Text

Amend 79 to License DPR-56,changing Tech Spec to Permit Reactor Operation W/Reload Number 4 Core (Cycle 5)
	ML20010J257
Person / Time
Site:	Peach Bottom
Issue date:	09/16/1981
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Atlantic City Electric Co, Delmarva Power & Light Co, Philadelphia Electric Co, Public Service Electric & Gas Co
Shared Package
ML20010J258	List: ML20010J257; ML20010J260; ML20010J261;
References
	DPR-56-A-079 NUDOCS 8109300051
	Download: ML20010J257 (47)
	v • d • e

{{#Wiki_filter:~ - =. >= i p ne2 o UNITED STATES d ~g NUCLEAR REGULATORY COMMISSION n ,1 wAmimoTow. o. c. 2 cess j/ ' PHILADELPHIA ELECTRIC COWANY l PUBLIC SERVICE ELECTRIC AND GAS COWANY l 'n ( DELMfRVA 90WER AND LIGHT COMPANY I. Y" -ATLANTIC CITY ELECTRIC COMPANY l DOCKET NO. 50-278 i PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 3 f[ AMEN 0 MENT TO FACILITY OPERATING LICENSE f Amendment No. 79 License No. OPR-56 >,.t i 1. The Nucl'ead Regulatory Commission (the Commission) has found that: I The applications for amendment by Philadelphia Electric Company, et A* al. (tne licensee) dated September 30, 1980, and March 30,1981, as t supplemented April 24, June 30 and July 15, 1981, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended l (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in confomity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health h and safety of the public, and (ii) that such activities will be I conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements j have been satisfied. 2. Accordinglj, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-56 is hereby amended to read as folicws: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Ar.endment No.79, are hereby incorporated j in the license. PECo shall operate the facility in accordance with the Technical Specifications. 8109300051' 81091E PDR ADOCK 05000278 { P PDR ~._,... _._. _ ~. _ _ _.. _,_- _. _ _ _ _. _.____ _ __

D - O , 3. This ifcense amendment is effective as of the'date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION b f John F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications . Date of Issuance: September 16, 1981 O 4 6 1 1 i i i l l l l l i I

__-.~. ATTACHMENT TO LICENSE AMENDMENT NO. 79 FAOILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 Replace the following pages of t'te Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. Remove Pages Insert Pages iv iv vi vi 4 4 i 9 9 10 10 lla lla 13 13 14 14 17 17 19 19 20 20 ~' 21 21 24 24 31 31 33 33 40 40 74 74 91 91 92 92 103 103 104 104 111 111 133a thru 133d 133a thru 133d 133e i 140 140 140a thru 140e 140a thru 140d 142 142 142a thru 142c 142a thru 142c 142e thru 142g 142e thru 142g i 142h 1421 152a 152a 157 157 241 241 i 4

P3APS Unit 3 LIST OF FIGURES _ Ficur_e Title Pace 1.1-1 APoM Flow Stas Scram Relationship To 16 Normal Operating Conditions 4.1.1 Instrument Test Interval Determination 55 Curves 4.2.2 Probability of Sv.sta: Unavailability 98 Vs. Test Interval 3.4.1 Required Volume and Concentration of 122 Standby Liquid Centrol System Sclution 3.4.2 Required Temperature vs. Concentration 123 for Standby Liquid Control System Solution 3.5.K.1 MCPR Cperating Limit Vs. Tau 142 8x8 and 8x8R Fuel 3.5.K.2 MCPR Operating Limit vs. Tau, P SXSR Fuel 142a igures 2.5.1.a. anc 3.s..l.3 (.x., e = r Fuel) dz.leted IP33 Cycle 5 - all SXS core) 3.5.1.C MAPLEGR Vs. Planar Average Exposure, 142b Unit 3, 3x8 Fuel, Type H 3.5.1.D MAPLEGR Vs. Planer A-erage Exposure, 142:. Unit 3, Sr3 Fuel, Type L 3.5.1.2 Kf Factor vs. Core Fl?w 142d 3.5.1.F MAPLEGR Vs., Planar Average Expcsure, 142e Unit 3, Ex8. PTA Fuel 3.5.1.G MAPLEGR Vs. Planar Average Exposure, 142f t.n.:. x :. :..... 1

v. AP _H GR * 's. Planar i.verage Exp sure.

14 g 2.5.'.H =..v e. :..... -,.: e. :... - c.,. =.. > 2 s 3.5.1.I MA?LEGR vs ? lana: Average 142h Exposure, Unit 3, PcXcR Fuel (PSDR3299) 3.5.1.J MA?LEGR vs. Planar Average Ex;:sure, 'Jni: 3, PSX5R Fuel (Generi:' 1421

u. c s -.. -.,-,.

a-.,.,...- 2. a. c..e..., ..,g g u.. s...'. _= s. =.. " 4.. = d. ' v,.C a...'.. X. .v. 4 4.........m.........,.... v.a.. = 4..,, =..,. =.... cr Cooldown felicwing Nuclear Shu:down a 3.6.3 Minimum Temperature for Core Operation 164b (Criticali:y) 3'.5.4 Transition Temperature Shift vs. Fluence 164c 244 6.2-1 Management Organization Chart 6.2-2 Crganization for Conduct of Plant Operatien 245 mandnent io'.,'A,AT,jd, g,jg,79 -iv-

?BAPS Unit 3 LIST OF TABLES l Table Title Race 4.2.3 Minimum Test and Calibration Frequency 31 for CSCS 4.2.C Minimum Tes and Calibration Frequency 33 for Con rcl Red 31ccks Actuation 4.,2.D Minimum Test and Calibration Frequency 84 for Radiation Monitoring Systems k.2.E Minimum Test and Calibration Frequercy SS for Drywell Leak Detection 4.2.? Minimum Test and Calibration Frequency 86 for Surveillance Instrumentation 4.2.G Minimum Test and Calibratien Frequency 58 for Recirculation Pump Tri? 3.5.K.~ Cperating Limit MCPR Values for 133d varicus Core Expcsures 3.5.E.3 Cperating Limit MCPR Values for 133e

~ ricus C:re Expcsures

.a ~= =~. a , :_1 a--- 3.5-2 DELETED 4.6.1 In-Service Inscection ?recram for Peach 150 20: tem.. units, anc 2 l 3.7.1 Primary Cen ainment :sclation Valves 179 .i .3 Testacle Pene: rations Wi-h Ocuble 134 0-Ring Seals 3.7.3 Testable Penetrations With Testable 154 se i.3 c w s 3.7.4 Primary Containment Testable Isclation 135 Valves t.S.1 Radicactive Liquid Waste Sampling 210 and Analysis 4.3.2 Radicactive Gasecus Waste Sampling 211 and Analysis 234d 3.11.D.1 Safety Related Shock Suppressors 240k 3.14.C.1 Fire Detec:crs Amendment,lo. frf, M 79 _vi-

PSAPS 1.0 DEFINITIONS (Cont'd) compenent, or device to perform its function are also capable of performing their related support function. O=eratinc - Operating means that a system or component is performing its intended functions in its required manner. C=erstinc Cvele - Interval between the end of cne refueling outage for a particular unit and the end of the next subsequent refueling cutage for the same unit. Primarv Containment Intecritv - primarv. containment integrity means na ne drywell anc pressure suppression chamber are intact and all of the follcwing condi; ions are satisfied: 1. All non-autcmatic centainment isolation valves on lines connected to the reactor coolant system er centainment which ~ - are not required to be open during accident conditions are ~ closed. These valves may be opened to perform necessary operational activities. 2. At least ene deer in each airlock is closed and sealed. 3. All autcmatic rentainmen; ise'.a:icn ealves are perable :r feactiva:ed in the isclated posi icn. 4. All blind flanges and manways are closed. Frr:ective Acticn - An acticn initia:ed by the prc ectica syster l wnen a :m

s reached.

A prctective action can ce a: a channel l cr system level. 1 re ective Function - A system protective action which results o frcm :ne protec :ve action of the channels menitcring a particular plant condition. Rated Power - Rated power refers to operation at a reac:ce power of 3,293.MWt; this is also termed 100 percent power and is the maxi =um ocwer level authorized by the operating license. Rated steam flow, rated ccclant flow, rated neutron flux, and rated l nuclear system pressure refer to the values of these parameterr when the reactor is at rated cower. l l Arendment No. X # I,79 _4_

PBAPS Unit 3 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Acolicabilitv: Acclicabilitv: The Safety Limits established The Limiting Safety System Settings to preserve the fuel cladding apply to trip settings of the instru-integrity apply to these ments and devices which are provided variables which =cnitor the to prevent the fuel cladding integrity fuel. thermal behavior.. Safety Limits frem being exceeded. Cbiectives: 'Ob iectives: The gbjective of the Safety The objective of the L'imiting Safety Limits is to establish limits System Settings'is to define the level which assure the integrity of of the process variables at which auto-the fuel cladding. matic protective action is initiated to prevent the fuel cladding integrity Safety Limits frem being exceeded. Scecification: Scecification: I The limiting safety system settings shall be as specified belev: A. R'eactor cressure 1S00 esia A. Neutron Flux Scram and Core Flow 210*. of Rated The exister.re of a minimur 1. A?RM Flux Scram Trio Set'tinc critical pecer ratio MCPR less 'Rur Mede) tnan 1.07 for two recirculation Icep cperatien, or 1.08 When the Mode Switch is in the for single locp cperation, RCN position, the AIRM flux shall constitute viclatien scram trip setting shall be: cf the fuel cladding integri:y safety limit. S.S 0.66W -545.-0.66 AW To ensure that this safety where: limit is not exceeded, neutrcn i flux sh-ll not be above the S - Se ting in percent of scram setting established :.n rated thermal pcwer specification 2.1.A for (3293 MWt) longer than 1.15 seconds as indicated by the process ccm-W = Lcce recirculating flew puter. When the process computer rate in percent of design is out of service this W is 100 for core flow safety limit shall be assumed ef 102.5 million lb/hr to be exceeded if the neutron er greater. flux exceeds its scram setting and a control rod scram does not occur. /cendnent t'o. 8. M 8.79. w. -av--- -w y m-w g or+ q = w.p4 n, -e m w --w, - --p--gy-- -,eenp-w

-M*

y v-

PBAPS e377 3 SA:srY LIMIT LIMITING SAFETY SYSTEM SETTING 2.1.A (Cont'd) In the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows. S < (0.66 W - 54% -0.66 AW) ( FRP) MFLPD

where, FRP = fraction of rated thermal power (3293 MWt)

MFLPD E m'aximum fraction of limiting power density vnere the limiting Power density is 13.4 KW/ft for all Sx3 fuel. The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual cperating value will be used. 2. APRM--When the reacter mcde switch is in the STARTUP pcsition, the APRM scram shall be set a: less than er equal to 15 percent of rated power. 3. IRM-The IRM scra= shall be set at less than er equal to 120/125 of full scale. 4. When the reactor mode switch is in the STARTUP or RUN pcsition, the reacter shall not be operated in the natural . circulation flow mode. Ar.endnent No. J C g, g K 79 PSAPS Unit 3 SAFETY LIMIT LIMITING SAFEiY SYSTEM SETTING

3. Core Thermal Power Limit 3.

APRM Red $1ock Trip Settino-(Reactor Pressure s 800 esia) SR3 <(0.66 W + 42%- 0.66 AW) (FRP) MFLPD t where: FRP = fraction of rated thermal power (3293 MWt). ~ MFLPD = =aximum fraction of limiting power d'ensity ~ where the limiting Power density is 13.4 KW/ft for all 3x3 fuel. The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating ~ value is less than the dehign value of 1.0, in which case the actual operating value will be used. C. Whenever the reactor is in the C. Scram and isolation--tS33 in. above shutdown condition with react:t icw water vessel ero irradiated fuel in the reactor level (0" on level vessel, the water level shall instruments) net be'less than 17.1 in, above the top of the normal active fuel :ene. Ar.endment No. X,79 -11a-l l

~ PBAPS Unit 3

1.1 BASES

FUEL CLADDING INTEGRITY A. Fuel Claddinc Inteority Limit at Reactor Pressure > 300 ~ ~ ~ ~ csia and Core Flow >10% of Rated l The fuel cladding integrity safety limit is set sud that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result'in damage to SWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient. limit. _However, the uncertainties in monitoring the core' operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit io defined as the critical. power ratio in the limiting' fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition-considering the power distribution within the core and all uncertainties. The Safety Limit.v.CPR is determined using the General Electric Thermal Analysis 3 asis described in references i and 3. * ~ '~ Amendmnt ilo. )(, 79

_32 A.D S LTn.4* 3 1.1.A BASES (Cont'd) 3. Cere Thermal Power Limit (Reactor Pressure < S00 esia on Core Flow < 104 of Rated) The use of the GEXL correlation is not valid for the critical pcwer calculations at pressures below SOO psia or core ficws less ..".an 10% c'. .a.ed. "..".=.. =. # - * .5 =. ' ". a. l . a d d i .c.

4... =. c,. 4..v, s a #. a..v.

limit is esrablished by other means. This. s cone by ~ establishing a limiting concition c:. core ther=al power operation wi,th the following basis. . Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core eressure drop at low pcwer.and all ficws will alwav.s be c.reater than 4.56 e.si. Analyses show that with a flow of 23 x 103 lbs/hr bundle flow, bundle pressure drop in near3y independent c wunc.3e power and-s has a value of 3,5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 2B x 103 lbs/hr irrespective,of total core ficw and independent cf bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at ressures from 14.7 psia to SCO psia indicate that the fuel eassembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thermal pcwer of =cre than 50%. Therefere a :cre thermal power ~~* -e.=

c...- a.

.' '. w 1 4... 4.. - 3.s :-. .==c,...- -. =. e s ".. a. s " =.1. w .ww s -- ~ 7 n -.. c a....,. 4.J a.. . =-. . u. : - ,.s a... ..a .2 4 . - e 5. e....

-.. _ r.

-L. 2. Plan: safetv analvses have shown that the scrams caused bv 1 s e... s. . u. s.-..u.=. $ c. : =...; " --..4.. C: t- ... s c _"...; e n.. 4.g y 4..' '. aX e.s.s.a..e, .:= .y f -w.1c... 4...a.g q. '. '. .n. - > a x e. s. s. ;. =. 4. s s.'.- ,~-. e.. 4 ., e.m. ).,..,s ne ca 4 w. c. =.. ". =. e. k a. ?. - =.. '. - A '. =.'.' v, . c ' s s ". - =.. ". =. i.. c =.. '. w-.. '... a. s = - =. y = A =. r.. e. s.. . h. a. . w g e... g. r wa.r. . =...c. e. =. r..... =. g..i. *

C.

...=.. a sC.:.

i. s

.-~ .r

1. p - -..~y. s. c h. a. d

e. =. %.. a. e. =... - e4 =. m. p

2. y - c. e. *. :. 5 q-....

.c. -. C.1 ' a.. w. J,.. G.s. . v .. y = .w .y. w. su.=. ... =. -.... -....".--4..=. ..+ .c........,. 1 3 m x :..,.: -w..., e valves) does not necessarily cause fuel damage. Arendaent.'!o. X M, g 79,

2.1 3ASES

FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Peach Bottom Atomic Power Station units have been analv ed throughout the spectrum of planned operating conditions up'to the thermal power condition of 3440 MWt. 3293 MWt is the licensed maxir.um power level of each Peach Bottom Atomic Power Station unit, and this represents the maximum steady state pcwer which shall not kncvingly be exceeded. Conservatism is incorporated in the analvsis of f ast pressurization transients as cescribec. in re erence 3. Conservatism is incorporated in all other, transient analyses in estimating the controlling fac ces, such as void reactivi y coefficient, control rod scram worth, scram delay time, ieaking ~ factors, and axial power shapes. These factors are selected . conservatively with resc.ect to their effect en the ac.clicable transient results as determined by the current analysis models. These transient models evolved over many years, have been substantiated in operation as a conservative teci for evaluating reac:cr dynamic performance. Results obtained frem a General Electric boiling water reactcc have been compared and results are sum =arized in Reference 1 for cold water events, and in Reference 2 for pressurization events. The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 25% greater than the nominal. maximum value expected to cccur during the core lifetime. The scram worth used has been derated to be ecuivalent .. = --. x i... a. =.1 v, 20.4 w.# . 5 =. ~..~. c.I s s-. c... ' c. *. 7. #. .'.=.--....1 .. c'. ry . ". =.. - -. = ~.. A. =. '.= v..i..=. =..d .1.=..= -.A. '...c=..'.... =. ' ' c u - ?. " v, .. '. =. c ... - =_ s.. =...= v,. =.. ' =...=.'~f.=.s.=-=..-..es.."..=.-"=..'v, s=....-"..=.' ..... =. c o r g n. a. _. c..a... g. c .a. g.,,. .y ,,... a.. c. .e.. w, u,.g.

2... g a... : -. - a. a.

7 , m. '. ~..~..=.1 s o. =. '1-Cw. ..k. a. e-S n' w.. ". =. c.c.'.=.. '. l ew '. s =.. u a ' ~. ~ woe

a. # # =... c ' s w-. a..

w c..,."a, s c. a... d a. l v, . i...e a n d.. ". 4.s e.. d -.. .c-=., .c .=. 1 -..s m.. 2. 4. 3.1.. =. 7 7. e. m.., c. s..s ,. 3. c.. s. e..

e. 4.,. 4. :. 4......

4....w.

s 2 f.;-. g:..u. a. .=. e g. : y a. - =..=.- ; n. ;.. n.

a. en...:.... -...

... a..cr.. =..=.. y .=. c,a. 4 ". a. . =.= c. i. v i..v. .i s a s - ". - =. d. ". "y .. ". =.. #... a. _- - =..'... a. ^. . ;.. a. . u. u a..s c., a.

. e u.. t....

... a. n -=....i.-=...a....e :.- - s. ,.. a ... c. -.. x 4.... c-. =., y : - r..- 4. .3 =, .o...=.<... .. = e. ;. 4.. - .f ... e m.

m. e., -

t.. ;.

r -...-.y... .,,-.g. u.,.. ,.a.,q., .c. y =. -,...ga...,.; ....w s. y .w acccmelishes the desired effect. The times for 50% and 90% .xy.r.. 4..e.....4..c.. g 4. v o....- as L... -.Cr.. .... y.. ~. 4.Cn c:. w.. 1 o s . f .a.

c... n c a.

4. 7..h. a.

=. a - 1 4. =. - - r..i .7 c '. ..s. e.. a.. s 1 =..., and ~~ w w e c establish the ultimate fully shutdown steady state condition. hendcent No.J3'; 79 qp

P3APS Unit 3 2.1.A 3 ASIS (Cont'd.) An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reacned. The APRM scram trip setting was determined by an analv. sis of mar-ins required to provide a reasonable range for 9 maneuvering during 0;eration. Reducing this cperating margin 4 wculd increase the frequency cf spurious scrams which have an adverse effect on tsarter safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because ~. it provides adequate margin fer the fuel cladding integrity Safety Limit yet allows operating margin that reduces the , pessibility of unnecessary scrams. The scram trip setting must be adjusted to assure that the LEGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFL?D) and reacter core thermal power. The scram setting.is adjusted in accordance with the formula in Specification 2.1.A.1, when the MFLPD is greater than the fraction of rated power (FRP). ~ An'alyses of the limiting transients show that no scram adjustment is required to assure MCPR greater than the fuel cladding integrity safety limit when the transient is initiated from MCPR greater than the operating limit given in Specification 3.5.K. s.=-.".y ..w-d a e. '.1 =. '..".a. .=.=e.'.w-. '.s.*. 1.,w- .,- - y. i 4. .7 i.n ..h a. a = n w eJ. e. 3 ; .e y...o. g.. a., m.

i. ;,

.m.e......=., . s. s. O...u. 3.=.=.. gm.. 4 c... i ..... y ....y y. ....g=..4.... cA 6. a. 2 4s . a s., =.g i... c.. y. 4 ..g ag - g..g 12.,.. f4 : 5 c.o. .y.. j y.. v.f

a. g -w.e.

..e..,... 4 s c c a.,.. a. a.

m

+3 e -. a. ;. .w. m

7. 3 f.. w n., w.

.c. c..,/ a -a .p. .c A a 2 .1 rs*. c s. .....c d a. a. .=.. s '. " '. c. a. d,,. .. c-. a. ' 1 v =.. s a s s w- '. - *. = c' w i.. h ycwa.. c w c f --c.. w y w s.=..." o,. 7.' # a. c *.. c. .., # i.. c. a. c s '.., - *. =. s s "s. - a. c'..-=.o c. 1-w vcid y .=...4...t. ,.4.. l ..3 c. s.- .g c. e. .y .e,.. = 3 w.. J. g Awge. . h. g .W=... g.i * =. 2 M.

5.**.
** a.

C.',/E".*.3.*.*., ..6

J y

e....- ....y

s. m1a.s......

..-f ec..=.s. g g e a. 3. s. ..,. i.= a. w.c. .J.4-4ge. 3 .m o .e =... c.I.1, c d. J a ..y .w... ..su.s c, ne.,ca.po.na.g .A . u... a. uy a.,y s, a.

a. :. n.....,/

.y. l g. at...a.; y. t,.~. ...,1 .c...e..=.... .gn. e.. s. .. a_ : w f ge. e

u. v....s.,..e.. = e.

c .c a. n.. a. n. w. -a 3 c '. 4..d..i v i d" a.'. s i.s v=..v. 1s i..n a u.'..' ... d e-..=..... -'.h u s, -d c of all pcssible scurces of reactivity inc.ut, uniform control red ..rocac e cause c: sicni:: cant e.cwer rise. c wi:..drawal is tne Ics: ecause the :1ux c..istribution associated witn uni:crm roc swithdrawals does not involve high local peaks, and because j l several rods must be. moved to change power by a significant percentage of rated cower, the rate of cower is very slew. i ~ Generally, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than 5 percent of rated pcwer per minute, and the APRM system wculd be more than adequate to assure a scram before the power could exceed the Safety Limit. Tne 15 percent APRM scram remains active until the mode switch is placed in the RUN pcsition. This switch occurs when the reacter pressure is greater than 350 psig..s..w. u._n n. - ~.o. i, f,p7.,79 i l

PSAPS Unit 3 l 2.1.A 3 ASIS (Cont'd.) The IRM syste= consists of 3 chambers, 4 in each of the reactor protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5-decades are covered by the IRM by means of a range switch and the 5-decades are broken dcwn into 10 ranges, each being one-half cf a decade in size. The IRM scram trip setting of 120 divisions is active i.c each range of the IRM. For example, if the instrument were en range 1, the scra= setting would be 120 divisions for tna: range; likewise, if the instrument were on range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accc=modate the increase in power level, the scram trip setting 'is also ranged up. The = cst significant sources of reactivity change during the power increase are due to centrol red withdrawal. For in-secuence control red withdrawal the rate of change of pcwer is sicw enough due to the physical limitation of withdrawing control rods, that heat flux is in ecuilibrium with

the neutron flux and an IRM scram would result in a reactor shutdown well before any Safety Limit is exceeded.

In order to assure that the IRM provided adequate protection against the single red withdrawal error, a range of red withdrawal accidents was analyzed. This analysis included starting the accident at varicus power levels. The = cst severe case invcives an initial conditien in which the react:r is just su= critical and the IRM system is not ye: en scale. This

end;; ion exis:s a: cuar:er red density.

Addi:icnal conserva:is: sas taken in this analysis Oy assuming tha the IRM channel closes: :: :he withdrawn rod is bypassed. The results of this analysis show that the rea::: is scra=ed and peak power limited

c :ne percen: Of riced pcwer, thus maintainin? MCPR above the fusi cladding integri y safety li=it.

Based c a the abcVe analysis, the IRM provides protection against lccal centr:1 ::d sichdrawal errors and cen inucus withdrawal Of : n:rel rods in-sequence and provides backup prc:e::ict f:r the APRM. 5. APRM Ecd 310ck Tric Se::inc The APRM system provides a control rod bicck to avoid conditiens which would result in an APRM scram trip if allowed to proceed. The APRM red block trip set ing, like the APRM scram trip setting, is autc=atically varied with recirculation lecp flow rate. The flow variable APRM rod block trip setting provides margin to the APRM scram trip setting over the entire recirculation ficw range. As with the APRM scram trip setting, the APRM red block trip setting is adjuste/ if the maximum fraction of limiting power density exceeds the fraction of rated =cweri thus preserving the APRM rod bleck safety =argin. As with .the scram setting, this =ay be accomplished by adjusting the APRM gain. ~ A..en caen t :!c.,E,,M,,6f,7 9

PBAPS Unit 3 2.1 BASES (Cont'd.) ~ C. Reactor Water Low Level Scram and Isolation (Excect Main Steam 11nes) The set point for the low level scram is above the bottem of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results reported in PSAR subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than the fuel cladding integrity safety limit in all cases, and system pr' essure does no.t reach the safety valve settings. The scram setting is approximately 31 in, below the normal operating range and is thus adequate to avoid spurious scrams. 3. Turbine Stoo Valve Closure Scram The turbine stop valve closure sc am trip anticipates the pressure, neutren flux and heat flux ir. crease that could resul.: frgm rapid clesure of the turbina stop valves. With a scr&m trip ~ setting of less than er ecual to 10 percent of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity safety limit even during the, worst case transient that assumes the turbine bypass is closed. This scram is bypassed when turbine s _ =_ _=... ' ' w i. s " =. i. w -S c '..=.=.d., c s. =..= s ". - =. A. . ".. - i..a. ..i s. .v. .-._=3 ,_ - =. c _e s.. -2 E. Turbine Cen rcl Valve Scram c n. - d. ea. a. s '..". =. . 's

m. w....,.. w 4..,...

., y a.iv. :=s. 1.s".=. s-..=m. u... a u - a..d...,"=.u., '-....., .., ' a, c c- ;. '.. ". -....-. =..= s e... _=..... d..=.eu ..=.....-.#r .. =. e s "u. =., + ..,sa.,, ..- -.....e.. ..,.s.a., c-c... reJec icn exceeding the capacity of the bypass valves er a l , s v. -. =. - *... u.".. =. s "..'. s.'. a- .. y s.. u. . - -. p..... .u_s -e w .....4., .w. 2 . - '. '.-. a. s s ". =... ". '. s e -. =... ' s

4.. '.. '..=. =. 4. '.... -. =. s s "w. =. s~".

.".=.s e r .u . r --in :n= nydraulic centrol system wnich sense icss of cil pressure l cue to the opening c: :ne :ast acting scienoid valves or a u t failure in the hv.draulic centrol syntem piping. Two turoine firs: stage pressure switches for each tr.; system initiate aunc=atic j bypass of the turbine control valve fast ciceure scram when the first stace -ressure is below that recuired to produce 30% of e i ra ed.=cwer. Control valve closure time is approximately twice t as long as that for stop valve closure. i l Amendment I;o. X jy,79 __

i d l* 2.1 3AEES (Cent'd) L. References 1. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Sciling Water Reactor", NEDO 10802, February 1973. 2. " Qualification of the One-Dimensional Core Transient Model for 3ciling Water Reacters", NEDO 24154 and NEDE 24154-?, Volumes I, II, and III. 3. " Safety Evaluation for the General Electr.ic Tcpical Report Qualification of the One-Dimensional Core Transient Model for Sciling Water Reactors.NEDO-24154 and NEDE 24154-P, Volumes I, II, and III. d A endrent i'o. X 79, \\

P3APS Unit 3 1.2 BASES The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all cperating conditions and whenever there is irradiated fuel in the reactor vessel'. The pressure safety limit of 1325 psig as measured by the vessel steam space pressure indicator assures not exceeding 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value is derived from the design pressures of the reactor pressure vessel (1250 psig at 5750F) and coolant system piping (suction piping: 1148 psig at 562cF; discharge piping: 1326 psig at 562cF). The pressure safety li=it was chosen as the lower of - the pressure transients permitted by the applicaole design codes: ASME Sciler and Pressure Vessel Code, Section III for the pressar2 vessel and ANSI 331.1.0. for the reactor coolant system piping. The ASME 3ciler and Pressure Vessel Code permits pressure transients up to 1C% over design pressure (110% X 1250 = 1375 psig), and the ANSI cede permits pressure transients up to 20% over the design pressure (120% X 1148 = 1378 psig; 120% X 1326 = 1591.psig). A safety limit is applied to the Residual _ Heat Removal System (RERS) when it is cperating i.- the shundcwn

ccling cde.

A: this ire it is in:luded in :ne reac:c ccolan system. t .hendr,ent l:o.79,

i P3APS Unit 3 2.2 BASES REACTC9 COOLANT SYSTEM INTEGRITY The pressure relief system for each unit at the Peach Bottem Atcmic Power Station has been sited to meet two design bases. First, the total' capacity of the safety / relief valves and safety valves has been established to meet the overpressure protection. criteria of the ASME Code. Second, the districution of this required capacity between safety valves and relief valves has been set to meet desicn basis 4.4.4.1 of subsection 4.4 of the FSAR which states that the nuclear system safety / relief valves ~ shall prevent opening of the safety valves during normal plant .isolations and Icad rejections. The details of the analysis which shows cc=pliance with the ASME Code requirements are presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical Report submitted in Appendix K. bleven. safety / relief valves and two safety valves-have been installed on Peach Bottom Unit 3. The analysis of the worst overpressure transient, is prov'.ded in the Supplemental Reload L'icensiag Safety Evaluation and demonstrates r.argin to the code allowable overpressure limit of 1375.psig. The analysis of -he plant isciation transient is pr'Ovided in the l Supplemental Rel:ad Licensing Safe:. Evalua icn and dem:nstrates

.ac. he sate:y valves Qill not open.

The safety / relief valve settings satisfy the Code requirements

hat the icwest valve se poin: be at er below the vessel design prescure of 1:50 psig.

These se::ings are also sufficia.n:1y above :ne normal cperating pressure range :: prevent nnecessary cycling caused by r.iner transients. The design pressure Of the shundevn c0 clin; piping cf the Residual Heat Removal System is not exceeded with :he reactor vessel steam dome less than 75 psig. Amendment !;c. g j (, K K 79

PBAPS Unit 3 NOTES FOR TABLE 3.1.1 (Cont'd) 10. The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high. 11. An APRM will be censidered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement. 12. This equation will.be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), where: ~ FRP = fraction of rated thermal power (3293 MWt). M'FLPD = maximum fraction of limiting power density where the limiting pcwer density is 13.4 K'4/ft for all Sx8 fuel. The ratio of FRP to MFLPD shall be set' equal to 1.0 unless the actual operating value is le's than the design value of s 1.0, in which case the actual cperating value will be used. W= Leop Recirculation flew in percent of design. W is ~ 100 for core flow of 102.5 million Ib/hr or greater. ft W= the ' difference'between two loop and single loop effective recirculatien drive ficw rate at the same core flew. During single lecp cperatica, the reduction in trip setting (-0.66t_W) is acccmplished by ccrrecting the ficw input of the flew biased High Flux trip setting to preserve the original (two Icep) relatienship between APRM High Flux setpcint and recirculation drive ficwier by adjusting tb.e APRM Flux trip setting.~ W = 0 fer two Icep cperation. Trip level setting is in percent of rated pcwer (3293 MWt). 13. 5cc Section 2.1.A.1. i Amendment No. M /W, Z.R',79 y-- ---i, e-m y-w

o PBAPS Unit 3 h*CTES FOR TABLE 3.2.C 1. For the sta: tup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems.fer each function. The SRM and IRM blocks.need not be operable in "Run" mode, and the APRM and REM rod blocks need not be cperable in "Startup" mode. If the first column cannet be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally testad immediately and daily there'after; if this condition lasts longer than seven days, the system shall be tripped. If the first column cannot be met for both trip systems, the ~ systems shall be tripped. ,2. 'This equation will be used in the event of operation with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP) where: FRP = fraction of rated thermal power (3293 MWt) MFLPD = maximum fraction of limiting pcwer density where the limiting power density is 13.4 KW/ft for all 8x8 fuel. The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual cperating value will be used. ~ W= Lecp Recirculation flew in percent of design. W is - l'O for core ficw cf 10t.5 millica lb/hr er greater. Trip level se: ing is in percent of rated pcwer (3293 MW:). AW is the differerte between two leep and single locp l effective recirculatien drive ficw rate at the same core i flev. During single lecp cperaticn, the reductict in trip setting (-0.66A W; is acccmplished by ccrrecting tne ficw input of the ficw biased Ecd Eleck Monitor (R3M) 00 preserve the original (two lcep) relationship betwaa- "a se:pcin: and recirculation drive ficw, cr by adjusting the R2M setting. W = 0 for twc lecp cpera icn. 3. IRM downscale is bypassed when it is en its icwest range. l This function is bypassed when the count rate is 2 100 cps. 4. 5. One of the four SRM inputs may be bypassed. 6. This SRM function is bypassed when the IRM range switches l are en range 8 cr above. 7. The trip is bypassed when the'reacter power is S 30%. 3. This functicn is bypassed when the mcde switch is placed in 1 Run. l ( 74 - er2ndment J1o. E, M, EE, X,79 i I

PSAPS 3.2 3 ASIS (Cent'd) Pressure instrumentation is provided to close the main steam isolation valves in RUN Mode when the =ain steam line pressure drops below 850 psig. The Reactor Pressure vessel thermal transient due to an inadvertent opening of the turbine bypass i valves when not in the RUN Mode is less severe than the loss of f eedwater analyted in section 14.5 cf the FSAR, therefore, 1 closure of the Main Steam Isolation valves for thermal transient protection when not in RUN mode is not required. The HPCI high flow and temperature instrumentation are.provided to detect a break in the HPCI ateam piping. Tripping of this instrumentation results in actuation of HPCI isolation valves. , Tr'ipping logic for the high flow is a 1 out of 2 logic. Temperature is monitored at four (4) locations with four (4) temperature sensors at each location. Two (2) sensors at each location are powered by "A" DC control bus and two (2) by "3" DC control bus. Each pair of sensors, e.g., "A" or "3" at each location are physically separated and the tripping of either "A" cr "3" bus sensor will actuate HPCI isolation valves. The trip settings of S300% cf design flow for high flow and 200cF for h,igh temperature are such that core uncovery is prevented and fission w- - product release is within limits, The RCIC high flow and temperature instrumentation are arranged the same as that for.the HPCI. The trip setting of S300% for high flow and 200c? for temperature are based en the same criteria as the HPCI. The Reac::r Water Cleanup System hi;h flew and tempera:ure instrumentation are arranged similar to that fer ne HPCI. The trip settings are such that c:re uncovery is prevented and fissien product rele,ase is within limits. The instrumentaticn which initia es CSCS action is arranged in a dual bus system. As for other vital instrumenta icn arranged in this fasnion, the Specificati:n preserves the effectiveness Of the syster even during periods wnen maintenance cr testing is being performed. An exception to this is when logic functional testing is being performed. The control rod block functions are provided to prevent excessive control cod withdrawal so that MCPR does not decrease to the fuel cladding integrity safety limit. The trip logic for this function is 1 out of n: e.g., any trip on one of 6 APRM's, S IRM's, or 4 SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel recuirements fer the R3M may be reduced oy one for maintenance, testing, er calibration. This time period is only 3% of the operating time in a month and dces not significantly increase the risk of preventing an inadvertent 4 i control red withdrawal. ol-Amendnent, No. X M,I9-

p3APS 3.2 3ASES (Cont'd) The APRM rod block function is flow biased and prevents a significant reduction in MCPR, especially during operation at reduced flow. The APRM provides grcss core protection: i.e., limits the gress core power increase from withdrawal of centrol rods in the normal withdrawal sequence. The trips are set so that MCPR is maintained greater :han the fuel cladding integrity safety limit. The REM rod block function provides local protection of the core; ~ i.e., the prevention of boiling transition in the local region of the core, for a single red withdrawal error from a limiting ~ ,centrol rod pattern. The IRM red block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. A downscale indication on an APRM cr IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in the control red motion and thus, cc7 trol red motien is prevented. The downscale trips are set u. 2.5 indicated on scale. The flow cc=parator a'nd' scram discharge volume high level com;0nents have only one icgic channel anc are not recuired fcr safety. The flew comparater must be bypassed when operating with one recirculation water. pump. The refueling interlocks also ccerate one Icgic channel, and are' required for safety only when the mede switch is in the refueling ecsition. Por effective emergency core c cling for small pipe breaks, the HPC: system mus functicn since reac:Or pressure dces not l fecrease rapidly encugh to alicw either c:re spray er ~_PC: :o Operate in time. The aut:matic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spurious operation. The trip se: tings given in the specification are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument channel out of service. Two air ejector off-gas monitors are provided and when their trip poin is reached, cause an isolation of the air ejector off-gas line. Isolation is initiated when both instruments' reach their high trip point or one has an upscale ! Amendment No. X, M',79 t [

PSAPS Unit 3 ~ LIMITING CONDITIONS FOR OPERATION SURVIILLANCE REQUIREMENTS 3.3.3 Control Reds (Cont'd) 4.3.3 Control Rods (Cont'd) 4. Control rods shall not be 4. Prior to control rod with-withdrawn.for startup or drawal for startup or during refueling unless at least refueling, verify that at two source range channels least two source range channels have an ocserved count have an observed count rate rate ecual to or greater of at leas: three counts per than three counts per second. ~ second. '5. During operation with 5. When a limiting en. trol rod limiting control rod pat-pattern exists, an instru-terns, as determined by the m.ent functional test of the . designated cualified person-R3M shall be performed nel, either: prior to withdrawal cf the designated rod (s). a. Both RM3 channels shall be operable, or j - b. Control red withdrawal shall be blocked, 0: c. The operating power 1-evel shall.be limited so that the MCPR will remain above the fuel cladding integrity safe:y limi: assuming a single error that results in complete withdrawal of a single cperas;e control roc. t C. Scram Insertion Times C. S_ cram Insertien Times 1. The average scram inser-1. After each refueling cutage, l tien time, based on the and prior to synchrent:ing deenergi ation of the scram the main turbine genera:Or pilot valve solenoids as initially following restar: time zero, of all operacle of the plant, all operable centrol rods in the reactor fully withdrawn insecuence cower coeration conditien rods shall be scram time 'shall b'e no greater than: tested during operational hydrostatic testing or during startup from the fully withi drawn position with the nuclea: S. Inser:ed from Avg. Scram Inser-system pressure above 800 psig. Fullv Withdrawn tien Times (sec) 5 0.375 20 0.90 50 2.0 90 3.5 l i f thendment ita. X g,79 -103-

PBApS LIMITING CONDITIONS FOR OPERATION SURVIILLANCE REOUIREMENTS 3.3.C (Cont'd) 4.3.C (Cont'd) After exceeding 30 percent power all previously untested operable control rods shall be tested as described above prior to exceeding 40 percent power. 2. Whenever such scram time ~ measurements are made (such as when a scram' occurs and the scram insertion time recorders 2. The average of the scram are operable) an evaluation insertion times for the shall be made to provide three fastest control reasonable assurance that rods of all groups of four proper control rod drive control rods in a two-by-two. performance is being maintained array shall be no greater than: ~~ ~ % Inserted From Avg. Scram Inser-Fullv Withdrawn tion Times (Sec) 5 0.399 20 0.954 50 2.120

s.

...e

3. The maximum scram insertion time for 90% insertion of i

l any operable control rod snall net exceed 7.00 seconds. l i i Amendmnt No..'g79 ~104-l e m.

t-I e PBAPS Unit 3 3.3 and 4.3 BASIS (Cont'd) C. Scram Insertion Times The control red. system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to' prevent the.MCPR from becoming less than the fuel cladding integrity safety limit. Analy' sis of the limiting power transients shows that the negative reactivity rates resalting frcm the scram with the average rasponse to all drives as given in the above Specification, provide the required protection. Th'e numerical values assigned to the specified scram performance are based on the analysis of data frem other BWR's with control red drives the same as these en Peach Bottom. The occurrence of scram times within the limits, but significantly longer than the average, should be viewed as an indication of a systematic problem with control rod drives especially if the number of drives exhibiting such scram times. exceeds one control rod of a (5x5) twenty-five control array. In the analytical treatment of the transients, which are assumed to scram ca high neutron flux, 290 milliseconds are allowed between a neutron senser reaching the scram point and the start cf negative reactivity insertion. This is adequate and conservative when ccmpared :: :he typical time delay cf abcut-010 tilliseconds estimated frem scrar tes results. The 290 mill seconds used in :he analyses centis:s of 90 milliseconds fer senser and circuit delay and 200 milliseconds :0 star: cf centrol red =ction. In addition the centrol red drop accident has ceen analyced in.NIOC-10527 and its supplements 1 & 2 er the scram times given in Specifica icn 3.3.C. Surveillance requiremen: 4.3.C was originally written and used as a diagncstic surveillance technicue during pre-cperational and startup testing cf Cresden ; & 2 fer the early discovery and identification of significant changes in drive scram performance follewing major changes in plant cperation. The reason for the ac=licatien cf this surveillance was the unpredica able and d'e' graded scram cerformance of drives at Dresden 2. The cause of the slower scrah performances has been conclusively Amendmentflo..M,.#,.E,79 -111. m v v

4 PSAPS Unit 3 LIMITING CONDITICNS FOR OPERATICN-SURVIILLANCE REQUIREMENTS 3.5.I Averace Planar LEGR 4.5.I Averace Planar LHGR During power operation, the APLEGR The APLHGR for each type of fuel for each type of fuel as a function as a function of average planar of average planar expesure shall not exposure shall be checked daily i exceed the limiting value shown in during reacter operation at l the. applicable figures g2 5 S. rated ther=al power i during two recirculation Icep operatien. During single loep operation, the APLEGR for each fuel type shall.not exceed the above values multipled by tne following reduction factors: 0.71 for 7X7 fuel; 0.83 for 8X8 fuel; 0.81 for ?TA, 8XSR and PSX8R fuel. If at any time during operation it is deter-mined by normal surveillance that the limiting value of APLHGR is being ex-ceeded, action shall be initiated within ene (1) hour to restore APLHGR to within prescribed limits. If the APLEGR is not returned to within pre-scr'ibed limits within five (5) hours reacter power shall be decreased at a rate which veuld bring the reactor to the' cold shutdown condition within 36 hours unless APLEGR is returned to within limits during this =eriod. ~ ~ Surveillance and correspending action shcil continue until react.=r cperation is within the prescribed limits. 2.; J Local LHGR 4.5.J Local LHGR The LHOR as a function of core During pcwer cperatic.7,,,G.Cthe lg.near height shall be checked daily n=at generatten rate ta= c: during terctor operation at anv rce in any :uel assembly a: >-c5

.w*___1

,',y*,-' -'*'"g,-- .... a x..,.3 ,, c ;...:. s.". a ' ' ac excaad 4 design LEGR. LEGRSLHGRd LHCRd = Design LHGR 13.4 kW/ft for all 3x8 fuel. 1 l l i -133a-Amendment No.,E,,jer, yg, g 79

Eq; PBAPS Unit 3 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5.J Local LHGR (Cont'd) If-at any time during operation it al

is determined by normal surveillance that limiting value for LHGR.is being exceeded, action shall be initiated within one (1)

-hour to restore LHGR to within prescribed limits. If-the LHGR is not returned to within prescribed limits within.five (5) hours, reactor-power _shall be decreased at a rate which would bring the reactor to the cold shutdown condition within 36 hours unless LHGR is returned to within limits during'this period. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits. 3.5*.K.1 Minimum Critical Power 4.5.K Minimum Critical cower Ratio (MCPR) Ratio (MCPR)

1. During power operation, the MCPR 1.

MCPR shall ce enecked daily I for the appl'icable incremental during reactor power operation cycle core average exposure and at >25% rated thermal power. for each type of fuel shall-be 2. Except as provided in Specifica-i:: ecual te or greater than the value 3.5.K.3, the verification of given in Specification 3.5.K.2 or the applicability cf 3.5.K.2.a 3.5.K.3 times Kf, where Kf is as Operating Limit MCPR Values shown in Figure 3.5.1.E. If at shall be performed every~120 any time during operatinn it cperating days bv' scram time is determined by nordal testing 19 or mote control surveillance that tne limiting reds en a rotating basis and value for.MCPR is being performing the follovirg: exceeded, action shall be initiated within one (1) hour a. The average scram time to res ore MCPR to within prescribed t'o the 20% insertion ~ limits. If the MCPR is not position shall be: returned to within prescribed 7 ave s 77B limits within five (5) hours, reactor power-shall be decreased b. The average scram time at-a rate which would bring the to the 20% insertion reactor to the cold shutdown position is determined condition within 36 hours unless as follows: MCPR is returned to within limits during this period. Surveillance n .and corresponding action shall ) Pave = I Nili-continue until-reactor operation. i=1 is within the prescribed 3 limits. I Ni i=1 where: n = number of surveillance tests performed to date in the cycle. -133b- ....s_.... ... To

-- m PSAP5 LIMITING

CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS i

3.5.K. Minimum Critical Power 4.5.d. Minimum Critical Power l Rat io (MCPP) (Cont ' d) Ratio (MCPR) (Cont ' d) f 2. Except as specified in 3.5.K.3, Ni = number of active control the Operating Limit MCPR Values rods measured in the ith l are as followc: surveillance test. l a. If requirement 4.5.K.2.a is ) met: The Operating Limit MCPR values are as given in Table 3.5.K.2 i = average scram time to l the 20s insertion position of all rods measured in b. If requirement 4.5.K.2.a is not the ith surveillance test. met: The Operating Limit MCPR c. The adjusted analysis mean values as a function of r scram time (TS) is. calculated are as given in Figures as follows: N1 3.5.K.1 and 3.5.K.2. N1 ~/ 3=ju +1.65 n i V ]E[Ni i=1 '4here: Where: i i l ~/ = / a v e - 73 j/ = mean of the distribution ~ i 0.90 3 for average scram insertic time to the 20% position = 0.710 sec l l I 3. fhe Coeratinc Limit MCPR values Nt = tc:al number of active shall'be as given in Table 3.5.K.3 centrol rods.easured in 4

f the Surveillance Recuirement spec:fication 4.3.C.1 cf Section

.5.K.2 to scram time

est control rods is not

() = standard deviation of the performed distribution for average i scram insertion time to 4 the 20% positien = 0.053. -133c-a_- a---. u. 70

PBAPS Unit 3 Table 3.5.K.2 OPERATING LIMIT MCPR VALUES FOR VARIOUS CORE EXPOSURES

MCPR Operating Limit **

~.

Fuel Tvoe For Incremental Cvele Core Averace Excesure' l BOC to 2000 MWD /t 2000 MWD /t before EOC Before EOC To EOC 8x8 1.24 1.27 f PTA &P. 8XSR 1.25

1. 3 0 J

'2x8R 1.24 1.27 If requirement 4. 5.K. 2.a is met, i These valves shall be increased by 0.01 for single loop operation. i t i Amendment No. f 2,- EZ, 77,79 -123d-

PSAPS Unit 3 Table 3.5.K.3 OPERATING LIMIT.MCPR VALUES FOR VARIOUS CORI EXPOSURES

MCPR Operating Limit **

Fuel Tvoe For Incremental Cvele Core Averace Ex=osure BOC to 2000 MWD /t 2000 MWD /t before EOC Before IOC To EOC 8x8 1.33

1. 39 PTA &P 8XSR 1.36 1.42

-8x8R 1.33 1.39 7 . l t If surveillance recuirement 4.5.K.2 is not perf.ormed. 4 These valves shall be incretsed by 0.01 for single lcop operation. i I 1 1 j l l l s l t t 1

i l

Amendinent No. 79 -133e-l -o i i

fe. A _ l PBAPS Unit 3 3.5 BASES (Cont'd.) H. Enoineerino Safecuards comoartments Coolino and Vantilation One_ unit cooler in each pump compartment is capable of providing adequate ventilation flow and cooling. Engineeri'ng analyses indicate that the. temperature. rise in safeguards compartments without adequate ventilation flow or cooling is such that continued operation of the safeguards equipment or associated auxiliary equipment cannot be assured. Ventilarion associated with the High Pressure Service Water Pumps is also associated with the Emergency Service Water pumps, and is specified in Specification 3.9. I. Averace Planar LHGR This specification-assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR Part 50, Appendix K. ~ i The peak cladding temperature (PCT) following a postulated less-of-coolant accident is primartly a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent, secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to'or-less than the design LHGR. This LEGR times 1.02 is used in the heat-up code along with the exposure dependent steady state gap conductance and red-to-red local peakin-factors. The Technical Specification AFLHGR is the LHGR )f the highest powered rod divided by its local peaking factor. The-limiting value for APLEGR is as shown in the applicable figures for each fuel type. The calculational procedure used to establish the AOLMG? for each fuel type is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR'Part 50. A cceplete discussion of each code employed in.the analysis is presented in Reference 4. Input and model changes in the Peach Bottom loss-of-coolant analysis which are different from the previous analyses performed with Reference 4 are described in detail in Reference 8. These changes to the analysis include: (1) consideration of the' counter current flow limiting (CCFL) effect, (2) corrected code inputs, and (3) the effect of drilling alternate flow paths in the bundle lower tie plate. s Anendment lio~. 33,- H, g,79 -140- .~

T' V c. l PBAPS Unit 3 3.5.K 3ASES (Cont'd.) i The largest recuction in critical pcwer ratio is then'added to the' fuel cladding integrity safety limit MCPR to establish the MCPR Operating Limit for each fuel type. Two codes are'used-to analyze the rod withdrawal error transient. The first code simulates the three dimensional 3WR core nuclear and thermal-hydraulic characteristics. Using this code a limiting control rod pattern is determined.; the following assumptions are: included in this determination: (T) The core is operating at full power in the xenen-free condition. (2) The highest worth control rod is assumed to be fully ' inserted. (3) The analysis-is performed fer'the most reactive point in tha - rycle. k4) Th'e centrol rods are assumed to be the worst pessible pattern without exceeding thermal limits. (5) A bundle in the v.icinity of the highest worth control rod is assumed to be cperating at the maximum.allcwacle lir.ar heat generation rate. (6) A cundle in the-vicinity of the highes: werth centrol red is assumed to be cperating a the m'.nimum allcwacle critical power ratio. The three-dimensicndl EWR code then simulates the core response to tha centrol red withdrawal error. The second code calculates the Red Block Menitor respense to the red withdrawal errc: This cde simulates the Red 31cck Meni:cr under selecte( failure cond::icns (LPRM) for the core respense (calculated by the 3-dimensional 3WR simulation code) for the con:rcl rod withdrawal. The analysis cf the rod withdrawal error for Peach Bot:cm Unit 3 censiders the continucus withdrawal of the maximum worth control rod at its maximum drive speed from the reacter which is cperating with the limiting control red pattern as discussed above. .Anendment No. 33, 77, 33, 79 _ ; 4 3_

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n PBAPS Unit 3 '3.5.K 3ASES(Cont'd.) A brief summary of the analytical nethod used to determine the nuclear characteristics is given in Section 3 of Reference 7 Analysis of the abnc mal coerational transients is presented in ~ Section S.2 of Reference 7. (nput data and operating conditicns used in this analysis are shown in Table 5-8 ef Reference 7 and in-the Supplemental Relcad Licensing Analysis. L. Averace Planar LHGR (APLHGR), Local LMGR, and Minimum Critical Power Ratto (MCPR) In the event that the calculated value of APLEGR, LHGR cr MCPR exceeds its limiting value, a determination is made to ascertain the cause and initiate corrective action to restore the value to within prescribed limits. The status of all indicated limiting fuel bundles is reviewed as well as inget. data associated with! the limiting values such as power distr bution, instrumentation l . data (Traversing In-core Probe-TIP, LocC. Power Range. Monitor - LPRM, and reactor heat balance instrumtatation), control rod configuration, etc., in order to deter:4ine whether the calculated values are valid. In the event that the review indicates that the calculated value exceeding limits is valid, corrective action is immediately undertaken to restore the value :: within prescribed ~ limits. Fcil: wing corrective action, which ray involve alteratiens :: :he cent.ci red c:nfiguration and censecuently changes to the ::re pcwer distributien, revised instrumenta:ien data, including l changes to the relative neutr.a flux distribution for up to 43 l in: re locatiens is cctained and the power distributien, APLHGR, LEGR and.MCPR calculated. Corrective acticn is initiated wi:hin l One hour of an indicated value exceeding limits and verifica-icn l-tha: the indicated value is within prescribed limits is cbtained within five hcurs cf the initial indicati:n. In the event that the calcula:ed value of.2?LEGR, LEGR cr MCPR exceeding its limiting value is not valid, i.e., due to an e.rroneous instrumentation indicatien etc., :=rrective action is initiated within one hour of an indicated valua exceeding limits. Verification that the indicated value is within prescribed limits is cbtained within five hours of the initial indication. Such an ' invalid indication would not be a violation of the limiting condition for cperation and therefore would not constitute a reportable occurrence. 5 Amendment No. 33, py, pg, 37,79 -140c-

PBAPS Unit 3 3.5.L BASES (Cont'd.) Operating experience has demonstrated that a calculatid value of .APLHGR, LMGR or MCPR exceeding its limiting value predominately occurs 6.e to thir. latter cause. This experience coupled with the extremely unlikely occurrence of concurrent operation exc'eeding APLHGR, LEGR or MCPR and a Loss of Coolant Accident or applicable Abnormal Operational Iransients demonstrates that th.e times required to initiate corrective action (1 hour) and restore the calculated value of APLEGR, LHGR or MCPR to within prescribed limits (5 hours) are adequate. 3.5.M. References 1. " Fuel Densification Effects on General Electric Sciling Water Reactor Fuel", Supplements 6, 7, and 8 NEDM-10735, August 1973. 2. Supplement 1 to Technical' Report on Densifications of Ge.neral El.ectric Reactor Fuels, December 14, 1974 (Regulatory Staff). 3. Communication: V. A. Moore to I. S. Mitchell, " Modified GE Model for Fuel Densification", Docket 50-321, March 27, 1974. 4. General Electric Company Analytical Medel for Less-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix ~ K, NEDE-20566

Draft), August 1974.

5. General Electric Refill Reficed Calculation (Supplen.ent to SAFE Code Description) transmitted to the USAEC by let:er, G. L. Gyorey to Vioter Stello, Jr., dated December 20, 1974. m=-:.vD 7. General Electric Eciling Water Reac:cr Generic Relcad Fuel J e-. ,.,11 ,r .s,. .s..-.. a. 4,.,.. .cs s 2 m S. Lcss-of-Coolant Accident Analysis For Peach Bottom Atomic ?cuer Sta:icn Uni: 3, NEDO-2405;, Decemoer 1977. A eno, ent No. M. H, f 2, 62,79 -140d- ~ -~.y-

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i y ?BAPS Unit ' 3.6.A &'4~.6.A.-Bases (Cont'd) The vessel pressurization temperatures at any. time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Figure 3.6.1, 3.6.2, or 3.6.3 in conjunction with Figure 3.6.4. Note: Figure 3.6.3 includes an additional 400F margin required by 10 CFR 50 Appendix G. Neutron flux wires and samples of vessel material are-installed in the reactor vessel adjacent to the vessel wall at the. core mi,dplane level. The wires and samples will be removed and tested to experimentally verify the values used for Figure 3.6.4. As described in paragraph 4.2.5 of the Safety Analysis report, detailed stress analyses have been made on the reactor vessel for both steady state and transient conditions wi'.h respect to material fatigue. The results of.these transients are compared .to allowable stress limits. Requiring the coolant temperature in an idle recirculation loop to be within 50c? of the operating - Icop temperature before a recirculation pump is started assures. that the changes in coolant temperature at the reactor vessel l nczzles and bottom head region are acceptcole. l The plant safety analyses (Ref: NIDI-24011-?-A) state that all MSIV valve closure - Flux scram is the event which satisfies the l ASMI 3ciler and Pressure Code r0guirements for prc:ection frem
ne consequences of pressure in excess of the vessel design i
pressure. The reactor vessel pressure code limit of 1375 psig, given in Subsection 4.2 of the FSAR, is well a cve the peak pressure produced by the above overpressure event. ~ i I I r 8 Amendment No. M $2,79 -15:a-
A .PBAPS Unit 3 -3.6.D &.4.6.D BASES .Safetv and Relief Valves The safety / relief and safety valves are required to 'be operable above the pressure (122 psig) at which the core spray system is not designed to deliver full f. low. The pressure relief system for each unit at the Peach Bot' tom APS has been sized to meet two . design bases. First, the total capacity of the safety / relief and the safety valves'has been established to meet the ovedpressure '~ protection criteria of the ASME coda. Second, the distribution ~ of this required capacity between safety / relief valves and safety valves has been. set to meet design basis 4.4.4.1 of subsection ,.,4.4 of the FSAR which states that the nuclear system safety / relief valves shall prevent opening of the safety valves l during normal plant isolations and load rejections. The details of the analysis which,shows compliance with the ASNE l code requirements is presented in subsection 4.4 of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical - Report presented in Appendix K of the FSAR. Eleven safety / relief valves and two safety valves have been installed on Peach So.ttom Unit 3 with a total capacity of 79.51% of-rated steam flev. The analysis of the worst overpressure transient demonstrates margin to the code allowable overpressure limit of 1375 psig. To meet the power generation design basis, the total pressure relidi system capacity of 79.51% has been divided into 65.96% safety / relief (11 valves) and 13.55% saf~ety (2 valves). The analvsis of the clan isolation transient shows that the 11 safeiy/ relief valves limit pressure at the safety valves below the setting of the safety valves. Therefore, the safety valves l will not cpen. Experience in safety / relief and safety valve operation shows that l a testing of 50 per cent of the valves per year is adequate to detect-failure or deteriorations. The safety / relief and. safety . valves are benchtested every second i i Amendment No. 33, E5, 57, FE, #2,79 -157- . - ~ ..y
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- l r -2 PBAPS Unit 3 5.0 MAJOR DESIGN FIATURES L 5 ~.1 SITE TIATURES The site is located partly in Peach Ecttem Tcvnship, York County, [ partly in Drumore Township, Lancaster County, and partly in Fulton-. Township, Lancaster County, in southeastern Pennsylvania cn the westerly there of Cenew n c Pend at the =cuth of Rock Rdn Creek. It is about 33 miles north-northeast of Saltimore, ' ~ Maryland, and 63 miles west-southwes: of Philadelphia, Fennsylvania. Figures 2. 2.1 - through 2. 2. 4 of the FSAR show the l site location with respect to surrounding cc=munities.
5. 2 - REACTOR A.
The core shall consist of nc: =cre than 764 fuel asse=clies. 7 x 7 fuel assemblies shall contain 49 fuel reds and S x 3 ~fJel acsemblies shall contain 62 or 63 fuel reds. j3. One Pressurized Test Assembly may be inserted in the Core'for u!p to four full fuel cycles. t C. The reactor core shall contain 135 cruciform-shaped control L-
ds.
The control material shall be bcron carbide powder 03.C). compacted to approximately 705 of :he theoretical d ens i ty. O. Cne Fas: Scrs= C:n:rel Red Orive.sy ce utilized during Operation. I 5.3 RIACTCR VISSEL The reac ce vessel shall be as described in Tacle 4.2.2 cf :he ?SAR. The applicable design codes shall be as described in Table L 4.2.. of -he FSAR. 2.,. C m,. ,..,.u.. 5.. 6 A. The principal design parameters fer the primary containment shall be as given in Table 5.2.1 cf the FSAR. The applicable design codes shall be as described in Appendix M of the FSAR. l 3. The secondary containment shall be as described in Section 5.3 of the FSAR. C. Fenetratiens to One primary containment and piping pa< sing througn such penetrations shall be designed in accer# rce with standards se: forth in. Section '5. 2.3. 4 cf the : .it. 9 W -241-Amendment No.- 33, )J, pg,79 ma}}

ML20010J257

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1.1 BASES

2.1 3ASES

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ML20010J257

Text

Attachment:

1.1 BASES

2.1 3ASES

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