, which would have been in conformance with the STS approach. However, the Modes and RCS temperatures for which these specifications apply prevented combining them y, f,f, p

12. Containment testing is a requirement mposed by Appendix J of 10 CFR 50. TL_c LCO ill be relocated; however, the values of parameters defining leakage limits from 3.6.1.2 will be retained under the Containment Integrity 1 1 Bases. SR 4.6.1.1.c will be modified to 4 //4 {

l c limic.c L a -cud r r:1ccotcd reference

.uoimymtog 8pecification@/

r:f e.uuue- f, g, /, /, .&./- I

= =c; ,..:.ca ; ,-,ke SaR 3.6.t. 4

.seu m s.a.1.1. Agen].c,xIscaso,

r* l

13. 3+ers )?pecificationYwill be relocated and a Containment Tendon Surveillance Program j

- e t a t ;;.e u . will be incorp I Section 6. 8.5. fl/aw ff Y. orated into newl./,/,e sof// de adde) \

b ' j'ltthexl~ N [.r ,*vvye m ,

k A w

14. SR 4.6.1.7.2 will be modified to eliminate reference to a specification that was relocated and instead reference ccrrcc- cding-rc= cmmuivu 16.6.1.1. n e w J T 4 4,/,/,2 ,

15. LCO 3.6.4.1 is deleted since it is redundant to LCO 3.3.3.6 and is obsolete per the STS.

16. Thic cpccification cill bc rciccated and c-

-Cnubbcr Incpcetics P*^0"== o*=*oman* u4

-included in ne- Occticn f.S.5. da/s-/so[he

17. 'Thic cpecification will be relocated 2nd n-

-Art; Tc...peraturcMonitoring"regrns.steteb.

l u Lm inmluded in new seucivu 6.0.5. ente /efe/

k A A. x

18. This specification places a lower limit on the amount of water above the top of the fuel assemblies in the reactor vessel during movement of control rods. The Bases state that this ensures the water removes 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly in the event of a fuel handling accident (FHA).

However, the movement of control rods is not associated with the initial conditions of an FHA, and the Bases do not address any concerns regarding inadvertent criticality which could lead to a breach of the fuel rod cladding.

Inadvertent criticality during Mode 6 is prevented by maintaining proper boron concentration in the coolant in accordance with LCO 3.9.1. Therefore, this LCO will be relocated.

19. The NRC review concluded that: (1) special test exceptions 3.10.1 through 3.10.4 may be included with corresponding LCOs which are remaining in Technical Specifications, and (2) special test exception 3.10.5 may be relocated along with LCO 3.1.3.3. LCO 3.10.1 is only applicable in Mode 2. As discussed in Note 1 above, the SDM requirements for Modes 1 and 2 are retained in other Reactivity Control System Technical Specifications. Retained Special Test Exceptions 3.10.2.and 3.10.3 address Special Test Exception 3.10.1 for LCOs 3.1.3.1 and 3.1.3.6. Therefore, Technical Specification 3.10.1 will be deleted. Also, per the stated NRC conclusion, LCO 3.10.5 will be relocated. LCOs 3.10.2 through 3.10.4 will be retained as they are.

V

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-o m m m m.ncu e wAAA ut: AuvAuuuu . . . . . - - .n. m m 5.O.D. 4 T e m b

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t REACTOR COOLANT' SYSTEM' '

3.4.2.1 'T'Y UATY

'3.4.5 STEAMGENERATORS[dMerehina/)

^

3 .'4 . 7 :

A_. n CHEMISTRY ,

-3.4.9.2 PRESSURIZER P/T LIMITS

~3.4.'10.' STRUCTURAL INTEGRITY 3.4.11. REACTOR COOLANT SYSTEM VENTS

- EMERGENCY CORE COOLING SYSTEMS NONE i CONTAINMENT SYSTEMS ,

3.6.1.2 CONTAINMENT LEAKAGE 3.6.l'.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY-PLANT SYSTEMS 3.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3.7.8 SNUBBERS 3.7.9 SEALED SOURCE CONTAMINATION 3.7.12 AREA TEMPERATURE MONITORING' f ELECTRICAL POWER SYSTEMS 3.8.4.1 CONTAINMENT PENETRATION COND'UCTOR -

OVERCURRENT PROTECTIVE DEVICES i

e

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL ' SPECIFICATION 3.3.3.6 ACCIDENT MONITORING INSTRUMENTATION (APPLICABLE MODES; 1, 2, and 3)

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

The instrumentation that satisfies criterion 3 or 4 are the Type A variables in FSAR Appendix 7A as well as the risk-significant variables listed in the discussion below.

Some of the 3/4.3.3.6 instruments may be relocated and some must be retained. Neutron flux and RVLIS will be added.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

Tf tiie answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

9. Containment Hydrogen Concentration Level TS 3/4.6.4, Combustible Gas Control, which requires the operability of the containment hydrogen analyzers, is evaluated on the TS Screening Form for LCO 3.6.4.1. In accordance with that form and the Safety Evaluation, Attachment 1, LCO 3.6.4.1 will be deleted since it is redundant to LCO 3.3.3.6 and is obsolete per the STS. .!

10. Radiation Level in RCS ,

R.G. 1.97 defines the purpose of monitoring this variable as

detection of breach (of the fuel cladding). Our exception to installing instrumentation for this variable was approved -

in NRC's SER dated 4-10-85.

P

11. Auxiliary Feedwater Flow Rate AFW flow rate should be retained for several reasons:

1. AFW flow rate indication is modeled in the Callaway PRA for cueing operators to restore MFW if AFW is unavailable. Basic event AE-XHE-FO-MFWFLO has a RAW of '

1.10 (10% increase in CDF).

2. AFW flow rate indication is vital to the Heat Sink Critical Safety Function (CSF) status tree.

3. Operating experience has proven this indication to be important.

4. AFW flow rate indication is included in NUREG-1431 Table 3.3.3-1.

5. SR 4.7.1.2.1 requires AFW flow rate indication.

6. AFW flow rate indication is being retained in T/S Table 3.3-9 for the ASP. If an LCO and SR for the ASP AFW flow rate indication is being retained, it only makes sense to retain the MCB AFW flow rate indication.

v -

t

12. PORV and PORV Block Valve Position Indicator y[// Je r-iltethl.

I PORV and PORV block valve position indicators k '-- '-

dclutuu [ium Tuuhuicel Spuuific& tion 3.3.3.0. Lcas of

-pmo 4 tion indication requir^- eks- *k^ "ctienc acccciatcd -

with LCO 3.4.4 bu untered; thcr:forc, thcrc is ne uuud uv

-alee h2rc there indicctorc under LCO 2.2.2.E It is-further noted that these indicators are not Type A variables at l' Callaway nor are they RG 1.97 Category 1. Ibathly uhauucl-

- ch;cho for -hec: indicators kave bccn addcd uo SR 4.4.4.3

- ,, e e n 4 e_e -

L ,

w

)

l l

13. Safety Valve Position Indicator i

This instrument is not a Type A or Category 1 indication.

It is a Type D, Category 2 variable and will be relocated.

14. Unit Vent - High Range Noble Gas Monitor This instrument is not a Type A or Category 1 indication.

It is a Type D, Category 2 variable and will.be relocated.

(4) CONCLUSION

This Technical Specification is retained.

As indicated above; neutron flux and RVLIS to be added.

- The Technical Specification may be relocated to the following controlled document (s):

g e N-FSAR Chapter 16 (containment pressure-extended range, safety valve position indicator, .-ee+ unit vent-high range noble gas monito ).

d l

EP" cnd PO~' bicch u_ ___..a.._...s

, e._ cal cc,,,pccit___._.

. _ _ . . __,,n_

< Z _s__.

_s n indicatore

., ,.hncc

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p l0$ V Non Inlo'cobr A^l lfdfi A$ V lleglc YelV6

[##[ #M [#1 s cm he-TS3H N

4 F

i l

I

s

' TECHNICAL SPECIFICATION SCREENING FORM

-(1) TECHNICAL $PECIFICATION 3.3.3.10 EXPLOSIVE GAS MONITORING INSTRUMENTATION

[ APPLICABLE MODES: During Waste Gas Holdup System operation)

(2) EVALUATION'OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_X_ (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity.of a fission product barrier.

X (3) A structure, system, or component that is part'of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The explosive gas monitoring instrumentation provides the capability to detect the concentration of oxygen and hydrogen in the waste gas holdup system (at the hydrogen recombiners) and provide an alarm if the concentrations exceed prescribed limits.

According to LCO 3.3.3.10, this TS assures the operability of the

instrumentation required for LCO 3.11.2.5, Explosive Gas Mixture of the Radioactive Effluents TS. According to the Bases of LCO 3.11.2.5, the purpose of the limits on explosive gas I

concentrations and the monitoring instrumentation is to prevent an explosion in the waste gas holdup system. (The Bases for i 3.3.3.10 were deleted in OL Amendment No. 50). An explosion could result in a release of radioactive materials contained in the gaseous waste holdup system. Although release of the ,

contents of a waste gas decay tank is an analyzed DBA, the analysis assumes that the tank ruptures non-mechanistically and ;

not as the result of a hydrogen explosion. Therefore, the explosive gas limits are not an initial condition of a DBA.

The explosive gas monitoring instrumentation is not applicable to installed instrumentation used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The explosive gas monitoring instrumentation is not applicable to a process variable, design feature, or operating restriction that is an initial condition of any DBA or transient analysis. Thus, this TS does not satisfy criterion 2.

The explosive gas monitoring instrumentation is not_ assumed to function in the safety analysis. It is not a part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes.the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this TS does not satisfy criterion 3. ,

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA nor is it known to be significant based on risk insights from other PSAs or operating experience. ,

Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

_X_ The Technical Specification may be relocated to the following controlled document (s):

SAR Chapter 16. (The LCO y be relocated but a '

program etatcmenLj will be ded to the new TS Section 6.8.5).

de.reroy tWon TS3K

. . . - - - -y

)

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.3.4 TURBINE OVERSPEED PROTECTION r

[ APPLICABLE MODES; 1, 2, and 3] l (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO-

_X_ (1) Installed instrementation that is used to detect, and indicate in ~he control room, a significant abnorma: degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part'of ,

the primary success path and which functions or :

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_X_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The Turbine Overspeed Protection System actuates to mitigate a potential turbine overspeed event. This prevents the generation of potentially damaging miasiles from the turbine. The turbine i overspeed event is not a DBA. This event is evaluated to determine the probability of damage to equipment needed for safe shutdown. The turbine has a favorable orientation from the standpoint of low trajectory missiles; however, the combination

of overspeed probability with high trajectory strike probability i must meet the NRC's requirements for overall probability, i.e.,

less than 1E-7 per year. .

.The Turbine Overspeed Protection System is not applicable to installed instrumentation used to detect, and indicate in the e control room, a significant abnormal degradation of the RCPB; '

therefore, this TS does not satisfy criterion 1. .

The Turbine Overspeed Protection System is not associated with a !

process' variable, design feature, or operating restriction that is an initial condition of any DBA or transient analysis. Thus, this TS does not satisfy criterion 2.

The Turbine Overspeed Protection System is not assumed to function in the safety analysis. It does not apply to any SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this TS does not satisfy criterion 3. !

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA nor is it known to be significant based on risk insights from other PSAs or operating experience.

Therefore, this TS does not satisfy criterion 4.

I (4) CONCLUSION This Technical Specification is retained.

X The Technical Specification may be relocated to the

following controlled document (s):

N FSAR Chapeer 16. '*ke 100 may bc rciccatcd but a

-program ctatc;cnt will hc ?ddcd v new is 5eucivu .

-6.6.3).'

TS3L r

l

l TECHNICAL SPECIFICATION SCREENING FORM 1

(1) TECHNICAL SPECIFICATION 3.4.5 STEAM GENERATORS

[ APPLICABLE MODES; 1, 2, 3, and 4)

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:.

YES NO

_X_ (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_X_ (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product ,

barrier. >

X (3) A structure, system, or component that is part of ,

the primary success path and-which functions or '

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a' fission product barrier. ,

_X_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has -

shown to be significant to public health and 6 safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical i Specifications.

If the answer to all four of the above questions is "NO",-the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION .

This TS establishes the inservice inspection requirements for the steam generator (SG) tubes which are part of the RCPB. It is intended to maintain the structural integrity of this portion of the RCPB. The LCO requires the SGs to be operable in Modes 1, 2, ;

. 3, and 4; operability in this case refers to the structural l integrity of the SG tubes by means of an augmented inservice inspection (ISI) program that is performed periodically during '

plant outages.

l 1

l l

'This specification is not applicable to installed instrumentation l that is used to detect, and indicate in the control room, a ,

significant abnormal degradation of the RCPB; and, therefore, i this TS does not satisfy criterion 1.

This specification is not applicable to a process variable or 'I operating restriction that is an initial condition of a DBA or i transient analysis that either assumes the failure of or presents 'i a challenge to the integrity of a fission product barrier. The

. specification is applicable to the design feature of SG tube strength which comes into play, for example, during a LOCA or MSLB to avoid a combined LOCA/SGTR or MSLB/SGTR evel t. However, tube integrity is neither an active design feature ;r monitored or controlled during plant operation, rather durinc shutdown -

conditions under the SG ISI program. Thus, the str'ctural

integrity and assumed passive post-accident performance of the SG tubes is maintained by periodic inspection. Therefore, this TS does not satisfy criterion 2.

The SG tubes are components of the RCS that are part of the primary success path and which function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The post-accident or post-transient performance of the SGs, which is a passive function, is maintained by the periodic inspection and ,

repair of the SG tubes specified in this LCO. However, the :

operability of the SG tubes is not maintained during operation of the plant through any actions performed or parameters monitored by the operating staff. Also, the SG tubes do not perform any i active function or actuation required for DBA or transient ;

mitigation. Therefore, this TS does not satisfy criterion 3. !

The equipment associated with this TS was modeled in the Callaway Level 2 PSA fault trees; however, based on Appendix A, the F requirements of this TS are not of prime importance in limiting plant risk. Therefore, this TS does not satisfy criterion 4. .

v N v Ref. 4 concluded that this LCO could be relocated out of TS but '

that the SRs musy be retained. damar Man es/s e,-/a mar 7~f ans[ -

r*

l AddafNCLSIONSMy,$,r*m (4) C ,

deree,f%, -yb faehon Q .j,{y, TT wi// Je o-

_)(_ This Technical Specification is retained.

-M- The Technical Specification may be relocated to the following controlled document (s).

. -The LCO mm, bc rciccated te TSAA ChayLei 16, f

_ -hcucvcz, a 5G cubc amrvcill2nce progrnm ctatcrent-(

, ill be cdded to .; TO SmvLisa C.C.C.

A TS4J l

\

i TECHNICAL SPECIFICATION SCREENING FORM

-(1) TECHNICAL SPECIFICATION 3.4.7 CHEMISTRY ,

[ APPLICABLE MODES; At all times] ,

(2) EVALUATION OF POLICY STATEMENT CRITERIA '

Is the Technical Specification applicable to: '

YES NO

_X_ (1) Installed instrumentation that is used to detect, o and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_X_ (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ,

_X_ (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or .

presents a challenge to the integrity of a fission product barrier.

_X_ (4) A structure, system, or component which operating ;

experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the '

Technical Specification may be relocated to a controlled i document. '

(3) DISCUSSION t

This specification places limits on the oxygen, chloride, and l fluoride content of the RCS to minimize corrosion of the RCPB. 1 The RCS chemistry TS is not applicable to installed I instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCPB.

The RCS chemistry specification does not satisfy criterion 1.

. Chemistry restrictions are not used as initial conditions for safety analysis. However, the chemistry requirements are applicable, albeit indirectly, to a design feature (RCS integrity) that is an initial condition of a DBA or transient analysis that either assumes the failure or presents a challenge to the integrity of a fission product barrier. But RCS integrity 4 is a passive rather than an active design feature. Thus, the RCS

chemistry TS does not satisfy criterion 2.

The chemistry requirements for the RCS are applicable to the integrity of the RCS which is a system that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

However, the chemistry requirements do not directly assure the RCS integrity, but provide an indication of a concern. RCS integrity is assured through ISI and engineering evaluations of structural integrity. Therefore, the RCS chemistry TS does not satisfy criterion 3.

The Chemistry Limits governed by this TS, dissolved oxygen, chloride, and fluoride, have little relation to post-accident fission product species of concern (e.g. noble gases, iodine forms, cesium, tellurium, etc.). Refer to Tables 4.7.3-1 and 4.7.3-2 of the Callaway IPE and to the draft NRC source term NUREG-1465. As such, this TS does not satisfy criterion 4. ,

(4) CONCLUSION This Technical Specification is retained.

X The Technical Specification may be relocated to the following controlled anonmant (s) :

V FSAR Chapter 16. 'The LCO a&y be relec&ied Let &

prograr ctatcscat ill bu added uv new 15 5cc uivu -

C.C.C).

TS4M 1

i

TECHNICAL' SPECIFICATION SCREENING FORM I

(1) TECHNICAL SPECIFICATION 3.4.10 STRUCTURAL INTEGRITY

[ APPLICABLE MODES; All Modes)

(2) EVALUATION OF POLICY STATEKENT CRITERIA Is the Technical Specification applicable to:

YES NO

_X_ (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the. reactor coolant pressure boundary.

_X_ (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION This specification provides the inspection requirements for the ASME Code Class 1,2, and 3 components to ensure their structural.

integrity.

This specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a l

significant abnormal degradation of the RCPB. Therefore, the structural integrity requirements do not satisfy criterion 1.

This specification is not applicable to a process variable, design feature, or operating restriction that is an initial condition of DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. While the TS imposes an operating restriction regarding pressure boundary integrity, it is not monitored or controlled during plant operation. The assumed integrity of Class 1, 2, and 3 components is assured by means of periodic inspections. Therefore, this TS does not satisfy criterion 2.

ASME Code Class 1, 2, and 3 components are part of the primary success path and function to mitigate DBAs or transients that either assume the failure of or present a challenge to the integrity of a fission product barrier. Individual ASME Code Class 1, 2, and 3 components may satisfy criterion 3 and the requirements that ensure the integrity /open vility of these components are included in the individual specifications that cover these components. However, as stated above, this specification addresses the passive, pressure boundary function of these components. Therefore, this TS does not satisfy criterion 3.

Loss of component structural integrity is not modeled in the Callaway Level 2 PSA (internal events and flooding only). Our 6 IPEEE program for seismic and fire external events is currently underway. However, failure modes important to risk from an IPEEE would not be identified by this TS. Therefore, this TS does not satisfy criterion 4.

Ref. 4 concluded that the LCO for this specification could be l relocated out of TS; however, the associated SR must be relocated j to the TS programmatic requirements. l (4) CONCLUSION l This Technical Specification is retained.

X The Technical Specification may be relocated to the following controlled document (s):

AR Chapter 16. (The L may be relocated but a I program _ _ . will be dded to new TS Section I

6.8.5).

desce,y-/fon TS4R

TECHNICAL' SPECIFICATION SCREENING-FORM !

(1) TECHNICAL SPECIFICATION 3.6.1.2 CONTAINMENT LEAKAGE' l

[ Applicable Modes; 1, 2, 3, .and-4] l

?

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to: ;

YES NO X (1) Installed instrumentation that is used to detect, ,

and indicate in the control room, a significant ,

abnormal degradation of the reactor coolant pressure boundary, j X (2) A process variable, design feature or operating ,

restriction that is an initial condition of a-Design Basis Accident or Transient analysis that either assumes the failure of or presents a ,

challenge to the integrity of a fission product barrier. j

_X_ (3) A structure, system, or fomponent that is part of ,

the primary success path and which functions or 4 actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or !

presents a challenge to the integrity'of a fission product barrier.

X (4) A structure,. system, or' component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Spe'cification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is'"NO", the Technical Specification may be relocated to a controlled document.

(3) DISCOsSIOh!

This TS identifies the allowable leakage rates for the .;

containment structure which are established.to meet 10 CFR 50, Appendix J. These requirements ensure that the leakage-rates ,

from containment will not exceed the value assumed in the safety '

analyses at the peak accident pressure.

This specification is not applicable to installed instrumentation that is used to detect, and indicate in the control room, a ,

i

1

. I significant abnormal degradation of a the RCPB; and, therefore, J the TS does not satisfy criterion 1.

This specification is applicable to parameters that are an I initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity i of a fission product barrier. However, the process variables for ;

which the requirements are applicable (containment design. I pressure and allowable leakage rates) are not variables that are ,

monitored and controlled during power operation such that process t values remain within the analysis bounds. Containment integrity t is assured by periodic inspection and testing. Therefore, this !

specification does not satisfy criterion 2.

The specification applies to containment leakage rate limits.

Thus, it is applicable to a structure that is part of the primary success path and which function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the intent of i criterion 3 is to capture only those SSC (and supporting systems) :

that are part of the primary success path of a safety sequence analysis. Operability of the containment is assured by a separate LCO (3 . 6.1.1) , and the limits imposed by the leakage rate requirements are neither monitored or controlled during i operation nor part of the primary success path of the containment function. Therefore, this TS does not satisfy criterion 3.

Parameters with design limits such as SDM, MTC, rod drop time, AFD, Fg, FAH, quadrant power tilt ratio, DNBR, pressurizer and I SG pressure and temperature limits are chosen to preclude events '

from occurring that are non-mechanistically examined in FSAR <

Chapters 6 and 15. These parameters are not modelled in the PSA >

which is a best-estimate study of plant design vulnerabilities.

Relocation of this specification does not remove the requirement ,

to perform leak rate testing per 10CFR50 Appendix J. As such, these limits are not significant for criterion 4.

Ref. 4 concluded that this LCO could be relocated out of TS.but that the limiting values of Pa and La must be retained in TS. '

(4) CONCLUSION i

_ This Technical Specification is retained.

The Technical Specification may be relocated to the following controlled docum

The LCO may be relocated to FSAR Chapter 16 but the limiting values of Pa and La will be retained in the CONTAINMENT INTEGRITY Bases. Relocation of LCO 3.6.1.2 requires that revisions be de to SR 4.6.1.1.c and SR 4.6.1.7.2j ma b reh e*nce A na, 4 4,4, j,j.) w, fi la

**' # Y' S' I'I' N* a no} Ha,t ;, J as /oces$ !

N m Ap.4 E _

~ v

TECHNICAL SPECIFICATION SCREENING FORM ,

r t

(1) TECHNICAL SPECIFICATION 3.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY [ APPLICABLE MODES; 1, 2, 3, and 4] _i

! (2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

l YES NO

_X_ (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant. i pressure boundary.

X (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that '

either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or i actuates to mitigate a Design Basis Accident or i

~

Transient that either assumes the failure of or presents a challenge to the integrity of a fission :

product barrier. j,

_X_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has .

shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION l

The containment serves as a barrier to prevent the release of l fission products following a LOCA or MSLB inside containment. To ;

mitigate the potential consequences of a DBA, it is necessary )

that the containment structure meet its structural requirements. !

This specification is intended to detect abnormal degradation of ,

the containment structural elements. This TS outlines an appropriate inspection and testing program to demonstrate this i

capability. The program consists-of the measurement of tendon.

liftoff force, tensile tests of tendon wires, and visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment.

This specification is not applicable to installed instrumentation >

that is used to detect, and indicate in the control' room, a l significant abnormal degradation of a the RCPB;-and, therefore, l this TS does not satisfy criterion 1. l This specification is applicable to a design feature (the l containment) that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product-barrier. .

Containment structural integrity is assumed to'be available for many DBAs. .However, containment structural integrity is not monitored or controlled during plant operation but, rather, via periodic inspections and tests. Therefore, this specification does not satisfy criterion 2.

I The specification applies to the detection of abnormal degradation of containment structures and therefore to '

containment structural integrity. Thus, it is applicable to a structure that is part of the primary success path which functions to mitigate a DBA or transient that either assumes the ;

failure of or presents a challenge to the integrity of a fission ,

product barrier. However,.the functional mode addressed by the q TS is maintaining the passive, pressure boundary integrity. This 'l TS does not address the capability of the containment to function !

or actuate in order to mitigate the consequences of a DBA or transient. Therefore, this TS is not required to ensure the -

operability of containment and, thus, does not satisfy criterion 3.

Ref. 4 concluded that this LCO could be relocated out of TS but ,

that the associated SRs should be retained to meet the l operability requirements for a retained LCO, in this case LCO ;

3.6.1.1. Ref. 2 incorporated the SRs regarding tendon i surveillance into Section 6 of the TS. I i

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA. PRAs indicate that risk is dominated by events in which the containment is bypassed, unisolated, or fails ,

structurally. Containment failure frequency is determined by ,

comparing containment failure pressure. As discussed in the l Callaway IPE Section 4.4.2, the containment failure pressure is based on realistic material properties as well as conservative- ,

calculations of containment stress. The material properties used ]

in these calculations do not change rapidly, so the testing and inspection requirements of this technical. specification are not !

critical. Therefore, this TS does not satisfy criterion 4. ;

l l

(4) CONCLUSION This Technical Specification is retained. j

-i X The Technical' Specification may be relocated to the following controlled document (s):

~

FSAR Chapter 16. (The LCO'may be relocated but a. :

l~

program etatemenej will be added to new TS Section l

**5

de' seep.}r,,s T56F-

/Ven EA f J./, /. a wi// Je add'ad h inf/emed .

i N O f esp en m .

L s -

w ,

l 6

9

TECHNICAL SPECIFICATION SCREENING FORM j i

(1) TECHNICAL SPECIFICATION 3.7.8 SNUBBERS

[ APPLICABLE MODES; 1, 2, 3, and 4 - also 5 and 6 for i[

snubbers on systems required'to be operable in Modes 5 and 1 6.]

(2) EVALUATION OF POLICY STATEMENT CRITERIA t Is the Technical Specification applicable to:

YES NO

_X_ (1) Installed instrumentation that is used to detect, !

and indicate in the control room, a significant ~

abnornal degradation of the reactor coolant pressure boundary. .

_X_ (2) A process variable,. design feature or operating restriction that is an initial condition of a i Design Basis Accident or Transient analysis that either assumes the failure of or presents a t challenge to the integrity of a fission product l barrier. ;

_X_ (3) A structure, system, or component that is part of the primary success path and which functions or !

actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or !

presents a challenge to the integrity of a fission ;

~

product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has ,

shown to be significant to public health and safety. ,

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the i Technical Specification may be relocated to a controlled document. ,

(3) DISCUSSION The snubbers are required to be operable to ensure that the ,

structural integrity of the RCS and all other safety-related systems is maintained during and following a seismic or other i event initiating dynamic loads. The restraining action of the !

snubbers ensures that the initiating event failure does not .

l i

propagate to other parts of the failed system or to other safety systems. Snubbers also allow normal thermal expansion of piping )

and nozzles tc eliminate excessive thermal stresses during heatup or cooldown. Snubber surveillance is conducted under the requirements of the Snubber Surveillance Program at Callaway.

The TS requiremente for snubbers are not applicable to installed instrumentation used to datect a significant abnornal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The snubber TS is associated with a design feature or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

However, the snubber requirements are not explicitly considered in the accident analysis. The availability of the snubbers is assumed based on the performance of a program of periodic augmented inspection and testing. Snubber operability is not required to be monitored and controlled during plant operation.

Some snubbers (inaccessib'e) can only be inspected during plant outages. Thus, this TS does not satisfy criterion 2.

Those snubbers that are required to function during DBAs or transients to prevent the initiating event from propagating to other systems or components that are part of the primary success path may be considered components that are part of the primary success path and which function or actuate to mitigate a DBA or transient that aither assumes the failure of or presents a challenge to the integricy of a fission product barrier.

However, snubbers are not explicitly considered in DBA or transient analyses but are a structural / design feature whose operability ir, assured by an inspection program. Therefore, the snubber requiraments do not satisfy criterion 3.

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA nor is it known to be significant based on risk insights from other PSAs or operating experience.

Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

X The Technical Specification may be relocated to the f ollowing controlled document (s) :

7 v -N FSAR Chapter 16. (Thc LCO may bc rciccatcd but a

-progr=~ ctatcacnt cill bc addud te ncw TO Ccction

_E.C.C' l

1 TS7M I

e' TECHNICAL SPECIFICATION SCREENING' FORM

-- /

(1) TECHNICAL SPECIFICATION 3.7.12 AREA TEMPERATURE MONITORING

[AFPLICABLE MODES; Whenever equipment _in area is required to be operable.)

(2) EVALUATION OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO

_X_ (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant cbnormal degradation of the reactor coolant p ressure boundary.

X (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or-Transient that either assumes the failure of or-presents a challenge to the integrity of a fission product barrier.

_X_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION This specification places a limit on the temperature of the areas l of the plant which contain safety-related equipment. This is '

required to ensure that the temperature of the equipment does not j exceed its environmental qualification temperature during normal operation. Exposure to excessively high temperatures may degrade the equipment and cause a loss of its operability.

I i

i The TS. requirements for area temperature monitoring are not -

applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1. ;

The area temperature monitoring TS is associated with the ;

variable-of room temperature which is not a process variable, design feature, or operating restriction that is an initial-condition of a DBA or transient analysis that either assumec the ;

failure of or presents a challenge to the integrity of a fission {

product barrier. Thus, this TS does not satisfy criterion 2. ,

The TS for area temperature monitoring does apply to the operability of SSCs that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the TS is only indirectly applicable to the operability of these systems and components. Therefore, this TS does not satisfy criterion 3.

Although room.heatup calculations were reviewed during the Callaway IPE to determine equipment survivability, the normal operation limits governed by this TS have only a secondary relationship to post-accident and'off-normal room temperatures ,

and no relation to the EQ test data used to determine equipment [

functionality. In general, room coolers were determined to be ;

risk significant; however, initial room conditions are not overly -

important. As such, this TS does not satisfy criterion 4.

e (4) CONCLUSION i This Technical Specification is retained.

X The Technical Specification may be relocated to the following controlled document (s):

FSAR Chapter 16. 'The LCC m2y bc rcloceted but a-

-progrnr rt2te nt .111 'ee mddmd Lv om. T3 Seccion- ,.

_f Q Ci TS70 f

P

TECHNICAL SPECIFICATION SCREENING FORM (1) TECHNICAL SPECIFICATION 3.11.1.4 LIOUID HOLDUP TANKS )

(APPLICABLE MODES; At all times.)

(2) EVALUATION OF POLICY STATEMENT: CRITERIA Is the Technical-Specification applicable to:

YES NO

_Z_ (1) Installed instrumentation that is used to detect, i and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_X_ (2) A process variable, design feature or operating l

restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_X_ (3) A structure, system, or component that is part of the primary success path and which functions or 1 actuates to mitigate a Design Basis Accident or j Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_X_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION

, The liquid holdup tank specifications impose limits on the l

quantity of radioactive material contained in specific outdoor tanks that may contain radwaste. Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentration would be less than the limits'Of 10 CFR 20, Appendix B, Table II, Column 2, at the

4

! l nearest potable' water supply and the nearest surface water supply in an unrestricted area. The tanks addressed by this specification are:

a. Reactor Makeup Water Storage' Tank

b. Refueling Water Storage Tank'

c. Condensate Storage. Tank-

d. Outside temporary tanks, excluding demineralizer vessels'and liners being used to solidify radioactive wastes.

These tanks are not addressed by the safety analysis of radioactive release from a subsystem or component.

The TS requirements for liquid holdup tanks are not applicaole to.

installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The liquid holdup tanks TS is not associated with a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thus, this TS does not satisfy.

criterion 2.

The TS for liquid holdup tanks does not apply to an SSC that is part of the primary success path ~and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Thereforc' this TS does not satisfy criterion 3.

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA nor is it known to be :,ignificant based on risk insights from other PSAs or operating experience. Accidents evaluated in FSAR Section 15.7 (other than FHA) result in insignificant offsite dose consequences when compared either to the design basis LBLOCA or to the beyond design basis scenarios examined in the Callaway IPE (e.g. scenarios with corium released from a breached reactor vessel, etc.). Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION This Technical Specification is retained.

_X_ The Technical Specification may be relocated to the f ollowing controlled document (s) :

SAR Chapter 16. (The LCO m v be relocated but a program stat;;;uy will be addyd to new TS Section

'$>^

Tsi A df^ M#6 1

- . . __ a

q V:

TECHNICAL SPECIFICATION SCREENING. FORM

'(1) TECENICAL SPECIFICATION 3.11.2.5 EXPLOSIVE GAS MIXTURE

[ APPLICABLE MODES; At all times]

.(2) EVALUATION'OF POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:

YES NO <

X (1) Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

_X_ (2) A process variable, design feature or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

X (3) A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a Design Basis Accident cnr

Transient that either assumes the failure of or presents a challenge to the integrity of a fission '

product barrier.

_X_ (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the tbove questions is "YES", then the Technical Specification shall be retained in the Technical Specifications. ;

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

)

(3) DISCUSSION This specification is provided to ensure that the concentration i of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of ;

hydrogen and oxygen. Maintaining these limits provides assurance i that the releases conformance of requirements with the radioactive materials of will be controlled in 1 GDC 60 of Appendix A to 10 CFR 50. The accident analysis concerning the gaseous radwaste ;

system' assumes that a storage tank ruptures, from unspecified causes, and releases its contents without mitigation.

-The TS requirements for explosive-gas mixture are not applicable to installed-instrumentation used to detect a significant abnor'al n degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The explosive gas mixture TS is not associated with a process- '

variable, design feature, or operating restriction that is.an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity i of a fission product barrier. Thus, this-TS does not satisfy criterion 2.

The TS for explosive gas mixture does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA nor is it known to be significant based on risk insights from other PSAs or operating experience. . Accidents evaluated in FSAR Section 15.7 (other than FRA) result in insignificant offsite dose consequences when compared either to i the design basis LBLOCA or to the beyond design basis scenarios .

examined in the Callaway IPE (e.g. scenarios with corium released ;

from a breached reactor vessel, etc.). Therefore, this TS does - i not satisfy criterion 4. !

1 (4)

CONCLUSION This Technical Specification is retained.

_X_ The Technical Specification may be relocated to the ;

following controlled document (s) :

FSAR Chapter 16. (The may be relocated but a program-ct2te entjwill be a'ded to new TS Section 6.8.5).

A 4 1 TSilB i b

,,- ,m, --

, r-- -+

l TECHNICAL SPECIFICATION SCREENING FORM l

(1) TECHNICAL SPECIFICATION 3.11.2.6 GAS STORAGE TANKS

.[ APPLICABLE MODES; At all times.)

(2) EVALUATION OF POLICY STATEMENT CRITERIA !

Is the Technical Specification applicable t o:

YES NO

_X_ (1) Installedinstrumentationthatisusedtohetect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

X (2) A process variable, design feature or operating restriction that is an initial condition of a !

Design Basis Accident or Transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

_X_ (3) A structure, system, or component that is part of ,

the primary success path and which functions or ;

actuates to mitigate a Design Basis Accident or ;

Transient that either assumes the failure of or '

presents a challenge to the integrity of a fission product barrier.

X (4) A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

If the answer to any one of the above questions is "YES", then the Technical Specification shall be retained in the Technical Specifications.

If the answer to all four of the above questions is "NO", the Technical Specification may be relocated to a controlled document.

(3) DISCUSSION The gas storage tank specifications impose limits on the quantity of radioactive material contained in those tanks for which the quantity of radioactivity contained is not limited directly or i indirectly by another TS. Restricting the quantity of radioactivity contained in each gas storage tank provides '

assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a member of I

I l

__J

the public at the nearest site boundary will not exceed 0.5 rem.

This is consistent with Branch Technical Position ETSB 11-5,

" Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure." The accident analysis concerning the gaseous radwaste system assumes a rupture of a storage tank without mitigation.

The TS requirements for gas storage tanks are not applicable to installed instrumentation used to detect a significant abnormal degradation of the RCPB; therefore, this TS does not satisfy criterion 1.

The gas storage tank TS is associated with a process variable or operating restriction (quantity of contained radioactivity) that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the barrier in this case is the tank itself which is not a barrier that is monitored and controlled during power operation of the plant.

Therefore, this TS does not satisfy criterion 2.

The TS for gas storage tanks does not apply to an SSC that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this TS does not satisfy criterion 3.

The equipment associated with this TS was not modeled in the Callaway Level 2 PSA nor is it known to be significant based on risk insights from other PSAs or operating experience. Accidents evaluated in FSAR Section 15.7 (other than FEA) result in insignificant offsite dose consequences when compared either to the design basis LBLOCA or to the beyond design basis scenarios examined in the Callaway IPE (e.g. scenarios with corium released from a breached reactor vessel, etc.). Therefore, this TS does not satisfy criterion 4.

(4) CONCLUSION 1

This Technical Specification is retained.

X The Technical Specification may be relocated to the ,

following controlled document (s): '

er 16. (The O may be relocated but a program etatement will be added to new TS Section z

ke.te ry}4n TS11C l

l Al

ML20087G382

Text

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