ML20197B669

Proposed Tech Specs Pages Re 970723 TS Amend Request Associated W/Pressure Temperature Limits Rept

ML20197B669
Person / Time
Site:	Farley
Issue date:	12/18/1997
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20197B622	List: ML20197B617 ML20197B639 ML20197B647 ML20197B669
References
NUDOCS 9712240019
Download: ML20197B669 (84)

Text

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Joseph M. Farley Nuclear Plant Technical Specification Change Pages i

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Pressure Temperature Limits Report g.a h

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Joseph M. Farley Nuclear Plant Marked Up Technical Specification Change Pages Associated with the Pressure Temperature Limits Report 1

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DEFINITIONS

$fCTION

ggf, 1.0 OEFINITIONS 1.1 ACTION.........................................................

11 1.2 AXIAL FLUX DIFFERENCE..........................................

11 1.3 CHANNEL CALIBRATION................................

11 1.4 CHANNEL CHECK.................................................. 11 1.5 CHANNEL FUNr, TION TEST.......................................... 11 1.6 CONTAINMENT INTEGRITY.........................................

12 1.7 C ONTROL L ED L EAKAG E..........................................

1.8 CORE ALTERATION............................................... 12 1.9 00SE EQU! VALENT I 131......................................... 12 1.10 E AVERAGE O!S INTEGRATION ENERGY........................

13 1.11 ENGINEERED SAFETY FEATURES RESPONSE T!ME....................... 13 1.12 FREQUENCY NOTATION.............................................

13 1.13 GASE9WS.RA9 WASTE TREATNENT. SYSTEM (Deleted)....................

13 1.14 IDENTIFIED LEAXAGE............................................

13 1.15 LIGWIO.RA9 WASTE TRF.ATMENT SYSTEM (Deleted)..................... 14 1.16 MAJOR CHANGE 5 TO RA010 ACTIVE WASTE-TREATMENT SYSTEMS 14 0FFSITE DOSE CALCULATION MANUAL (00CM).........

1.17 14 1.18 OPERABLE OP:RA81LITY.........................................

14 1.19 OPERATIONAL MODE MODE........................................

15 1.20 PHYSICS TESTS............................................

15 Nd 9 21 PRES $URE 8OUNDARY LEAXAGE......................................

15' 1

22 PROC ESS CONTROL PROGRAM ( PC P)............................... 15 1.23 PURGE PURGING................................................

15 1.24 QUADRANT POWER TILT RATIO...................................... 15 1.25 RAT E D TH ERMAL POW ER............................................

1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME.........................

14 1.27 REPORTA8LE EVENT...............................................

16

1. 21s SHUTDOWA MARGIN...............................................

16 1.29 50k191FICAT10N (Deleted SOURCE CHECK..........).............................

16 1.30 16 1.31 STAGGERED TEST RAS!$.............................

16 1.32 THERMAL POWER.................................................,

17 1.33 _UNIDfMTIFIED LEAKAGE.......................................~....

17 1.34 VEr:1 '.!!'JN EXHAUST TREATMENT SY STEM...........................17 1.35 VENTING........................................................

17 TABLE 1.1 OPERATIONAL MODES.............

18 TABLE 1.2 FREQUENCY NOTATION........................................

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'7 FARLEY UNIT 1 AMEN 0MOff No. 57 @

i IEE ADMIN!!TRATIVE C0Kfacts 1

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Asvies.....................................................

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Aandits.....................................................

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Authority..................................................

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Records....................................................

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6.5.3 ft00l! CAL REV! W A M CoffROL l

Activities.................................................

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Records....................................................

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E. 8 Rf9ETAKI EVWT ACT IM......................................

6 14 8.7 SAFffY LIE T VIK Aft s.......................................

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6. 3 PEKIERES AM pumass......................................

6-14 8.9 Mp mTIM M allE M N 1 4:

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$tertop OEpert.............................................

6*lla Ammea l Amport..............................................

6 16 l

Amasal Radielegical Enviremmastal operaties tapert........

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Ammeal Radienstive Effl eest tel ease Report.................

6 17 l

Rest 41y Operettog asport...................................

6-19 Redial Peaking Facter Listt Aspert.........................

6 19 Ammeel Memel temeratar heltaM11tLy Data taport............

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Annant toestar Caelast System specific Activity Amport.....

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OPERAff0NAL MODE M00[

1.19 An CPERATIONAl. MCDE (i.e.. H00E) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature spectfied in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests perfomed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR. 2) authorized under the provisions of 10 CFR 50.59 or 3) otherwise approved by the Commission.

PRESSURE BOUN0aRY LEaO GE 1.21 PRES $URE BOUNDARY t.EAKAGE shall be leakage (except steam generator tube leakage) through a non isolable fault in a Reactor Coolant Systee Component body pipe wall or vessel wall.

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3g,4f, PROCESS CONTROL PROGRAM f PCP) 1.22 The PROCESS CONTROL PROGPM (PCP) shall contain the current fomulas.

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sampling, analyses, tests, and deterstnations to be made to ensure that

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processing and packaging of solid radioactive wastes based on demonstrated processing of actual or staulated wet solid wastes will be accomplished in such a way 4: to assure compliance with 10 CFR Parts 20, 61, and 71; State

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regulations burial ground requirements: and other requirements governing ths disposal of solid radioactive waste.

U PURGE PURGIMG 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition in such a sanner that replacement tir or gas is required to purtfy the confinement.

00ADPJuff POWER TILT RATIO 1.24 QUADRANT POWER T!1.T RATIO shall be the ratto of the maximum upper excore detector calibrated output to the average of the upper excore detector calt-brated outputs, or the ratto of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable the remaining three detectors shall be used for computing the average.

AMENOMEXT No. 26h FARLEY UNIT 1 15

FARLEY NUCLEAR PLANT - UNIT l PTLR SUBMITTAL TECHNICAL SPECIFICATIONS MARKUPS INSERT 3 PRESSURti TEMPERATURE LihilTS REPORT IPTLR) 1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR)is the unit specific document that provides the reactor vessel pressure and temperature (P/D limits, including heatup and -

- cooldown rates, for the current reactor vessel fluence period. These P/T limits shall be determined for each fluence period or effective full power years (EFPYs) in accordance with Specificatien 6.9.1.15. Plant operation within these operating limits is addressed in LCO 3.4.101, RCS Pressure / Temperature Limits.

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REACTOR COOLANT SYSTEN 3/4.4.10 PRESS 11RE/TEMPERAftfRE LIMITS REACTOR COOLANT SYSTEM LIMITT ".flt.lFic-o alW B'eWll24.

s "TeMFb:f.4TclE4, Lt AirT5 WT(PTL LIMITING CONDITION FOR OPERATION

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3.4.10.1 The Reactor Coolant Systein (except the ressuriser) temperature and pressure shall be limited in accordance v4.th the - --

3.^ 2 ::d 3.5 3 during heatup, cooldown, etiticality, and inservice leak and hydrostatic testing."i W

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'-"m4eatup-of--LOO *-F in-any-one -hour 9er4+d-a.

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leum-cooldown-of-100* F-in--any--one4eur - F r i ed -

444aximum-temperature-change-of less-than-or-equal-te -10* ' ! a-say.one hour--period-during-inservice hydros tatic--andJeak-tasting.operutons abee; the-heatep-end-eeeldevr. li it rc;;.

APPLICABILITY: At all times.

ACTION:

.JP4///~/4-B /A/Th16-I' Vith any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutest perform an engineering evaluation or inspection to determine the effects of the out-of-limit condition on the fracture toughnees of the Reactor Pressure Vessel determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY ytthin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200 F and 300 psig, respectively, within the foll8ving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SlfRVEILLANCE RE0llIREMENTS gggjAf 74 F/7.C 4.4.10.1.1 The Reactor Coolant Sy tem temperature and pressure shall be determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.

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FO,R THE SERVICE PERIOD UP TO 16 EFPY

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INDICATED TEMPERATURE (O F)

FIGURE 4-2 FARLEY UNIT 1 RCACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLIC LE FOR THE FIRST 16 EFPY FARLEY - UNIT 1 3/a 4 29 AttENDMENT NO. 58 p

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o do too iso roo tso soo 4o0 40 soo INDICATED TEMPERATURE (O F F aVRE 3.4 3 FARLEY UNIT 1 REACTOR COOLANT SYSTEM CCOLCOWN L.!TATIONS APPLICABI.E FOR THE FIRST 16 EFPY FARLEY UNIT 1 3/4 4-30 AMENOMENT NO.58 y e

RE ACTOR COOL ANT $0$?$M 845E5 1.

Reducin' T g av to less than 500*F prevents tr.:

• steam genera!or tube rupture since the saturation pressure of tne primaryr coolant is below the lif t pressure of the atmospneric steam relief valves.

The survelliance requirements provide adequate assurance that excessive spe:tf te activity levels in the primary coolant will be detected in suf f tetent time to t

take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with spiting phenomena.

A reduction 16 frequency of isotopic analyses followlag power changes may be permissible if justified by the data obtained.

3/4.4.10 PRESSURE / TEMPERATURE t.!MITi The temperature and pressure changes during heatup and cooldown are limited to y/

be consistent with th quirements given in the ASPT Boiler and Pressure Yessel Code. Sectiog

,' Ap din G as required per 10 CFR Part 50 Appendt G.

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1) The reactor coolant te ure and prissure and 5 sten heatup and coolcown rates (with the exception of the pressurizer) t accordance with re 7 3.4 3 =: 3,e _;i f45fdtd TeMR-Et.Tild6-a) Allowable comoinations of pressure and tm:r.L,(Uh.TsGFbKr(

f:r :o temnerature change rates are telow and to the -ight of tne limit lines showr Limit lines for cooldown rates between those presented may be p* g obtahedb (ntereolation.

r__ :.=e Pic)3.' ?-define 31tmits to assure prevention of u

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nonductile failure onl For nomal operation, other inherent plant characteristics, e.g.,y. pump heat addition and pressurizer heater a

capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2) These limit linat chall be entrulaw y-Wv uninn - tan ds_ r - ' t #/,

"' - @ acw NWA/4 A shyt.smWz, MermarxMnso 70 a-& C SG - %

• M/7%t71Ab 3dreMJtTRrA05DD RJhht7Up4D MMW L/N _

3) e Ebnda

% vr ihrFaii pntPTt5eYstWbe pressuruno above 200~.

psig if the temperature of the steam generator is below 70*F.

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sve F art,CY. UNIT 1 0 3/4 4 6 AMEND S T NO. 2 h

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REACTOR COOLANT SYSTEM BASES 4)

The pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200'F/hr respectively. The spray shall not be used if the temperature dif ference between the pressurizer and the spray fluid is greater than 320'F.

5)

System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code.Section XI.

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WC AM^:' A. "" t4t-Or "se twp--and-Gcci d;ur l'-i t Curv :,

'pr 1075."

2 /.9 Heatup and cooldown limit curves are calculated u' g the most 11 ting alue of the nil-ductility reference temperature. RTn

, at the end of e f fective full pouer years (EFPY) of service life. The EFPY service life period is chosen such that the limiting Rindt at the 1/4T location in the core region is greater than the Rindt of the limiting unirradiated material.

The selection of such a limiting Rindt assures that all components in.the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

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P n r may be used until the next l

Values of RT,$e, detern,ined in t.!: material surveillance program, evaluated accordinE to i

results from t ASTM E185-82, are available. Capsules vill be removed in accordance with the requirements of ASTM E185-82 and 10 CTR $0, Appendix 8.

The surveillance specimen withdraval schedule is shown in ?!'" 5:::i:: 3*

The heatup and cocidown curves must be recalculated when the MT determined from the surveillance capsule exceeds the calculated k for the eq 1ent capsule radiation exposure.

P12.2 ;.

A11ovable sure-temperature relationships for various heatup and-cooldovni gates are calculated using methods derived from Appendix G in Section iTT of the ASME Soiler and Pressure Vessel Code as required by Appendix G to 1 r ""

d n n -thods are discussed in detail in

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CAP.

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NI Indt Rindt Upper

• I Energy Component Code No.

Typt (1)

(1)

(1)

(*F)

(*r)

[c]

umu[d]

~

Closure head done 96 A533,B C1.1 0.16 0.009 0.50

-30

-20[a]

g40 Closure head. segment B6902-1 A533,B.C1.1 0.17 0.007 0.52

-20

-20[8 ]

138 Closure head flange B6915-1

. C1.2 0.10 0.012 0.64 t ]

60[8]

75[8]

- Vessel. flange 86913-1 A5 C1.2 0.17 0.011 0.69 60[a]

60[8]

106[a3 Inlet nozzle 86917-1

A508,

.2 0.010 0.

60[*]

60[8]

110 Inlet nozzle' B6917-2 A508, C1.

'O.008

.80 60[8]

60[aj go o,

Intet nozzle 86917-3

'A508 C1.2 O.

0.87

'60[8]

60[*]

98

. %.5 Outlet nozzle B6516-1 A508 C1.2 0

1 0.77 60[8]

60[a]

2 Outlet nozzle B6916-2 A508 C1.2

.011 0.78 60[8]

60[8]

'97.5 o

l

. Outlet nozzle B6916-3 A508 C1.2

.009 0.78 60[8]

60[8]

100 Nozzle shell 86914-1 A508 C1.2 0.

0 0.68 30 30[a]

148 I.nter. shell 86903-2 A533,B,C1.1 0.13 0.01 0.60 0

0 151.5 97 Inter, shell 86903-3 A533,8, 0.12 0.014

.56 10 10 134.5 100 1.ower shell B6919-1 AS3} C1.1 0.14

.0.015 0.

-20 15 133 90.5 tower shell B6919-2 F33.B.C1.1 0.14 0.015 0.56

-10 5

134 97 Bottom head ring.

B6912-1 A508 C1.2 0.010 0.72 0

10[*]

163.5 Bottom head segment 86 A533,8,C1.1 0.15 0.011 0.52'

-3

-30[8]

147 Bottom head dome 7-1 A533,8,C1.1 0.17 0.014 0.60

-30

-30[8]

143.5 E Inter, shell long.

M1.33 Sub Arc Weld 0.25 0.017 0.21 0[8]

[8]

G weld ' seam

• Y Inter..to lowe G1.18 Sub Arc Weld 0.22 0.011 (0.20[b] 0[8]

0[8]

G shell weld ans:

tower s I long.

G1.08 Sub Arc Weld 0.17 0.022

<0.20[b] 0[8]

0[8]

5 weld seams-

[a] Estimate per NUREG-0800 "USNRC Standard Review Plan" Branch Technical Position HiEB 5-2.

~

[b] Estimated (10w nickel weld wire used in fabricating vessel weld seams).

4, t

[c] Major working direction.

[d] Normal to major working direction.

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":00:10 0 ;/4.4-2 74:7 NCUT10 : PLU:N: (: ' " '!! " !~ ' A:

FUNCT: N CI TULL ?Okt?. : ?,V!CE (CI??)

FA R EY - UNIT 1 B 3/4 4-10 A AMENDMENT N

REACTOR COOLANT SYSTEM i

• If4M PAgTkdTicsJt^1LY M WM i

SASES V

The ASME approach for calculating the allowable Ilmit curves for var us he up and cooldown rates specifies that the total stress intensity fa or, K,

e the combined thermal and pressure stresses at any time durin heatup l

g or cool wn cannot be greater than the reference stress intensity actor, K gg, j

for the al temperature at that time.

K is oMabed from t nfennce j

gg fracture to hness curve, defined in Appendix G to the ASME C e.

The K gg curve is given y the equation:

Kgg = 26.78 +.223 exp (0.0145(T-RTNOT + 160 D (U

s Were K is the refere e stress intensity facto as a function of the metal 4

gg tempurature T and the met nil ductility refer ce temperature RT

Thus, NDT.

the governing equation for t e heatup-cooldo analysis is defined in Appendix G-of the ASME Code as follows:

CKgg + KggiK, (2) g Where K is the ctress'intensi fact caused by membrane (pressure) gg stress.

/

/

K is the stress jntensity factor caus by the themal gradients.

it

/

is provide [by the code as a function of temperature relative K gg to the Ri of the material.

g C = 2.

for level A and B service limits, and C=

.5 for inservice hydrostatic and leak test operat ns.

A any time during the heautp or cooldown transient, Kgg is etermined by the spetal temperature at the tip of the postulated flaw, the approp ate v ue for RTNOT and the reference fractur' toughness curve.

The the al tresses resulting from temperature gr,sents through the vessel wall are FARLEY-UNIT 1 8 3/4 4-11 AMENDMENT NO

REACTOR COOLANT SYSTEM 8ASES L%

alculated sad then the corresponding thermal stress intensity factor, K fo the reference. flaw is computed.

From Equation (2) the pressure str ss inten ity factors are obtained and from these, the allowable pressur are calcula d.

COOLDOWN For the cal latien of the allowable pressure versus oolant temperature during cooldown, t code reference flaw is assumed to e st at the inside of the vessel wall. Du ng cooldown, the controlling loc ion of the flaw is always at the inside o the wall because the thermal

adients produce tensile stresses at the inside, hich increase with increas.ig cooldown rates.

Allowable pressure-temperature rela ons are generated for th steady-state and finite cooldown rate situations.

om these relations omposite limit curves are constructed for each cooldown ate of interest The use of the composite cu in th cooldown analysis is necessary because control of the coolco.n proc u is t'ased on measurement of reactor coolant temperature, whereas the limi pressure is actually dependent on the material temperature at the t of th assumed flaw. During cooldown, the 1/4T vessel location is at a hi her temperat e than the fluid adjacent to the vessel 10. This condition.

~ course, is not ue for the steady-state situation.

It follows that at any giv reactor coolant temp (ature, the delta T developed during cooldown results a' higher value of K at W e 1/4T location for IR finite cooldown rates an for steady-state operation.

Furthermore, if conditions exist su that the increase in K exceeds K the calculated IR gg allowable pressur during cooldown will be greater than the teady-state value.

The ove procedures are needed because there is no direct cont ol on temperaj re at the 1/4T location; therefore, allowable pressures may unkno ingly be violated if the rate of cooling is decreased at various inte als along a cooldown ramp. The use of the composite curve eliminat th problem and assures conservative operation of the system for the enti c oldown period.

AMEN 0MENTNO.h FARLEY-UNIT 1 B 3/4 4-12

REACTOR COOLANT C'r5 TEM MI@h N1M 8ASES

~

REATUP hree separate calculations are required to detemine the limit curve for fin (te heatup rates.

As is done in the cooldown analysis, allowable pressure'teeperature relationships are developed for steady state con tions

/4T as well as 'f(nite heatup rate conditions assuming the presence of a defect at the'inside of the vessel wall. The thermal gradients ring heatup produce compresst(e stresses at the inside of the wall that ap viste the tensile strasses pPoduced by internal pressure. The metal reture at the crack tip legs the c ant temperature; therefore, the K for the 1/4T crack during heatup is lower an the E for the 1/4T crack uring steady-state gg conditions at the same coe nt temperature. During atup, especially at the end of the transient, condit ns may exist such t t the effects of coegressive thennal stresses and different K g's for steady state and finite heatup rat'es do.not offset each other and the assure-teeperature curve based on steady-state conditions no longer represent a1 r bound of all stellar curves for finite heatup rates when trae 1/4T flew ( considered. Therefore, both cases have to be analyzed in order to assu t at any coolant temperature the lower yalue of the allowable pross ire ealcu ted for stea# state and finito heatup rates is obtained.

/

The second portion of the heatup analysis co ras the calculation of pressure temperature 11mit4tions for the case in whi a 1/4T deep outside surface flew is assumed snel inside surface, the thenmal gradientsp/ Unlike the situation at the stablished at the outside surfac during heette produce stresses which are t6hsile in nature and thus tend to ret orce.any pressure stresses present.

hose thereal stresses, of course, are ndent on both the rate of heat and the time (or coolant temperature) ale the heatte ramp.

Furt.he

, since the thereal stresse, at the outside re tens 11e and increate with Ancreasing heatus rate, a lower bound curve canne be defined.

Rather, eac weatup rate of interest must be analyzed on an indiv 1 basis.

Fo1 ing the generation of pressure-temperature curves for be the steady-te and finite heatup rate situations, the final limit cur'tes re produ d as follows. A composite curve is constructed based on a point-poi comparison of the steady state and finite heatup rate data. At any gi en temperature, the allowable pressure is taken to be the lesser of the ree values taken from the curves under consideration.

s FARLEY-UNIT 1 a 3/4 4-13 M DMENT NO 89

REACTORC00LAITSYiMg i

aASEs I

L The :: ef t'= tr;::!te '."ree is 9e *ssary te set cen"r 't!": h:*tep

! M!tet!-

3-M!e it-ta-

-:: r :0 Of t'- '" *"; "

poss4hle-fer rend!!! ens-te e:464 t he -coat rell 'ac teadi t ie" !"i tch"'

f"^" t'^ '*"!de t^

th: ::t:!d: and t'^

j th: :::t e44t:"1 r;?trica.r-Stre '! sit-aust-at-411-ti-" he b""ed : * "ly:i" ^f l

t F4*elly, th: !? CFRJet 50,- Append 44-4 Aula "ki'k ' d d -- " * - ' ' - - * '

t:r;r:tr: :! the 04e:r: 5:ad 414nge ted vesse! '!!N0e -"St '. 2 ne:!dr:d.

Th+s-A#1*-+444ee-tht t50-eteleum-se641--t:r;reture f-t5: :Jeure-flange 4

Pet 4#" 50 et !^^tt !!0er gg;gey gg3, gg, i!,tt!.; or,m so. tk... 7.; ten egg th: ;r000r: :=d 30 ;reent-of -the pre:==" tee-hydr 464 tte tett-pr:sevre-(641 -;;!Wfr forl'y ""(4-1),-In-addItien -the-new-le-CEA Ser* 50 N!: et t=

r th44-a -p' ant-spec 6f4c-fracture-evaluat4on-say-be-perfomed te j=tify !=:

1i ttir.; :;;t7: xte.

i

".: : esuttr-such-44racture-analyds-wu ;rf:.md fu Fr? y W it 3.

" c = F r!:y Unit-3 fracture-analy444-resulte re

--u - '-

g g

Feeley Wit !

!Ne^ the ;rt! ret per--*ars treJd'^tice! fer 5^th O! =*.

h=d :;r th!: fr*:te: :=Lyster-the-14-UM-heatup-4M entd:r. :== r:

b t ;nt:d b" th: := 10 CFMet 50-Aule-a: th:r = '?; J n 3.1 2 =d 3.?-3.

i

=

j Although the pressurizer operates in temperature ranges above those for.wht W i

there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis perforised in accordance with the ASME Code requirements.

i I

The OPERA 8!LITY of either RNR reitef valve or an RC5 vent opening of greater than or equal to 2.45 aguare inches ensures that the RC5 will be protected fros ii pressure transients which could exceed the limits of Appendix G to 10 CFR Part (

50 when one or more of the RC5 cold legs are less than er equal to 310*F.

I Either RHR reitef valve has adequate relieving capability to protect the RCS fk(

)

free overpressurization when the transient is limited to either (1) the start

/

(

of an idle RCp with the secondary water temperature of the steam generator less l

than or equal to 50*F above the RCS cold leg temperatures provided measures are

(

(

taken to cushion the overpressure effects at RCS temperatures above 250*F or (2) the start of and their injection into a water solid RC5.

)

4.11 situcTUman IniicalTY ~

^

-- ^

MM The inservice inspectica and testing programs for ASME Code Class 1, 2 and 3 i

components ensure that the structural integrity and operational readiness of these coepenents will be maintained at an acceptable level throughout the lifa of the planer.

These programs are in accordance with Section XI of the ASME i

Boiler and ppessure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g Commission pur)suant to 10 CFR Part 50.55a(g)(6)(1).except where sp J

10ecr4 WTHe956cF%luncTied iW w cwic-no RMPs. Tie AWAtists is BAftD

~

00M 5TAcr ce h M4f Mr.M UdWed CF CR4AEU-CHtAsids PJMP1.

At.LCWED Bd 7% ThWdlML3Ft:GFicErscwJ5.

FARLEY-UNIT 1 B 3/4 4-14 AMDOiO(T No. 47,7th e

I l

MCM %t: PAEAWAFMC id(.4.dDet) id The &. Held, A MooiFl6D Ace Mo/e o To fh 6 3/4 4-8 ftc cedTiddW OF M65 3/4.4.10.

4 k

n

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,--..cw,.m,m,

.-..n,,wwoy._,.%-.m._--...,.._.-m,v.,~_,,m,

.-__._._,.,_.___-,_.,,.rm,_ _. - _,.

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40=1N1571AT1'il ccNT e ts ANNUat 8 tac

• R ctotxt-..........................................................

sys;g. SPte:T!! aci!'t!!? stremi 4.3.1 13 This annua. report activity analyses if t.e penar-ts sniv resutred d en the results of spectitt Isettftcation )...)

/ tsount save etteetes tne itssts of tur*.as tse year.

• 5e !C. lowing inforsattan smal' se tac hded: (1) the 11st t.as pteeded tin graonts and tacular forsatiReactor ;o.or his In.ntch

• • e last isotopic anal /s ts !st tantotoetne peristsed prior to e=ceeding the t!) Results of ustt.

results of analysis.ntle itsts.as onceesed one reswits of one snelysts after the radiotodine activity.as reduces to less taan the 11 sit.

shoul6 include date and taso of saspling and the r6diotodine concentrations:

Each result Clean up tlow nistory startang.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> petor to the first sample in vnich the (3) limit was eucteded (6) Crapn of the 1 131 toncentration teiero Citgs) other radiotodine isotore concentration (sters Ci'gs) as a function of tise Isr and see tto durati;n of the spect!!c activity above the steady. state levels and (3) t se duration when the spectfit activity of the primary coolant exceeced tne*he adiciodine limit.

A MJat SEALID $0tfRCE LEAKAct REPORT 6.9.1.la A report shall be prepared and submitted to the Commission on an annual basis it sealed source or tission detector leakage tests reveal the presence of gesater than or equal to 0 005 sacrocuries of renovable y

contamination.

WWE f.)

Y g

SPECIAL ptPORTS b

6.9.2 Special reports ehall be submitted to the Comstssion in accordance vtt*t Reports should be subeltted to the V. 5. Nuclear Regulatory Comet Document Control Desk. Vasnington. D.C. 20333.

6.to atCORD prTttrION In addition to the a

' Federal Regulations,ppitcable receed retention requirements of Title 10. Code of minimus period indicated.the following records shall be retained for at least the 6 10.1 The following records shall be retained for at least five yearsi Records and loss of unit operation covering time interval at each power a.

level.

b.

Records and logs of principal esintenance activities inspections.

repair and replacement of principal items of equipment related to nuclear safety..

ALL REPORTASLZ'tVENTS subeltted to the Commission.

c.

d.

Records of surveillance activities, inspections and calibrations requited by these Technical Specifications, Records et changes made to the procedures required by Specification s.

6.g.1.

f.

Records of cadioactive shipments.

Records of sealed source and tission detector leak tests and results.

g.

FARLEY. UNIT 1 6 10 AMEN No.

$1

@n: %cncJ,6.4.1.1eic9.2s w o G.lo W /60 To % G*20h*

FARLEY NUCLEAR PLANT. UNIT 1 PTLR SUBMITTAL TECl{NICAL SPECIFICATION MARKUPS I

l INSERT 5 PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 4 6.9.1.15 The reactor coolant system pressu4e and temperature limits, including heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1, The analytical methods used to determine the RCS pressure and temperature limits shall be those

_ previously reviewed and approved by the NRC, specifical y t ose escr eib d in the (NRC approval l h d

document)

The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

1 s

3 i

s

+

i 5

.. ~, _.,,

,.--,...,,,w-,

,,,, ~.. -,..

E '.

.,-m,...-

MfDil CEFINIT!0NS SECTION ggg 1.0 DEFINITIONS 1.1 ACTION..................................................,,,,,,.

1.}

[

1.2 AXIAL FLUX DIFFERENCE..........................................

11 1.3 CHANNEL cal!8AAT!0N..........................

1.4 CHANNEL CHECK.................................................

11 11 1.5 CHANNEL FUNCTION TEST.........................................

11 1.6 CONTAINMENT INTEGR1TY.................................................

1 2 1.7 CONTROLLED 1.EAXAGE............................

12 1.8 CORE Al.TERATION................................................

12 1.9 QOSE EQUIVALENT I 131..........................................

12 1.10 E AVERAGE DIS!NTEGRATION ENERGY...............................

13 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME...........

13 f

1.12 FREQUENCY NOTATION.............................................

13 G

1.13 GASE0WS.RA9 WASTE TREATMENT. SYSTEM (Deleted)...................

13 1.14 IDENTIFIED l.EAKAGE...........................................

13 g'

1.15 LIOWI9.RA9 WASTE TREATMENT SYSTEM Deleted).....................

14 1.16 MAJOR. CHANGES.TO.RASIGAGTIVE WAST TREATMENT. SYSTEMS (D 14 1.17 0FFSITEDOSECALCULATIONMANUAL(00CM).........................

14

/

1.18 OPERABLE OPERABILITY...............................

14 1.19 OPERATIONAL MODE MODE.......................................

15 1.20 PHYSICS TESTS.................................................

15

/g D 7 21 PRESSURE BOUNDARY LEAKAGE......................................

15 PURGE PURGING........(PCP)

PROCESS CONTROL PROGRAM 15 22 1.23 15 1.24 QUADRANT POWER TILT RATIO......................................

15 y

1.25 RATED THERMA 1. POWER............................................

16

\\

1.26 REACTOR TRIP SYSTEM RESPONSE TIME................

16 REPORTA8LE EVENT...............................................

16

(

1.27 SHUTDOWN MARGIN........................................'.......

1.28 16 1.29 SGk19IFICATION 16

(

SOURCE CHECK..(Deleted).......................................

1.30 16

/

1.31 STAGGERED TEST BASIS............................

16 1.32 THERMAL POWER..................................................

17 1.33 UNIDENTIFIED LEAKAGE...........................................

1-7 1.34 VENTILATION EXHAUST TR 17 V ENT I NG.............. EATMEN T S Y ST EM........

1.35 17 TABLE 1.1 OPERATIONAL MODES........................................

1-8 TABLE 1.2 FREQUENCY NOTATION........................................

1*9 IUSttri 1.21a R&~GJ2e Te1AE-447u're UAftu EsMT % d M

AMOmENTNo.H.h h.1 LEY-UNIT 2 1

2EREE mou!wIsTnaTIvs casrrnoos O

i moviev...................................................

b10 And1te...................................................

6-11 Authority................................................

6-12 mecords..................................................

6-12 6.5.3 TscuerICAL m3YIEW ha D coerracL Activities.............................................. 6-12 Recorde..................................................

6-13 5.4 RIPomTm.m tvruT AcTIGq..................................... 6-14 6.7 sAFETT LIMIT YIDL&TTos[..................................... 6-14 6.0 PRDcEDUREs AltD P30GRaatq........... /......................... 6-14

6. f.. RENNt!ING. REDITIEEMENT4 6.9.1 ROUTINI REPostTS 4

startup poport.......................................... 6-15a Annual Report............................................

6-16 Annual Radiological Revironmental operating poport.......

6-17 Annual medinactive Esti m nelaase moport...............

6-17 Nonthly operating Report.................................

6-19 l

PaaM ag Facter Limit Report..............................

6-19 Amausi Diesel Gemarator Reliability Data Repart.......... 6-19 Aamual meaetor coolant systen spectric activity toport... 6-20 f,

Asemal sealed semese Leakage mapart......................

4-20 IUklY 2

6..: :Pscrar. = >= irs...........................................

6->p 6.10 m23MID RETINTIG4....................,...................

6-33dt.

6.11 RADIATICEf FECTEcT1trf PROCRha(............................... 6-214

$ U NIGE R&DIATIM 1RE1........................................ 6-22 FARIEJ-OstIT 2 III

• M 180. 444--

Ikk2TE

.fOSSdsk*I6'MR:LATt./Al-UM/75 0-/DCT fo* EO'A I

i l

DEFINITIONS OPfRATfanal M00f. M00E 1.lg An OPERATIONAL MODE (i.e., MODE) sha11 correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

1 PHYSICS TEST 1 1.20 PHYSICS TESTS shall be those tests perfonned to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and i

1) described in Chapter 14.0 of the FSAR. 2) authorized under the provisions of 10 CFR 50.59, or 3) othenvise approved by the Coassission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAXAGE shall be leakage (except stens generator tube Teakage) throuqh a non.lsolable fault in a Reactor Coolant System Component body, pipe wal' or vessel wall.

IU:tL1~3 PROCESS CONTROL PROGRAM f PCP) e rou.c%

p i

b 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, i

sampling, analyses, tests, and deterutnations to be made to ensure that

[

processing and packaging of solid radioactive wastes based on demonstrated j

processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 21: State regulations: burial ground requirements: and other requirements governing thei disposal of solid radioactive waste.

(

' PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of dischartling air or gas from a confinement to maintain temperature, pressure, humidLty, concentration or other operating condition, in such a sanner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIQ l.24 Qu40 RANT POWER TILT RAT!0 shall be 'the ratto of the maximus upper excore detector calibrated output to the average of the upper excore detector call-brated outputs, or the ratto of the max < sus lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever i

is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

AMOOMENT Mf,g-)

FARLEY-UNIT 2 15 l

l -

c FARLC'lNUCLEAR PLANT UNIT 2 PTLR SUBMTITAL TECIINICAL SPECIFICATIONS MARKUPS INSERT 3 PR F9SURE TEMPERATURE I IMITS REPORT (PTLR) 1.2 Ia The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P/T) limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These P/T limits shall be determined for each fluence period or effective full power years (EFPYs) in accordance with Specification 6.9.1.15. Plant operation within these operating limits is addressed in LCO 3.4.10.1, RCS Pressure / Temperature Limits.

REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS

= _ - - -

REACTOR COOLANT SYSTEM U 4 Ts N Cido I N M M M M.lTddh 0Mf7% % R2T(PFLR h-

~

LIMITING CONDITION FOR OPERATION

~

3.4.10.1 The Reactor. coolant System (except the ressuriser) te perature and pressure shall be limited in accordance with the lici" !!::: :h

Tig.

;

..--:...?

.- ! during heatup. cooldovn. criticality, and inserv.co leak and 4 -

hydrostatic testing. e I

^ : ' r- '::te; :! 1^^*?

!- 127 ::: 5::: ;::!:d.

-c.

L.

A. ix ; ;;;1h;.

L.cf 7

y

S::: ;;:i:d.

1 : taum-tempera ture th e :e r f 1--- -- "r ^;""' *

• 1^* ' '"

r ^^:

e.

i h ;r ;;;!;d d;;!:; !a:::v4:: hyd:::t 4: : d-1**k * : !:; ;;;;;ti:::

at :: th: 5::tu; cro ::el t r- !!e!! turver.

i APPLICABILITY: At all times.

e i

pg,/p/g /A/ 7% / )*/f, ACTION:

Vith any of the above limits exceeded, restore the temperature and/or pressure to within the Itait within 30 minutes perform an engineering evaluation or inspection to determine the effects of the out-of limit condition on the fracture toughness of the FAsetor Pressure Vessel determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least il0T STAND 8Y ytthin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200 F and 300 psig, respectively, within the follIving 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, t

SURVEILLANCE REQUIREMENTS

... M w

........u..........

.5Ebc///dD it/ 77+ 9TZ.E )

4.4.10.1.1 The Reactor Coolant Syg M emperature anc'pYtT1 Tie shall be determined to be within the limitsrat least once per hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material triadiation surveillance specimens shall be removed and examined, to determine changes in material pecperties, as required by l

- 10CTR50, Appendtx 5.

a 1

AMENDMENTN0h FARLET. UNIT-2 3/4 4 27 P

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F gure 3.4-3 Farley Unit 2 Reactor Cooling System Cooldown Limitat ns Applicable for the First 14 EFPY, Farley-Unit 2 3/4 4-30 AmendmentNo.55,4 v

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4 REACTOR C00LMff 5fSTEM BASES Rducing Tayk rupture since the saturation pressure of the primary coolant isto less

[

generator tu below the Itft pressure of the atmospheric steam relief valves. The surveillance requirements provide. adequate assurar.ce that excessive specif te activity levels in the take corrective action. primary coolant will be detected in sufficient time to Information obtained on todine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic ar.slyses following power changes may be permissible if justified by the data obtained.

3/4.4.10 ptE55URE/TEMPERATUit LIMIT 5 The tesserature and pressure changes during heatup and cooldoen are limited to v

be consistent with the requirements given in the ASME Boiler and Pressure Yessel Code. Section Appendix G as required per 10CFR Part 50 Appendix G.

e/

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1)

The reacto colant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance w_ith_"l;;;ge 3.f 2-eng 3.5-2 fr CM fiin f.Il-pec eer,6 E 8'e%dEe: Tc'kTFi.7/Jd4 UMIT5 R6 Poet' (PTLE.D.

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nte ation.

b) f f =

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nonductile fatture onl For normal operation, other inherent plant I

charscteristics. e.g.,y. peep heat addition and pressurizer heater capacity, any limit the h'estup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2)

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320'F.

(

I 3) systes preservice hydrotests and in service leak and hydretaats shall be performed at pressures la accordance with the requirements of

• DER Seller and Pressure Vessel Cede, Sectica II.

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FARLEY. UNIT 2 33744..j AME.DMENT NO N

REACTOR COOLANT SYSTEM BASES The ASME approach for calculating the allowable limit curve: for vario hea arj cooldown rates specifies that the total stress intensity facto,

K, fo the combined thermal and pressure stresses at any time during atup g

or cool n cannot be greater than the reference stress intensity fa or, KIR' for the metg1 temperature at that time. K is obtained from the eference IR

-fracture toughness curve, defined in Appendix G to the ASME Cod. The K IR curveisgivenlytheequation:

KIR = 26.78 +

23exp-[0.0145(T-RTNOT + 160)]

(1) where K is the referen stress intensity factor a function of i.he metal gg

Thus, temperature T and the m6ta il ductility refere e temperature RTNDT.

the governing equation for th heatup-cooldown nalysis is defined in Appendix G of the ASME Code as follows:

CKgg + kit !KIR Where, K is the stress intensity.act caused by membrane (pressure) gg

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stress.

K is the stress in ensity factor cause by the thermal gradients.

It K

is provided y the : ode as a function of mperature relative IR to the RT f the matarial.

NOT C = 2.0 r level A and B service limits, and C=

for inservice hydrostatic and leak test operati s.

is d ermined by At ny time during_the heautp or cooldown transient, KIR the al temperature at the tip of the postulated flaw, the 'appropr te val e for RTNOT, and the reference fracture toughness curve. The therm 1 resses resulting from temperature gradients through the vessel wall ar FARLEY-UNIT 2 B 3/4 4-11 AMedMedT do.

1 REACTOR COOLANT SYSTEM i

E BASES w

=

=-

ulated and then the corresponding thermal stress intensity factor, K

.ca for e reference flaw is computed.

From Equation (2) the pressure str ss intens factors are obtained and from these, the allowable pressur are calculate COOLDOWN For the calcu tion of the allowable pressure versus colant temperature during cooldown, the Code reference flaw is assumed to ist at the inside of the vessel wall. Dur g cooldown, the controlling loc fon of the flaw is always at the inside o he wall because the thermal radients produce tensile stresses at the inside, ich increase with increas' g cooldown rates. Allowable pressure-temoerature relat ons are generated for th steady-state and finite cooldown rate situations.

Fr a these relations omposite limit curves are

. constructed 'for each cooldown te of interas The use of the composite cury in th /cooldown analysis is necessary ce-cause control of the cooldown proced e s based on meast;rement of reactor coolant temperature, whereas the limi i g pressure is actually dependent on I

the material temperature at the ti of th ssumed flaw.

During cooldown, the 1/4T vessel location is at a hi er temperatore than the fluid adjacent to the vessel 10. This condition, o course,'is not trqe for the steady-state situation.

It follows that at any give reactor coolant tempagature, the delta T developed during cooldown results a higher value of K at' e 1/4T location for gg finite cooldown rates an for s'teady-state operation.

Furthermore, if conditions exist su that the increase in KIR exceeds kit the calculated allowable pressur during cooldown will be greater than the teady-state value.

The ove procedures are needed because there is no direct con el on tempera re at the 1/4T location; therefore, allowable pressures may unknow gly be violated if the rate of cocling is decreased at various inter als along a cooldown ramp.

The use of the composite curve elimina qs thi problem and assures conservative operation of the system for the entiFq c - 1down period.

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FARLEY-UNIT 2

REACTOR COOLANT SYSTEM 8ASES Nb HEATUP

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hree separate calculations are required to determine the limit cur es for fin te heatup rates. As is done in the cooldown analysis, allowa e pressure-mperature relationships are developed for steady-state e ditions as well as f nite heatup rate conditions assuming the presence of a 1/4T defect at the side of the vessel wall. The thermal gradient during heatup produce compress stresses at the inside of the wall that leviate the tensile stresses p duced by internai pressure.

The meta temperature at the crack tip lags the co lant temperature; therefore, the for the 1/4T crack IR during heatup is lower an the K f r the 1/4T cra during steady-state IR conditions at the same co ant temperature. Durin heatur, especially at the end of the transient, condit ns may exist such at the effects of compressive thermal stresses and different R s f r stea -state cnd finite heatup rates do not offset each other and the ressure-te perature curve based on steady-state conditions no longer represen a

wer bound of all similar curves for finite heatup rates when the 1/4T f1a s considered.

Therefore, both cases.

have to be analyzed in order to ass e t at at any coc1 ant temperature the lower value of the allowable pres re cale ated for steady-state and finite heatup rates is obtained.

ThesecondportionofIheheatupanalysisco erns the calculation of f

pressure-temperature li tations for the case in wh ch a 1/4T deep 'outside surface flaw is assume Unlike the situation at th vessel inside surface, the thermal gradients tablished at the outside surfa during heatup produce stresses which are nsile in nature and thus tend to r nforce any pressure stresses present, hese thermal stresses, of course, are ependent on both the rate of heat and the time (or coolant temperature) a ng the heatup ramp. Furthe e, since the thernal stresses, at the outsi are tensile and increase with nereasing heatup rate, a lower bound curve cann t be defined.

Rather. eac eatup rate of interest must be analyzed on an ind idual basis.

Fo1 wing the generation of pressure-temoerature curves for b h the steady tate and finite heatup rate situations, the final limit curv are produ d as follows. A composite curve is constructed based on a poin -by-po comparison of the steady-statt and finite heatup rate data. At a g

n temperature, the allowable pressure is taken to be the lesser of th ree values taken from the curves under consideration.

FARLEY-UNIT 2 B 3/4 4-13 Mb 9

REACTOR COOLANT SYSTEM

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Although the pressurizer operates in temperature ranges above those for which-there is reason for concern of non-ductile failure operating limits are provided to assure compatibility of operation with the fatigue analysis perfonned in accordance with the ASME Code requirements.

The OPERASILITY of either AHR relief valve or an RCS vert opening of greater than or equal to 2.85 square ir.ches ensures that the ACS will be protected from h

pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the P.CS cold legs are less than or equal to 310*F. ~

(

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start l

of an idle RCP with the secondary water temperature of the steam generator less d than or equal to 50*F above the RCS cold leg temperatures provided seasures are 3-taken to cushion the overpressure effects at RCS temperatures above 250*F, or

(

(2) the start of geha ing pumps and their injection into a water solid RCS.

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q /4.4.11 STRUCTURAt h *" m I 2

L _ PC&

Au Idsegr4 The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of-these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section II of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a g Commission p(ur)suant to 10 CFR Part 50.55a(g)(6)(1).except where specific decr4 ed M CA46 of THe. Id.76cTuJ SVMC#6MG RJM% TW6 AdALW5 K BM6D odu $rAE.7 Lv% AMMdM UdM&ed CF oREAFA: Cabda R.lMP5 i

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1 FARLEY-UNIT 2 8 3/4 4-14 ANENOMENT NO. 8 Mars % ReuAP45 dOJC6D W% SJh !4 MOOlFico. Af6 h60 To 06e B BM La W 0>tnudm/ o; eases 3/4.4.1o.

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ADM3NISTRATIVEICONTROLS:

'AN!;UAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT.

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6;9.l'.13 'his: annual report-t's only requiredLvhen the results of specific activity a.talyses of-the primary coolant have exceeded the limits of 3

= Specification-3.4.9 during the year.- The following information shall be i included: ' (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample-J

'in which'the limit was exceeded'(in graphic and tabular format): (2). Results of i

the last isotopic analysis for radiciodine. performed prior to exceeding-the limit,-results of analysis while limit was exceeded and results of one-analysis after the.radiciodine= activity was reduced to less'than the limit.

Each result should include-date and time of-sampling-and the radiolodine concentrations: (3)

' Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to_the first sample in which the limit:vas exceeded:- (4) Graph _of the I-131 concentration (micro Ci/ga) and one other radiciodine-isotope' concentration (micro Ci/gs)-as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary' coolant exceeded the

-radiciodine limit..

ANNUAL SEALED SOURCE LEAKACE REPORT 6.9.1.14.A report shall be prepsred and submitted to the Commission on an annual basis:if sealed source or fission detector leakage tests reveal the presence of_gteater than or equal to 0.005 microcuries of removable contamination-idklIS

=

9. h oe 4SFECIAL REPORTS Y

Ps.e -

6.9.2 Special reports shall be submitted to the Commission in accordance with the requirements of 10CFR50.4 vithin the time period specified for each report.

Reports should be submitted to the U. S. Nuclear

• Regulatory Commission, ATTN:

Document Control Desk, Vashington, D.C. 20555.

t 6.10 RECORD RETElfrION In addition to the applicable record retention requirements of Ticle 10, code of Fede'ral Regulations, the following records shall be retained for at least the -

minimum period indicated.

~

6.10.1 The following records shall be retained for at least five years:-

Records and logs of unit' operation covering time interval at each power a.

level, b._ Records and logs of principal maintenance _ activities;_ inspections.

repair and replacement of principal items of equipment related to-

. nuclear safety.

c.-

aLI. REPORTASLg EVDrtS submitted to the Commission, d'.

decords.of surveillance activities, inspections and: calibrations-required by these Technical Specifications.-

e.

Records.of changes,made-to the procedures required by,' Specification 6.8.1.-

f.

Records'of radioactive shipments.

g.1 Records of sealed source and, fission detector leak tests and results.

FARLET-UNIT'2L 6-20 AMENDMENT NO. O Mrem 6C[cd; e 69.I.15i GA.2 t.@ 6 IO MOAD To S.L, 2Oa.

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FARLEY NUbLEAR PLANT - UNIT 2 PTLR SUBMITTAL

- TECHNICAL SPECIFICATION MARKUPS ~

INSERT 5 PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.15 The reactor coolant system pressure and ter yerature limits, including beatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

The analytical methoda used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the {NRC approv.si document}

- The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

4

Joseph M. Farley Nuclear Plant Unit i Pressurs Tempenture Limits Report Technical Specincation Changes Pane Ch=a=* Instruc'Jong Remove Page ;!' '

Replace Page-:

-I I

XIX XIX l5-15 3/4427 3/4427 3/4 4-29 3/4429 3/4430 3/4430 B 3/4 4 6 B 3/4 4-6 B 3/4 4 7-B 3/4 4 7 B 3/4 4 8 B 3/4 4 8 B 3/4 4-9 B 3/4 4 9 B 3/4 410 B 3/4 4-10 9

B 3/4 410a B 3/4 4-10a E 3/4 411.

B 3/4 4 11

~

B 3/4 412 B 3/4 4 12.

B 3/4 413 B 3/4 4-13 B 3/4 4-14 B 3/4 4-14 '

6 20

=6-20 6-20a 4

4 l

4

4 4

s INDEX.

T DEFINITIONS s

-- SECTION PAGEJ 1.04 DEFINITIONS 4 1

1.1 ' ACTION........................'.........................'..........1-1 1.2

-AXIAL FLUX: DIFFERENCE............'................................

1-1 1.3,

. CHANN E L - CALI B RAI ON............... '................. -.........'..... 1 - 1 e

1.4- - C HANN EL CH EC K........................................... '......... 1 - 1 1;5

' CHANNEL ~ FUNCTION TEST....-........................................

1-1 1.6 CONTAINMENTJ INT EGRI TY............................................ 1-2.

1.7 CONTROLLED LEAKAGE........................................

4.....1,

1.8 CO RE ALT E RAT I ON.................................................. 1 - 2 ;

1.9-

_ DOSE EQUIVALENT I-131............................................ 1-2

- 1 ~.10

~ E -AVERAGE DISINTEGRATION ENERGY.................................. 1-3 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME......................... 1-3

1'.12 FREQUENCY NOTATION............'...................................

1-3 v1.13-

-4Ve9aM (Deleted)......................~.1-3 1.14-- IDENTIFIED LEAKAGE............................'...................

1-3

' 1.15 LI^"n a '""?."TE T"..?r" ?? :Y:T::" ( Dele ted)....................... 1-4 1.16-

'P SVGMiM6 (Deleted)....'............................................ 1-4

-1.17 OFFSITE DOSE CALCULATION MANUAL-(ODCM)...........................

1-4

{

1.18 OPERABLE - OPERABILITY...........................................

1-4 l

1.19 O P E RA. I ONAL MO D E - MOD E.......................................... 1 - 5 1.20 PHYSICS TESTS....................................................

1-5 1.21 P RES SU RE BOUNDARY LEAXAG E........................................ 1-5 1.21a PRESSURE TEMPERATURE LIMITS REPORT (PTLR).......................

1-5 l

L-1.22 - PROCESS CONTROL PROGRAM (PCP)...........

........................ 1-5

1. 2 3 - P U RG E-PU RG I NG.'................................................... 1 - 5 1.241 -QUADRANT POWER TILT RATIO........................................

1-5 1.25 RATED THERMAL POWER..............................................

1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME................................

1-6 1.27 REPORTABLE EVENT.................................................

1-6

1. 7. 8 SHUTDOWN MARGIN..................................................

1-6 1.29 GGMGWMANGN ( Dele t ed )......................................... 1 - 6 1.30 SOURCE CHECK.....................................................

1-6 l '. 31 - S TAGG E RED ~" EST BAS I S............................................. 1 1.32 ' THERMAL-POWER.................................................... 1-7 1

1.33 UN I D ENTI FI ED LEAKAG E.............................................. 1 - 7

1.34 VENTILATION EXHAUST TREATMENT SYSTEM............................. 1-7 1.35^ VENTING.......................................................... 1-7 TABLE 1.1U iOPERATIONAL MODES..........'...............................U1-8

{

iTABLE 1.2~

-.IREQUENCY NOTATION........................................ 1-9' 4

'FARLEY-UNIT-1 I

AMENDMENT NO.

t

- INDEX

-ADMINISTRATIVE CONTROLS-

.fC.CTION

~

PAGE Review...............................................

6-10 Audits...............................................

6-11.

Authority........................................... 0-12 Records........^......................................

6.

6. 5. 3 ~

TECHNICAL REVIEW AND CONTROL

= Activities........................................... 6-12 R e c o r d s...... '........................................ 6 - 13

.6.6 REPORTABLE EVENT ACTION..................................... 6-14'

6. I SAFETY LIMIT VIOLATION...................................... 6-14 6.8 PROCEDURES AND' PROGRAMS..................................... 6 6.9 REPORTING REQUIREMENTS 6.9.1

-ROUTINE' REPORTS Start p Report....................................... 6-15a Annual Report........................................ 6-16 Annual Radiological Environmental Operating Report............................................... 6-17 t

Annual Radioactive Effluent Release Report............................................... 6-17 Monthly Ope rating Report............................. 6-19 Radial Peaking' Factor Limit Report................... 6-19 Annual Diesel' Generator Reliability Data R e p o r t............................................... 6 - 19 Annual-Reactor. Coolant System Specific._

Ac t ivi t y Re p o r t...................................... 6 - 2 0 Annual Sealed Source Leakage-Report..................

6-20 Pressure Temperature Limits Report '(PTLR)............ 6-20a 6.9.2 SPECIAL REPORTS......................................

6-20a

't 6.10 RECORD ~ RETENTION........................................... 6-20a 6.11 RADIATION PkOTECTION-PROGRAM....................

.......... 6-21a

' :6.12-HIGH' RADIATION AREA........................................ 6-22 SFARLEY-UNIT 1 XIX AME4DMENT NO.

R

" DEFINITI,ONS OPERATIONAL' MODE - MODE 11.19 L An OPERATIONAL MODE - (i.e.,: MODE) shall corresponti to_any one inclusive

. combination of core reactivity condition,' power level and average reactor' coolant = temperature.specified in Table'l.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall-be those tests performed:to measure the fundamental nuclear: characteristics of the reactor core and related instrumentation and it described in Chapter 14.0 of the FSAR, 2)-authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE-BOUNDARY LEAKAGE shall be leakage (except str.am generator tube leakage)- through a non-isolable fault in a Reactor Coolant System Component

-body,. pipe wall or vessel wall.

PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 1.21a The PRESSURE TEMPERATURE LIMIT 3 REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P-T) limits, including heatup and-cooldown rates, for the current reactor vessel fluence period. These P-T limits shall be determined for each fluence period or effective. full-power years (EFPYs) in accordance with Specification 6.9.1.15.

Plant operation within-these operating limits is addressed in LCO 3.4.10.1, RCS Pressure / Temperature Limits.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS -CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; burial ground requirements; and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING ~

1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to, purify the confinement.

QUADRANT POWER TILT RATIO

'l.24 -QUADRANT POWER TILT RATIO shall:be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated

output.to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall~be used for computing the average.

FARLEY-UNIT l' 1-5 AMENDMENT NO.

i s

i

_m_-m.____--_._

a-.....-_-.i.---

t REACTOR COOLANT SYSTEM-3/4.4.10- PRESSURE /TD(PERATURE LIMITS REACTOR COOLANT SYSTEM-LIMITING CONDITION FOR OPERATION 3.4.10.1-The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in-accordance with the limits specified in the PRESSURE TEMPERATURE LIMITS REPORT (PTLR) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

~ APPLICABILITY: At all times.

ACTION:

With any of the above limits specified in the PTLR exceeded, restore the I

temperature and/or pressure to within the limit within 30 minutes; perform an engines ring evaluation or inspection to determine the ef fects of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to 1ssa than 00*F and 500 psig, respectively, within av thefol$owing30 hours.

SURVEILLANC?. REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR at least once per l

hour during system heatup, cooldown, and inservice leak and hydrostatic testing cperations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.

FARLEY-UNIT 1 3/4 4-27 AMENDMENT NO.

t-THIS 2 AGE INTENTICNALLY LEFT BLANK FARLEY - UNIT 1 3/4 4-29 AMENDMENT NO.

THIS PAGE INTENTIONALLY LEFT BLANK FARLEY - UNIT 1 3/4 4-30 AMENDMENT NO.

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

REACTOR COOLANT SYSTEM.

T BASES-Reducing Tavg to less than 500*F_ prevents the release. of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief-valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient i

time to take corrective action.

Information obtained on iodine spiking-will

- be used to assess the parameters associated with. spiking phenomena.. A reduction.in frequency of isotopic analyses following power changes may be

- permissible.if justified by thi data obtained.

- 3/4.4.10

' PRESSURE / TEMPERATURE LIMITS t

The temperature and pressure changes during heatup and cooldown are limited to i

be consistent with the requirements given in the ASME Boiler and. Pressure Vessel 1 code,.Section XI, Appendix G, as required per 10 CFR Part 50, Appendix G. l

' ll The reactor coolant temperature and pressure and system heatup and

?couldown rates (with the exception of the pressurizet) shall be limited in accordance with the PRESSURE TEMPERATURE LIMITS REPORT (PTLR).

l a) Allowable combinations of pressure ard ' temperature for specific temperature change rates are below and to the right of the limit lines shown in the PTLR.-

Limit lines for cooldown rates between l those presented may be obtained-by interpolation.

b)'The PTLR defines limits to assure prevention of nonductile l

~

failure only.

For normal operation, other inherent plant characteristics, e.g.,

pump heat addition and pressuriser heater capacity, may limit-the heatup and cooldown rates that can be achieved over certain pressure-temperature ranget.

2)

These limit-lines shall be calculated periodically asing methods approved 4

by the NRC.

j 3)

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.

FARLEY-UNIT-1 B 3/4 4-6 AMENDMENT NO

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~

~ REACTOR COOLANT SYSTEM.

-BASES

.4)-

=The pressurizer.heatup'and cooldown-rates shall not exceed 100*F/hr:and 200*F/hr'respectively. 'The spray shall not be used if.the temperature diffurence-between.-the pressurizer'and the spray fluid is greater than-

.320'F.

I 5) system preservice hydrotests and!-in-service leak and hydrotests shall be.

performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section.XI.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 21.9 effective full power years- (EFPY). of service life. The 21.9 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in.the core region is greater than the RTndt of the limiting unirradiated material.

The selection of such a limiting-RTndt assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code.

reeuirements.

i FARLEY-UNIT.1 B 3/4 4-7 AMENDMENT No.

REACTOR COOLANT SYSTEM BA1ES Values of 4RTndt datermined in accordance with the NRC-approved methodology l

may be used until the next rs.,lts from the material surveillance program, evaluated acceedino to ASTM E185-82, are available.

Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H.

The surveillance specimen withdrawal schedule is shown in tt.e PTLR. The l

heatup and cooldown curves must be recalculated when the ARTndt determined f rom the surveillance capsule exceeds the calculated ARTndt for the equivalent capsule radiation exposure.

\\

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section XI of-l the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part50andthesemethodsarediscussedindetailinWCAP-14040-NP-A, Revision 2.l Although the pressurizer operates in temperature rangen above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements, r

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 310'F.

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with tne secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures provided measures are taken to cushion the overpressure ef fects at RCS temperatures r.

ve 250*F, or (2) the start of al' operable charging pumps and their inject 2 nto a water solid RCS.

In the case of the injection by the charging pumps,

' analysis is based on the start of the maximum number of operable charging pumps allowed by the *echnical Specifications.

r L

FARLEY-UNIT 1 B 3/4 4-8

. AMENDMENT NO.

1

L x

M M

5 t

u d

a si FARLEY-UNIT 1' B'3/4 4 9 AMENDMENT No.

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e 8

4 k-t FARLEY-UNIT 1 B 3/4 4

.0.

AMENDMENT NO.

l

THIS PAGE INTENTIONALLY LEET BLANK h

I

(

L rARLEY-UNIT 1 B 3/4 4-10a AMENDMENT No.

~...

f REACTOR COOLANT SYSTEM-i BASES-THIS PACE INTENTIONALLY LEIT BIANK l.

t a

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es a

m-

' FARLEY-UNIT 1 B 3/4 4-11 AMENDMENT NO.

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g

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_+-A.

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REACTOR COOLANT SYSTD4 BASES

.THIS PAGE INTENTIONALLY'LEFT BLANK t

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L 4

0 4

4 FARLEY-UNIT 1.

- B 3/4 4-12 AMENDMENT No,

REACTOR COOLANT SYSTEM

-gggg3-THIS PAGE INTENTIONALLY LEFT BLANK I

b 1

FARLEY-UNIT-1 B 3/4 4-13 AMENDMENT NO.

REACTOR C001 ANT SYSTEM

-BASES-E.

3/4.4.11

- STRUCTURAL INTEGRITY

.The inservice inspection'and testing programs for.ASME Code Class.1,.2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life a

of the plant.- _These programs are in accordance with Section XI of the ASME-Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR-Part.50.55a(gliexcept where specific written relief has-been granted by the Comenission pursuant to 10 CFR Part 50.55a sg) (6) (1).

-FARLEY-UNIT 1 _

B 3/4 4-14 AMENDMENT NC,.

ADMINISTRATIVE CONTROLS ANNUAL RFACTOR COOLANT SYSTD4 SPECITIC ACTIVITY REPORT 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of Spa cification 3.4.9 during the year. The following information shall be included (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded (in graphic and tabular format)s (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit.

Each result should include date and time of sampling and the radiolodine concentrationes (3) Clean-up flow history starti g 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first aample in which the limit was exceededs (4) Graph of ti.e I-131 concentration (micro C1/gm) and one other radioiodine isotope concentration (micro ci/gm) as a function of time far the duration of the specific activity above the steady-state levels and ($) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

NjNUAL SEALED SOURCE LFAXAGE REPORT 6.9.1.14 A report shall be prepared and submitted to the Commission on an 1

annual basis if sealed source or fission detector leakage testa teveal the presence of greater than nr equal to 0.00$ microcuries of removable m ntamination.

h TARLEY-UNIT 1 6-20 AMENDMENT No.

F

AiNINISTRA' RIVE CONTROLS PRES $11RE_ TEMPERATURE LfMITS_jtEPORT fPTLR1 6.9.1.15 The reactor coolant system pressure and temperature limits, including heatup rnd cooldown rates, shall be established and documented in the PTLR for LCo 3.4.10.1.

The analytical methods used to determine the RCS prossure and temperature limits shell be those previously reviewed and approved by the NRC, specifically those described in the (NRC approval document)

The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or suppiament thureto.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Commission in accordance with the requirements of 10CTR50.t within the time period specified for each report.

Reports should be submitted to the U. 3. Nuclear Regulatory Commission, ATTN Document Control Desk, Washington, D.C.

20355.

6.10 RECORD RETENTION In addition to the applicable record rutention requirements of Title 10, Jode of ruderal Regulations, the following records shall be retained fer at 11ast the minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a.

Records snd logs of unit operation covering time ir.terval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c.

ALL REPORTABLE EVENTS submitted to the Commission.

d.

Records of surveillance activities, inspections ar.d calibrations required by these Technical Specifications, e.

Records of changes made to the procedures requitert by specification 6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

FARLEY-UNIT 1 6-20a AMENDMENT NO.

Joseph M. Farley Natear Plant Unit 2 Pressure Temperature Limits Report

' Technical SpecOulon Changes PagLCMctructions Remove Page-Replace Paga -

1 I

XIX XLX l5 l$

3/4427 3/4427 3/4429 3/4429 3/4430 3/4430 B 3/4 4 6 B 3/4 4 6 B 3/4 4 7 B 3/4 4 7 B 3/4 4-8 8 3/4 4 4 B 3/4 4 9 B 3/4 4 9 B3/4411 B 3/4 4 11 B 3/4 412 B 3/4 412

. B 3/4 413 83/4413 83/4414 B 3/4 414 6 20 6 20 6 20a

INDEX DEFINITIONS SECTIO 1(

RAGE L.0 natFINITIQgg 1.1 ACTION...........................................................

1-1 1.2 AX1AL FLUX DIFFERENCE............................................ 1-1 1.3 CHANNEL CALIBRATION..............................................

1-1 1.4 CHANNEL CHECK....................................................

2-1 1.5 CHANNEL FUNCTION TE8T............................................ 1-1 1.6 CONTAINMENT INTEGRITY............................................ 1-2 l

1.7 CONTROLLE D LE AKAG E............................................... 1 ~ 2

+

1.8 CORE ALTE RATI ON.................................................. 1 - 2 1.9 DOSE EQUIVALENT I-131............................................ 1-2 1.10 E - AVERAGE DI S INTEGRATION ENERGY................................. 1 '.

1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME......................... 1-3 1.12 FRE20ENCY NOYATION............................................... 1-3 1.13 (Deleted)......................

1-3 1.14 I D E N T I F I E D LE AKAG E............................................... 1 - 3 1.15 E,49944 (Deleted).......................

1-4 1.16 44s-WAses-ensA9W4W9 SW498MS ( D e l e t ed )................................................ 1 - 4 1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)...........................

1-4 1.18 OPERAP'.8 - OPERABILITY...........................................

1-4 1.19 OPERA'10NAL MODE - MODE..........................................

1-5 1.20 PHY te ! CS TE S TS.................................................... 1 - 5 1.21 PRE 13URE BOUNDARY LEAKAGE........................................ 1-5 1.21a FRL18URE TEMPERATURE LIMITS REPORT (PTLR)........................ 1-5 l

l 1.22 'PROIESS CONTROL PROGRAM (PCP)....................................

1-5 1.23 PURGE-PURGING....................................................

1-5 1.24 QUADRANT POWER TILT RATIO........................................ 1-5 1.25 RATED THERMAL POWER.............................................. 1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME................................ 1-6 1.27 REPORTABLE EVENT................................................. 1-6 1.28 SHUTDOWN MARCIN..................................................

1-6 1.29 sOE494MGAMON ( De l e t ed )......................................... 1 - 6 1.*O SOURCE CHECK.....................................................

1-6 1.31 STAGGERED

• TEST BASIS............................................. 1-6 1.32 THERMAL POWER....................................................

1-7 1.33 UNIDENTIFIED'LEAKRGE.............................................

1-7 1.34 VENTILATION EXHAUST TREATMENT SYSTEM............................. 1-7 1.35 VENTING.......................................................... 1-7 TABLE 1.1 OPERATIONAL MODES...........................................

1-8 TABLE 1.2 FREQUENCY NOTATION..........................................

1-9 i

~ 7ARLEY-UNIT 2 I

AMENDMENT NO.

m

. ~.

_ ~.. -.

- ~.. _.

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

. ~ _. _

- ~..

i i

ljEEX

. s

-j

- ADMINISTRATIVE CONTh0LS

,p.

SECTIQE F& rag

{

VI R M iew..................................................

6-10

.Jdits..................................................

6-11 i

Au t ho r i t y................................. '.............. 6 - 12 i

Records.................................................

6-12 6.5.3 TECHNICAL MVIEW AND CONTROL

-f i

t Ac t iv;.t i e s.............................................. 6 - 12 a

Records.................................................

6-13 6.6 REPORTARLE EVENT ACTTON........................................

6-14 6.7 shPET1r LIMIT VIOLATION........................................ 6-14 F

6.8 PROCEDURES AND PROGRAMS........................................

6-14

.t r

6.9 REPORTING REQUIMMENTS i

6.9.1 ROUTINE REPORTS a

startup Report..........................................

6-15a l

Annual Report...........................................

6-16 Annual Radiological Environmental Operating Report......

6-17 3

Annual Radioactive Effluent Release Report..............

6-17 Monthly Operating Report................................

6-19

}

Peaking Factor Limit Report........................'.....

6-19

- Annual Diesel Generator Reliability Data Report.........

6-19 Annual Reactor Coolant System Specific Activity Report.. 6-20 j

Annual sealed source Leakage Report.....................

6-20 Pressure Temperature. Limits Report (PTLR)................ 6-20a 6.9.2 SPECIAL REPORT8..............................................

6-20a:

[

6'.10 RECORD' RETENTION........

.............................. 6-20a i

6.11 RAD I AT I ON PROTECTION PROGRAM......................... '......... 6 - 214 i

16.12 M I G H RAD I AT ION ARIA........................................... 6 - 2 2 -

f FARLEY-UNIT.2.

XIX AMENDMENT NO.

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PEFINITIONS OPERATIONAL MDDR - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combinwhion of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSIC 8_ TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the inndamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of th9 PSAR, 2) authorimod unuer the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PREESURE BOUNDARY LEARAQE 1.21 PRESSURE DOUNDAhY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant system Component

- body, pipe wall. or vessel wall.

EBE31MBE_IEMERB&lVBE_Mji1I1_BEPORT fPTLR)

.21a The PRES 5URE TEMPERATU3E LIMITS REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P-T) limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These P-T limite shall be determined for each fluence period or effective full-power years (IFPYs) in accordance with specification 6.9.1.15.

Plant operation within these operating limits is addressed in LCO 3.4 10.1, RCs Pressure / Temperature Limits.

PROCE55 CONTROL PROGRAM iPCP1 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastet based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; burial ground requirements; and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGINO~

1.23 PURGE and PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gar is required to purity the confinement.

QUADRANT POW'tR TILT RATIO 1.24 _QUAORANT' POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower escore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one.excore detector inoperable, the remaining three detectors shall be used for computing the average.

FARLEY-UNIT 2-1-5 AMENDRENT No.

W'

. - _ - - _ - _ _ _ _ _ - =

REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be 1Lmited in accordance with the limits specified in the PRESSURE TEMPERATURE LIMITS REPORT (PTLR) during heatup, cooldown, criticality, and inservice leak and hydrostatic testing.

APPLICA3TLITY:

At all times.

ACTION:

With any of the above limits specified in the PTLR exceeded, restore the l

temperature and/or pressure to within the limit within 30 minutes; perform an engineeritig evaluation or inspection to determine *.he effects of the out-of-1Lmit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200'F and 500 peig, respectively, within ava the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR at least once per-l hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.

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Peducing T to less than 500*F prevents the release of activity should a ayg steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Information obtained on lodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.10 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASKE Boiler and Pressure Vessel Code,Section XI, Appendix 0 as required per 10 CFR Part 50 Appendix 0.

l 1)

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with the PRESSURE TEMPERATURE LIMITS REPORT (PTLR).

l a)

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown in the PTLR. Limit lines for cooldown l

rates between those presented may be obtained by interpolation, b)

The PTLR defines limits to assure prevention of nonductile l

failure only.

For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressuriser heater capacity, may limit the heatup and cooldown rates that can be achieved over c6rtain pressure-temperature ranges.

2)

These limit lines shall be calculated pr eiodically using methods approved by the NRC.

3)

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F.

FALLEY-UNIT 2 B 3/4 4-6 AKENDKENT NO.

i BEACIOR COOLANT SYSTEM sAsrs 4)

The pressurieer heatup and cooldown rates shall not exceed 100'F/hr and l

200'F/hr respectively. The spray shall not be used if the temperature difference between the pressuriser and the spray fluid is greater than 320'F.

5) system preservice hydrotests and in-swrvice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel code, section XI.

I Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reforence temperature, RTndt, at the end of 33.8 effective full power years (EFPY) of service life. The 33.8 EFPY service life period is chosen such that the almiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unitradiated material.

The selection of such a ilmiting RTndt assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable code requirements.

FARLEY-UNIT 2 B 3/4 4 ~/

AMENDMENT No.

BEACTOR_CQQLANT_8V.AIE8 bases

. _. ~

determined in accordance with the NRC-approved methodology, may l values of ARTndt be used until the next results from the material surve111. :e program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirements of ASTM Elb5-82 and 10 CFR 50, Appendix H.

The surveillance specimen withdrawal schedule is shown in the PTLA.

The heatup and l cooldown curves must be recalculated when the ARTndt determined from the next sv.rveillance capsule exceeds the calculated 4RTndt for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown-rates are calculated using methods derived from Appendix 0 in section XI of l

the ASME Boiler and Pressure Vessel Code as required by Appendix 0 to 10 CFR 50 and these methods are discussed in detail in WCAP-14040-NP-A, Revision 2.

l Although the pressuriser operates in temperature ranges above those for which there is reason for concern of non-ductile f ailure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code reqvirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of gteater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 310*F.

Either RHR relief valve han adequate relieving capability to protect the RCS from overpressurisation when the transient is 1Laited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50'F above the RCS cold leg temperatures provided maakures are taken to cushion the overpressure ef fects at RCS temperatures above 250*F, or (2) the start of all operable charging pumps and their injection into a water solid RCS.

In the case of the injection by the charging pumps, the analysis is based on the start of the maximum number of operable charging pumps allowed by the Technical specifications.

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.B 3/4 4-13 AMENDMENT NO.

1 REACTOR COOLANT SYSTEM BASE 3

/

2],1 4.11 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughoist the life of the plant. -These programs are in accordance with section XI of the ASME Boiler and Pressure Vessel Code and applicable Addendts as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the commission pursuant to 10 CFR Part 50.55a(g)(6)(1).

4 9

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FARLEY-UNIT 2 8 3/4 4-14 AMENDMENT NO.

I

ADMINISTRAT7VE COWTROLS MEVAL. REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY. REPORT 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded t!.e limits of specification 3.4.9 during the year.

The following in'armation shall be included (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radiotodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3)

Clean-up flow hiotory starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Crnph of the 1-131 concentration (micro Ci/gm) and one other radioiodine isotope concentration (micro ci/gm) 2s a function of time for the daration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

M1[UAL SEALED SOUPCE LEAYACE REPORT 6.9.1.14 A report shall be prepared and submitted to the commission on an annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable contamination.

d FARLEY-UNIT 2 6-20 ARENDMENT NO.

ADMINISTRATIVE CONTROLS PRESSVRE TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.15 The reactor coolant system pressure and temperature 1Laits, including heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

The analytica'. methods used to determine the RCS pressure and temperature limita shs11 be those previously reviewed and approved by the NRC, specificelly those described in the (NRC approval document)

The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

SPECIAL R2 PORTS 6.9.2 special reports shall be sabmitted to the Commission in accordance with the requirements of 10CFR50.4 within the time period specified for each report.

Reports should be submitted to the U. s. Nuclear Regulatory Commission, ATTN Document Control Desk, Washington, D.C. 20555.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at Isant the minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a.

Records and logs of unit operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c.

All REPORTABLE EVENTS submitted to the Commission.

d.

Records of surveillance activities, inspections and calibre.tions required by these Technical specifications.

e.

Records of changes made to the procedares required by Specification 6.8.1.

f.

Records of radioactive shipments, g.

Records of sealed source and fission detector leak tests and results.

FARLEY-UNIT 2 6-204 AMENDMENT No.

Safety Analysis e

4 s

1

ENCLOSURE 5 Joseph M. Farky Nuclear Plant - Units 1 and 2 Pressure Temperature Limits Report -

Technical Specification Changes Safety Assessment Introduction Tuc Farley Nuclear Plant (IWP) Unit I and Unit 2 Technical Specifications (TSs) for the reactor coolant system (RCS) pressure and temperature (P T) limits curves are proposed to be relocated to the Pressure Temperature Limits Report (PTLR), consistent with the guidance provided in Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits. The P.T limits contained in the proposed PTLR incorporate the NRC approved methodology described in WCAP 14040-NP.A, Revision -

2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS lhtup and Cooldown Limit Curves, as modified at the direction of the NRC Staff, and the fluence awociated with uprated power. The license amendment request for power uprate was provided to the NRC by SNC letter dated February 14,1997. In addition to use of the revised NRC-apptcved methodology and the fluence associated with uprated power, the proposed Unit 1 P-T limitt, incorporate the results from testing of Capsule W described in WCAP-14196, Analysis of Capsule W from the Alabama Power Company Farley Unit 1 Reactor Vessel Radiation Surveillance Program. The proposed Farley Unit 2 P.T limits are based on based on the fluence associated with the NRC-approved fluence methodology contained in WCAP 14040-NP A, Rev.

2, as modified at th ' direction of the NRC Staff, the fluence associated with uprated power, and the results from testuig of Capsule X described in WCAP-12471, Analysis of Capsule X from the

- Alabama Power Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program.

The following provides a summany of the proposed changes:

1.

Revise Farley (Jnit 1 (valid through 21.9 EFPY) and Unit 2 (valid through 33.8 EFPY) RCS P-T limits, based on uprated power, in accordance with the NRC -approved methodology described in WCAP-14040-NP-A, Revision 2, as modified at the direction of the NRC Staff.

2.

Revise the Unit.1 and Unit 2 surveillance capsule withdrawal schedules based on the NRC.

approved methodology contained in WCAP-14040 NP A, Revision 2, as modified at the.

- direction of the NRC Staff, and operation at uprated power.

3.

Addition of the definition of the l'ressure Temperature Limits Report (PTLR) to Section 1.0 of the TSs.

_.mm.____

1 1

Enclosurc 5 Page 2 Safety Assessment 4.

Relocation of RCS P T limits from Technical Specification 3.4.10.1 to the PTLR in accordance with the guidance provided by Generic Letter 96-03.

5.

Revise TSs Bases 3/4.4.10 to reference the PTLR as the source document for the P T limit curves required by Technical Specification 3.4.10.1 and to incorporate, by reference, the NRC-approved methodology for generating P-T limits contained in WCAP-14040-NP-A, Revision 2, as modified at the direction of the NRC Staff. Additionally, the proposed change will revise Bases 3/4.10.1 to reference ASME Section XI, Appendix G,instead of ASME Section III, Appendix G, consistent with changes to 10 CFR 50, Appendix G (60 FR 65456, December 19, 1995).

6.

Delete Unit 1 TSs Bases Table B 3/4.4 1, Farley Unit 1 Reactor Vessel Toughness Properties, and Unit 2 Technical Specification Bases Table B 3/4.4-1, Reactor Vessel Toughness Data. This information is contained in FSAR Tables 5.2-24 and 5.2 25 for Unit I and Unit 2, respectively.

7.

Revise Technical Specification Bases Section 3/4.4.10 to eliminate detailed information regarding the methodology used to generate the P-T temperature limits that is either no longer applicable or is described in WCAP-14040-NP-A, Revision 2, which is incorporated by reference into the Bases.

8.

Clarify the description of the worst-case mass input transient used in the low temperature overpressure protection (LTOP) contained in Bases 3/4.4.10.

9.

Addition of Technical Specification 6.9.1.15 which provides the reporting requirements e

associated with the PTLR.

Safety Analysis Discussion and Evaluation During the development of the improved standard technical specifications (STS), the NRC staff agreed to allow licensees to relocate the RCS pressure temperature limits from the TSs to a licensee controlled document, provided the parameters for constructing the curves and setpoints were derived using an NRC-approved methodology. Generic Letter (GL) 96-03, Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, provided guidance to licensees for implementing this line item TS improvement. The guidance contained in GL 95-03 specifically requires licen:ees wishing to implement this line item TS improvement to:

(1) reference a methodology for developing the curves and setpoints that has been approved by the NRC; (2) develop a PTLR or a similar document that contains the figures, values, parameters, and any explanations derived from the methodology; and (3) make appropriate changes to the applicable sections of the Technical Specifications.

Page 3 Safety Assessment The following provides a description of the Farley Unit I and Unit 2 compliance with the requirement: of GL 96-03:

(1) The P-T limits contained in the proposed PTLR are applicable through 21.9 effective full-power years (EFPY) for Unit I and 33.8 EFPY for Unit 2, and were generated in accordance with the methods described in WCAP 14040 NP A, Revision 2, as modified at the direction of the NRC Staff, consistent with the requirements of 10 CFR 50, Appendix G, and Regulatory Guide 1.99, Revision 2. Additionally, the proposed P-T limits have been adjusted to account for the static and dynamic pressure differential between the reactor vessel beltline and the residual heat removal relief valves (RHRRVs) which provide low temperature

- overpressure protection for the RCS. Specifically, a 25 psi AP correction has been incorporated into the proposed P-T limits at RCS temperatures below 110'F, and a 60 psi AP correction incorporated at RCS temperatures at or above 110F, consistent with the limits on RCP operation contained in Paragraph 2.2 of the PTLR based on the maximum number of allowed operating reactor coolant pumps (RCPs). Incorporation of the AP associated with the number of operating RCPs in the P.T limits and restricting the number of operating RCPs

- at RCS temperatures below 110*F provides additional assurance that the RCS pressure at the reactor vessel beltline will not exceed the limits of Appendix G during an RCS pressure transient.

The NRC haa reviewed the methods described in WCAP 14040 NP A, and approved the topical report by issuance of Safety Evaluation Report (SER) dated October 16,1995. The NRC concluded in its SER that WCAP 14040, Revicion 1, satisfies the provisions described in a draft generic letter published in the Federal Register (60 FR 28805) for public comment on June 2,1995, which was subsequently issued as GL 96-03, January 31,1996. Revision 2 to WCAP 14040 NP-A simply incorporates the Westinghouse Owners Group response to NRC comments on Revision 1; incorporates the NRC SER approving WCAP 14040 NP-A, Revision 1; and adds the suflix NP A to the report number to designate NRC approval of the report.

(2) The proposed PTLRs for Farley Unit I and Farley Unit 2 meet the requirements contained.in GL 96-03 and are included as Enclosure 2 and Enclosure 3 of this submittal, respectively. It should be noted that Farley utilizes the residual heat removal system suction relief valves (RHRRVs) for low-temperature overpressure protection (LTOP) in lieu of a microprocessor-based cold overpressure mitigation system (COMS)in conjunction with the power operated relief valves (PORVs). Consistent with the NRC SER for WCAP-14040-NP-A, Revision 2, and GL 96-03, the LTOP requirements will be retained in the Technical Specifications in their current form and verified to provide adequate protection in accordance with the proposed methodology.

SNC has evaluated the ability of the RHRRVs to provide LTOP based on the proposed P-T limit curves and determined that a dngle RHRRV provides adequate relief capability to i

prevent the RCS pressure from exceeding the 10 CFR 50, Appendix G, steady-state limit during the worst-case heat or mass input transient at RCS temperatures less than or equal to rv-- - -

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l Page 4 Safety Assessment 310*F. The worst-case heat input transient is dermed as the start of a single RCP with a temperature difference of 50'F between the primary and secondary side of any one steam generator. The worst-case mass input transient is dermed as the injection of the maximum number of operable charging pumps allowed by the Technical Specifications for a given temperature range into a water solid RCS. The Technical Specifications specifically preclude the start of an RCP with one or more of the RCS cold leg temperatures less than or equal to 310'F unless (1) the pressurizer water volume is less than 77011' or (2) the secondary water

)

temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures, thus eliminating the potential for a heat input transient more severe than the worst-case heat input transler.t assumed in the LTOP analysis. Additionally, the Technical Specifications specifically limit the number of operable charging pumps to one whenever the temperature of one or more of the RCS colJ legs is less than or equal to 180'F, thus eliminating the potential for a mass input transient more severe than the worst case mass input y

transient assumed in the LTOP analysis.

WCAP-14040 NP-A, Rev. 2, includes the method for determination of the LTOP enable temperature, consistent with Branch Technical Position (BTP) RSB 5 2, which is defined as the water temperature corresponding to a metal temperature of at least RTmn + 90'F. The RTmn+ 90'F values for Farley Units 1 and 2 are 251'F and 276'F, respectively. Based on analysis performed by Westinghouse documented in WCAP-14689, Revision 2, Farley Urits I and 2 lieatup and Cool * 'm Limit Curves for Normal Operation and PTLR Support Documentation, the RCS water temperature that corresponds to a metal temperature of 251*F for Unit I has been conservatively established to be 279*F which bounds all possible cases during heatup and cooldown. The Unit 2 RCS water temperature corresponding to a metal temperature of 276*F has been conservatively established to be 302'F which bounds all possible cases during heatup and cooldown. Therefore, alignment of the RCS to the RHR system at RCS temperatures equal to or less than 310'F, in accordance with the requirements of Technical Specification 3.4.10 3, meet.5 the criteria set forth in Branch Technical Position RSB 5 2 for the LTOP enable temperature.

As discussed above, the LTOP enable temperature of 310'F is conservative in comparison to the value determined in accordance with WCAP-14040-NP A, Rev. 2. Due to the 8'F margin (minimum) between the proposed LTOP enable temperature and the LTOP enable temperature determined in accordance with the me6 hods of WCAP-14040-NP A, Rev. 2, additional margin for instrument uncenainty associated with the LTOP enable temperature need not be incorporated into the proposed P T limits. Since RCS LTOP is provided by mechanical relief valves, instrument uncertainty associated with the relief valve setpoint is accounted for in the calibration of the RHRRVs (445 i 5 psig); the analysis assumption that that flow does not start until inlet pressure reaches 450 psig + 10% accumulation; and the increased surveillance frequency. This eliminates the need to incorporate additional margin for instrument imcertainty into the proposed P-T limits.

GL 96 03 also requires that licensees address the minimum boltup temperature for the reactor vessel head and closure flange. Consistent with the methods described in WCAP-

E.iclosure 5 Paga5 Safety Assessment 14040 NP-A, Revision 2, the minimum boltup temperature for Farley Unit I and Unit 2 is conservatively established at 70*F as required by paragraph 2.1.1 of the proposed Unit I and Unit 2 PTLRs.

(1) Consistent with the guidance provided in GL 96-03, SNC provides the proposed Technical Specifications changes associated with the PTLR as Enclosure 4 for Farley Unit I and 2. The proposed changes consist of the following:

(a) hiodify TS Defmitions, Section 1.0, to incorporate the dermition for the Pressure Temperature Pressure and Temperature Litnits Report (PTLR).

(b) Modify the requirements of TS 3.4.10.1 to reference the PTLR as the source document for the P-T limits required by TS 3.4.10.1.

(c) hiodify Bases Section 3/4.4.10 to reference the PTLR as the source document for the P T limits required by TS 3.4.10.1, and identify that the methodology to be used to recalculate the P-T limits is NRC approved.

(d) hiisce'laneous changes to TS Bases 3/4.4.10 to remove detail that is either no longer applicable or duplicates the infonnation contained in WCAP-14040 NP A, Revision 2.

Additional clarification to the limiting mass input transient for the reactor coolant system wrs incorporated into TS Bases 3/4.4.10.

(e) hiodify TS Administrative Controls, Section 6.9.1.15, to specify NRC-approved methodology for determining the RCS P T limits required by TS 3.4.10.1.

(f) hiodify TS Administrative Controls, Section 6.9.1.15, to require submittal of the PTLR to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

Based on the items (1), (2), and (3) above, the proposed Farley Unit I and Unit 2 PTLRs, and the proposed changes to the TS, meet the requirements of GL 96-03, The Parley P T limits will be generated in accordance with NRC-approved methodology, and the plant will continue to be operated in accordance with the RCli P-T limits as required by TS 3.4.10.1. Therefore, Farley will continue to meet the requirements of 10 CFR 50, Appendices G and H, thus assuring that the integrity of the reactor vessel will be maintained.

a

10 CFR 50.92 Evaluation e- - - - - - -, -. - - - - - - - -

ENCLOSURE 6 Joseph M. Farley Nuclear Plant Units 1 and 2 Pressure Temperature Limits Report Technical Specification Changes 10 CFR 50.92 Evaluation Pursuant to 10 CFR 50.92, SNC has evaluated the proposed amendments and has determined that operation of the facility in accordance with the proposed amendments would not involve a significant hazards consideration. The basis for this determinatbn is as follows:

1. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed removal of the Reactor Coolant System (RCS) pressure temperature (P-T) limits from the Technical Specifications (TSs) and relocation to the proposed Pressure Temperature Limits Report (PTLR) in accordance with the guidance provided by Generic Letter (GL) 96-03 is administrative in that the requirements for the P-T limits are unchanged.

The P-T limits proposed for inclusion in the PTLR are based on the fluence associated with 2775 MW thermal power and operation through 21.9 effective full power years (EFPY) for Unit I and 33.8 EFPY for Unit 2. GL 96-03 requires that the P-T limits be generated in accordance with the requirements of 10 CFR 50, Appendices G and H, documented in an NRC approved methodology incorporated by reference in the TSs. Accordingly, the proposed curves have been generated using the NRC-approved methods described in WCAP.

14040-NP A, Revi: ion 2, as modified at the direction of the NRC Staff, and meet the requirements of 10 CFR 50, Appendices O and H. TS 3.4.10.1 will continue to require that the RCS pressure and temperature be limited in accordance with the limits specified in the PTLR. The NRC-approval document will be specified in TS 6.9.1.15 and NRC approval will be required in the form of a TS Amendment prior to changing the methodology. Use of P-T limit curves generated using the NRC approved methods will provide additional protection for the integrity of the reactor vessel, thereby assuring that the reactor vessel is capable of providing its function as a radiological barrier.

TS 3.4.10.3 for Farley Nuclear Plant (FNP) Unit I and Unit 2 provides the operability requirements for RCS low temperature overpressure protection (LTOP). Specifically, TS 3.4.10.3 requires that two residual heat removal (RHR) system suction reliefvalves (RHRRVs) be operable or that the RCS be vented at RCS cold leg temperatures less than or equal to 3 In'F. Cons: stent with GL 96-03, the Farley Unit I and Unit 2 requirements for LTOP will be retained in TS 3.4.10.3 and will be evaluated in accordance with the proposed methodology.

Based on the above evaluation, the proposed changes are administrative in nature and do not involve a significant increase in the probability or consequences of an accident previously evaluated.

1 Page 2 Significant Ilazards Evaluation

2. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

As stated above, the proposed changes to remove the RCS P-T limits from the TSs and relocate them to the proposed PTLR is an administrative change. Consistent with the guidance provided by GL 96-03, the proposed P-T limits contained in the proposed PTLR meet the requirements of 10 CFR 50, Appendices G and H, and were generated using the NRC-approved methods described in WCAP-14040 NP A, Revision 2, as modified at the direction of the NRC Staff. The proposed changes do not result in a physical change to the plant or add any new or different operating requirements on plant systems, structures, or components with the exception oflimiting the number of operating RCPs at RCS temperatures below 110'F. Limiting the number of operating RCPs below 110*F results in a reduction in the AP between the reactor vessel beltline and the RHRRVs, thereby providing additional margin to limits of Appendix G. Provisions are made to allow the start of a second RCP at temperatures below llo'F in order to secure the pump that was originally cperating without interrupting RCS flow. The LTOP enable temperature exceeds the minimum LTOP enable temperature determined as described in WCAP-14040 NP A, Rev. 2, thereby providing t.dditional assurance that the LTOP system will be available to protect the RCS it' the event of an overpressure transient at RCS temperatures at or below 310'F. Based on the methods contained in WCAP 14040 NP-A, Rev. 2, the minimum boltup temperature for the reactor vessel flange region is conservatively established as 70'F.

As stated in the above response, implementation of the poposed changes do not result in a significant increase in the probability of a new or different accident (i.e., loss of reactor vessel integrity). The RCS P T limits will continue to meet the requirements of 10 CFR 50, Appendices O and H, and will be generated in accordance with the NRC approved methodology described in WCAP-14040-NP-A, Revision 2, as modified at the direction of the NRC Staff. Therefore, the proposed changes do not result in a significant increase in the possibility of a new or different accident from any previously evaluated.

3. The proposed change does not involve a significant reduction in a margin of safety.

The margin of safety is not affected by the removal of the RCS P.T limits from the TSs and relocating them to the proposed PTLR. The RCS P-T limits will continue to meet the requirements of 10 CFR 50, Appendices G and H. To provide additional assurance that the P-T limits continue to meet the requirements of Appendices G and H, TS 6.9.1.15 will require the use of the NRC-approved methodology to generate P.T limits. The RCS LTOP requirements will be retHned in TS 3,4.10.3 due to use of the RHRRVs for LTOP, consistent with the guidance proviAd by GL 96-03, and will be verified to provide adequate protection of the reactor coolant system against the limits of Appendix G. The LTOP enable temperature exceeds the LTOP enable temperature determined in accordance with the NRC-approved methodology, thus protecting the RCS in the event of a low temperature overpressure transient over a broader range of temperatures than required by WCAP-14040-

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Significant Haznds Evaluation NP-A, Rev. 2, Administaktive procedures will preclude operation df the RCS at temperatures below the minimum boltup temperature for the reactor vessel hud, thus precluding the

. possibility of tensioning the reactor vessel head at RCS temperatures below the minimum boltup temperature. Operation of the plant in accordance with the RCS P.T limits specified in -

the PTLR and continued operations of the LTOP system in accordance with TS 3.4.10.3 will J continue to meetihe requirements of 10 CFR 50,~ Appendices G and H, and will therefore, Lassure that a margin of safety is not significantly decressed as the result of the ; roposed

-changes.

Based on the preceding analysis, SNC has determined that removal of the RCS P.T limits from the TS and relocation to the proposed PTLR will not significantly increase the probability or.

consequences of an sccident previously evabated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a signi6 cant reduction in a margin of safety. SNC therefore concludes that the proposed change meets the requirements of 10 CFR 50,92(c) and does not involve a significant hazards consideration.

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