ML20127C652

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Proposed TS for License Amend Request 92-03 Re Reactor Vessel Integrity & RCS Heatup & Cooldown Limitation Curves
ML20127C652
Person / Time
Site: Zion  File:ZionSolutions icon.png
Issue date: 09/01/1992
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20127C650 List:
References
NUDOCS 9209100010
Download: ML20127C652 (47)
v • d • e


Text

- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _

'a ENCLOSURE 4 PROPOSED CHANGES MARKED *JP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FF " TTY OPERATING LICENSES JPR-39 AND DPR-48 FOR LICENSE AMENDMENT REQUEST NO. 92-03 920910001o 92o901-PDR ADOCK 05000295 P

PDR ZNLD/2110/26 j

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TABLE OF CONTENTS t.

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- uA %t y

- 1.0 DEFINITIONS.

r 1.26 Offsite Dose Calculation Manual Page 1 1.1

. Action 1.2.

Actuation Device 1.27 Operable - Operablitty l

1.3 Actuated Equipment 1.4 Actuation Logic Test Page 5 1.28 Operating i

1.5 Axial Flux 01fference 1.29 Operating Cycle 1.30 Operational Mode - Mode

-l 1.6 Batch Release 1.7' Channel Calibration, Instrument

(

1.31 Physics Tests l

1.32 Pressure Soundary Leakage-1.8 Chann 1 Check 1.33 Process Control Program (PCP)

Page 2 1.9 Channel Functional Test

'1.34 Protection Logic Channel 1.35 Protection System 1.10 Composite Sample 1.11 Containment Integrity 1.12 Continuous Release Page 6 1.36 Purge - Purging 1.13 Controlled Leakage 1.37 Quadrant Power Tilt Ratto-1 1.14 Core Alteration 1.38 Rated Thermal Power i

j ' Page 2a 1.14a Core Operating Limits Rc, ort 1.39 Reactor Pressure 1.40 Refueling Outage i

I

'Page 3 1.15 Defined Terms'-

1.41 Reportable Event-1.42 Shutdown Margin 1.16 Degree of Redundancy.

1.43 Site Soundary 1.17

. Dose Equivalent I-131

-1.18

-E Average _Olsintegration Energy 1.44 So11dification t'

1.19-Gaseous Radwaste Treatment. System

1.20 Ident1fted Leakage Page 6a 1.45 Source Check 1.46 Surveillance Frequency Notat1gn

[

Page 4 1.21.

Instrument Channel 1.47 Thermal Power 1.48 Unidentified Leakage 1.22 Leakage-1.23 Master Relay Test 1.49 Unrestricted Area-l 1.24 Member (s)'of the Pubile 1.50-Ventilation Enhaust Treatment System t

I t

1.25 Off-Site AC Pouer 1.51 Venting t

. Sources LIMITING SAFETY SYSTEM SETFOINT i

SAFETY LIMITS 2.1 Protective Instrumentation.Setp0lats 4

i 1.1 Reactor Core-2.2 Protective Equipment Setpoints

-[

j

-Bases "I

1.2 Reactor' Coolant Systen Pressure L -

____ _ ___ _ _ _ LL~agtmR D& M

Page

.. figure L ?,. ';. 2-1 Reaetcr C;;!&;;t-Syst:= " :te; L!-lt:t?m 84 AuQ a: A -

Reacter Ccel::t Sy:t. Coutdown-tiettations.

85 hFm.ie.b ' I ~..._._...

a

-....i -.r u a..._.s..

v.

r.....w,,

'Ts11-Te.;;r ' cv;c7 ttfu.~iiro for Lion unit 1 hich. l 87

c......__.__.,......s..,

. m,,. n..._r..... v.

4-,. 4--_.

.. c-

-Fen Twer- 0 rete: L!fe-4558VF for Ilon ^6 nit :

'3.3.6-1 Dose Equivalent I-131 RC Limit versus Percent of 124c

' Rated Thermal Power 131a Deleted 3.11-1 Restr nted Area Boundary 226 6.1-1

.Minimus Shift Crew Composttion' 327 4

5 a

4 f'

L LIST OF FIGURES (Continued)-

t i

yl1_

. Amen &Ent 1.os.'i e..." 123 a

m

...E

t Table EAnt-1.1 Operational Modes 6b 1.2 Survel11ance Frequency Notation 6c 3.1-1 Reactor Protection System-Limiting Operation Conditions and-Setpoints 30 3.1-2

' Deleted 33

% +,4 3r3re-1

-Zh; U.lt-b&Mter _Wssal Te;;haesi 0;ta 88 wte %4 -

JJ. 3. 2 2-Ziwe ;14-AeM4cr_VesssLTc ;t ::s Datt 89 3.3.3-1 Reactor Coolant Pressure Isolation Valves 97b I-3.3.4-1 In Service Inspection Program 106 3.3.5-1 Reactor Coolant Systemi and Chemistry Specifications 122 3.4-1 Engineered Ieguards' Actuation System-Limiting Conditions on 129

_ Operation anu Setpolhts 3.4-2' Deleted 132 3.7-1 Neutron Flux _High Trip Points _wlth Steam Generator Safety Valves 160a Inoperable-- Four, Loop Operatlons h

3.7-2' Deleted 160b 3.8.9-1

. Accident Monitoring Instrumentation 192a 3.9-3a-

' Deleted 205 l

3.9-3b Deleted 206 3.9-3c Deleted 207 3.9-3d Deleted' 207b i

1 e

LIST Of TABLES viiI

' Amendment Nos. L~' end 130

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

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'1.0 DEFINI'TIONS t.

1.28 OPERATING 1.33 PROCESS CONTROL PROGRAN (PCP)

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OPERATING is defined as perforetag the inhnded The PROCESS CONTROL PR63 RAM'(PCP) shall contain i

function in'the intended manner.

the current formulas, sampling, analyses, tests I

and determinatters to be made to ensure that ihe 1

1.29 OPERATING CYCLE

. processing and.puckaging of solid radioactive wastes will be accomp11shed in such a way as to l

The OPEKATING CYCLE shall be the interval between assure compliance with 10 CFR parts 20, 61 and

'l l-the endM = taajor refueltag outage and the end 11, and Federal and State regulations and other 1

j' of the n m W equent major refueling outage per requirements governing the shipment and disposal 1

unit.-

of radioactive weste.

t

[

1.30 OPERATIONAL NOK -- MODE '

1.34 f_ROTECTION LOGIC CHANNEL l

l An OPERATIONAL N00E (1.e. NOK ) shall ce respond A PROTECTION LOGIC CHANNEL M 1 be an i

to any one inclusive combinatten of core

. arrangement of relays, contacts or other L

reactivity condition, power level, and everage components which operate in response to reactor coolant temperature specified'in Table INSTRUNENT CHANNEL outputs to produce a decision L

1.1, when fuel' assemblies are present'in the output.' The decision output is the initiatten l

l reactor vessel.-

of a f.retective actian signal. At the system leve?, the decisten output is the operation of a 3

1 1.31 PHYSICS TESTS suff1 dent numer of ACTUATION KVICES and the associded ACTUATES EQUIPNENT as required to PHYSICS TESTS shall be those tests performed to place or restore the Nuclear Steam Supply System i

l measure the fundamental nuclear charesteristics of to a design safe state. The channel is deemed l

l' the reactor. cees and related instrumentatten and

- to include theiACTUATION KVICES.-

1) described in Chapter 14.0.of the FSAR, 2).

4 j

authortred.under.the previstens of 10 CFR 50.5g, 1.35 fROTECTION.515 TEM or 3) otherwise approved byL the Commission..

4 t

The PRfTECTION SYSTEM shall consist of both the-q 1.32 PRESSURE BOUNDARY LEAKAGL Reacter Protection System'and the Engineered Safeguards System. The PROTECTION SYSTEN

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l ls PRES $URE. BOUNDARY LEAKAGE shall'be. leakage (except encompasses all electric and mechanical devices

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steam generator. tuba' leakage) through'a' and c1rruitry (from sensors through ACTUATION I

t non-isolable fault in the Reactor Coolant Sysi s DEVICES) which are required to operate in order-

[

component body, pipe wall, or vessel wall..

to place or restore the Nuclear Steam Supply

.[

j.

~ System to a design safe state.

'QMm: L7 j

gu r t> en Amendment Nos. 96 and;86 l

09150/09!80 5

gg g j

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crt R h

da W+

spa d % ~.d.

N pc. < ;dy M4-rd Vcm. I und grosac _

b edwc li ~.: h,

inchl u) kedy

%d cooldown p

reies

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.cw emt. '

n=he veue.1 fl w es EOk

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LIMITING CONDITION F01 OPERATION SURVEILLANCE REQUIREMENT 3.3.2 PRESSURIZATION AND SYSTEM I1ITEGRITY 4.3.2 PRESSURIZATION AND SYSIEM ikIEGRITY A. Heatup and Cooldown A. The reactor coolant temp rature and pressure Q kg%

shall be determined to ts within the limits,at eA c The Reactor Coolant System + temperature and least once per 30 minutes during system f

(rales pressur}{with the exception of DRe) heatup, cooldown, and inservice leak and (pressur g shall be 14mited 1 ;;cerder.ce hydrostatic testing operations.

j with 16.. tt4es-shii ".. T igui c 3 3. 3. 2 -1 aie 3.3.2-23during heatup, cooldown and inservice leak and hydrostatic testing,w4+M

{, 4,

1. a. A maximum heatup rate of 20*F/hr applicable up to and including 180*F RCS indicated temperature. A maximum heatup.

rate of 60*F/hr applicable for RCS indicated temperatures greater than i

180'F.

-b.

x cooldown of 100*F in any 1 ud% %.

l M3 M

b yQ4{'0 )

c. A maximum temperature cha'nge of <10*F in

.any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period during inservice s (nag hya.ostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all tines.

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ACTION:

With any of the e4Mase limits / exceeded, restore the temperature and/or pressure to within the limli within 30 minutes; perforn an engineeting evaluation to determine the effects of the out of limit condition on the structurcl integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least MODE 3 within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and reduce RCS Tays and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

10580/10590 79 Amendment Nos. kH-4-H

4 i

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4 1.IM111NG-CONDITION FOR Ol' IRA 110N SURVEILLANCE REQUIREMENT 3.3.2 J(Continued) /Nguladt + '%,6 powG

.p (,.I s 4.3.2

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Ther.'-t" !SM 2.=.

in i b.pu o.s. o. I.....

.B.

Not Appiicable

.i.2.2 2 shall'be recalculated' periodically as required, based on results from the material surveillance program.

. C.

The secondary side of the steam generator must not be.pressurited above-200 psig if-C.

Not Applicable the. temperature of the primary and secondary i -

coolant is below.70*F.

D.-

The pressurizer heatup rate shall not exceed D.

Not Applicable 100*F/hr and the pressurizer cooldown rate not exceed 200*F/hr.

The spray shall not be i

used'if the-temperature difference between the pressurizer and'the spray fluid is -

greater than 320*F.

t E.1 Hydrostatic Testing E.

Not Applicable i.

1.

System inservice leak and hydrotests i

shall be performed in accordance with the requirements of ASME Boller and' Pressure Vessel Code.Section XI and applicable addenda; except as stated in Specification'4.3.4.C.1.

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LIMITING COWITION FOR OPERATION i

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SURVEltt. Anti REQUIREMENT L

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S.3.2.G.

Low Temperature Overpressure Protection 4.3.2.C.

Low Temperatwe Overpressere Protection (Continued) 1 (Continued) 4.

Verifying each PORV's isolation valve is open at least once per shift when this method is being used for low temperatwe overpresswe protection.

S.

Testing perssant to Specification 4.0.5.

j t

i 1

b.

The Reactor Coolant System (RCS) b.

The RCS pressure shall be vertfled to be l

l-pressere shall be less than 100 less than 100 psig, and presswizer level i

psig, and the pressertzer level shall be vertfled to be less then 25Y. at i

less than 251, or least once per shift, when this method is l

j l

being used for low temperatore i

l overpresswe protection.

l c.

The ACS is depressertzed and one c.

Verifying one PORV and St's isolatton PORV and it's isolation valve are valve are open at least once per shif t, open.

j when this method is being used for low temperatwo overpresswe protection l

s W>

D'd Q

2. a.. A maalaum of one* charging pump er 2.

At least WIof the Atme

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-[<;hw bq n";ty !;;;;tt =,.

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_ M x ";^, ;;;. U - ;

..T. and all i

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ce.pw cr,

Injectlen into the RCS. ame=eo 4 accumulators, shall be vertfled to be locapable accamed m.

of injecting lato the RCS prior to entering a b.Ne sa!d

.c e. he.s 4 shcu conditlen In_which they are reptrod_to l

, C,x

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  • b-.

( W a Weal N e w shift ecf %

cf W A:n Mo A ECS.

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thereafter while they are required to be Q bu

- c. Ne acomm t Aes

,s N U b_

y P:;;; E gg,+ g,g%

m g en

(

j eq*W ef Mj.6n a

h c.

l For short durations of tles during pump

\\ g/cg

.me.A:.s

(

l s dd switchover, two charging pumps any be:ApfemebE for

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m the purpose of maintainleg seal Injection flow to the reactor coolant pumps.

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(%ru s 7 Amendment Nos, 130'antM25

)

ii

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 303.2.G. Low Temperature. Overpressure Protection 4.3.2.G. Low Temperature Overpressure Protection (Continued)-

.(Continued)

3. ' When starting a reactor coolant pump, 3.

Not applicable.

- when no reactor coolant pumps.are

. running, the_ temperature in the secondary side.of the steam generator in the loop-in which the reactor coolant. pump is to be started shalI be less than :50*F higher than"the RCS temperature.

APPL.ICABILITY:' '~ Mode 4 when the temperature of any CS cold leg. is. less than or equal l

D.

to WJ0 4, M00E 5 and M00E 6'with the reactor vesseI head on.

ACTION:.

a.-

With one.PORY inoperable,-restore

the inoperable PORY to OPERA 8tE status within'7l days, or within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> either; Depressurize the RCS to less

.than 100 psig and-lower pressurizer level to.less then-l 25%, ' or,.

l' Depressurize the RCS ar.d open' I.

l- : at least one PORY and. it's '

l-block valve.-

12470/12480:

83a Amendme.t m N mew

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OPERATION -

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CRITICALI7Y LIMIT i

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BASED ON INSERVICE i

C HYDROSTATIC TIST*F) l TEMPERAnlRE (341 l

l FDR THE SERVICE PERIOD UP 70 'A E M t

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00 So too ISO 200 250 300 350 400 450 Soo l

l INDICATED TEKPERATURE G*D. F) l

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The 20'F/hr heatup rate is applicable for all RCS indicated temperatures.

The 60'F/hr heatup rate is applicable for RCS indicated temperatures greater than 180*F.

Zion 1 and 2 Reactor Coolant System Heatup Limitations applicable for up

to 14 EFPY and Heatup Rates up to 20*F/hr and 60*F/hr. Curves contain j

margins of 10*F and 60 psig for possible instrument errors.

REACTOR COOLANT SYSTEM HEATUP LIMITATIO:tS Figure 3.3.?-1

.-..-.-~.s 84 Amendment Nos.-l^1 f.

^ 1-7 0580/10590 0657A.

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50 too 150 200 250 300 350 400 450 Soo

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DOICATED TEMPERAWRE (Dir. F) i r

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i Zion 1 and 2 Reactor Coolant System Cooldown Limitations applicable for up to 14 EFPY and Cooldown Rates up to 100*F/hr.

Curves contain margins of 10*F and 60 psig f or possible instrument errors.

REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS

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Amendment Nos. iti-tr~9l-

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Amendment Nos.-4014-91 86 10580/10590-

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  • ION UNIT 2, g-lo iiiiiiiii,4.,,,,,iiiii,,,,,

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5 10 15 to 25 30 35 /

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87 M580/10590-065.7 A-

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89094-2 A5338 CL.1

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.12 49

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.32

. 56

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~ Girth WeM Seam L

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NF4(C)4 SAN

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0 d

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MF8(d).

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0

/

Girth Held Eeami (a). Estimated using Methoes of U.S.NRC NUREG-0800. Branch Technical Fosition NTES 5-2, July. 1981

/

(b) Neld Nire Heat No.-~72105 and Linde 60 Fivu Lot No. 8669.

/

j

-i.

(c) Neld M1re Heat No. 8T1762 and Linh 80 Flus Lot No. 8597

/

\\1 (d) Weld Wire Heat No. 8T1762 and Linde 80 Flux Lot No. 8632

/

~

. (e)~ Meld Mlre Heat No.'406L44 and Linde to Flux Lot No. 8720

/

\\(f)WeldMlreHeatNo.~71249andLindenoFluxLotMo.8738 h,

%gN-ZION UNIT 1 REACTOR VESSEL TOUGNNESS DATA x

TA8LE 3.3.2-88 Amendment Nos. IM=ent-tte e

=-

- =

p -, e

x

/

x m

Cu di P

T RTNOT J MS(a)

NOT 50FT-Ly</35 MIL

/

CCMPONENT HEAT NO.

MATERIAL TYPE m

in m

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(a) Estimated using Methods of U.S.NRC NUREG-0800, Branch Technical Position NTE8 5-2. July, 1981 i

i

\\

(b) Weld Mlre Heat No. 71249 and Linde 80 Flux Lot No. 8738

\\

(c) Held M1re Heat No. 72102 and Linde 80 Flux Lot No. 3650

\\

(d) Meld Wire Heat No. 72105 and Linde 80 Flux Lot No. 8669

\\

(e) Neld Mlre Heat No. 821T44 and Linde 80 Flux Lot No. 8773

/

/e ZION UNIT 2 REACTOR VESSEL TOUGHNESS DATA

'w_,

_ y E 3.3.2-2 89

____ N -rcn,. m

Bas Y

b 5

b

  • - 4

/,-

ww u-h 1 0 z. "

[

^

f 3.3.2 & 4.3.2 FRACTURE TOUGHNESS PROPERTIES The temperature and pressure changes during heatup and cm,idown are limited to be consistent with the requirements given in the ASME Boller and Pressure Vessel Code, Sertion III, Appendix G, an,QCFR50 Appendix G.

,A I M ai k.Q

\\

The f racture toughness properties of the ferritic materials in the reactor pessel are determined in accordance

\\

with the NRC Standard Review Plan. A12ftf;t85mM;-end-4*-eccordance=witbedditionsi=reactemessel requirements.

These properties are then evaluated in accordance with Appendix G of the/LSIfrSummer-Addendede Section III of the

\\

ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-MP4-A, " Basis-for-H*a tep-and 4Cwidowrr-timit-C es" Apr44-49M.

as',d)

+ M 2.

AU, yh d,)

jNg "gda) m Ms

^

,J Heatup anu co ldown imit curves are calculate using the most limiting value f thein11-ductilit reference tempera ture,(/RTNDr,)at the end of 15/ef fective' full power years (EFPY)f of service life.

The 15/ FPY service E

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-se. lect 4efFbf=5W1F8 dlimitingt RTNOT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Codt

% requirements.

N g_

(oa m a.

nsena m an *a um<r c.ans N N w L o%t& M M u*o rs idV o,n sal ~ d-H J

M the reactor vessel materials have been tested to determine their initial R NDT; the results of these/ tests are shown in 'dM 2.2.2 1 ad 0.0.2-2.

Reactor operation and resu)djusted rp,ference temperat r s b tant fas neutron (E greater than,3 Mev)

Irradiation can cause an increase in the RTNDT.

hercNO,a ue astad upon the fluence, copper content, and nickel content of the material in question,ican be predict using Regulatory Guide 1.99, Revision 2. " Effects of Residual Elements on Predicted Radiation Damage to Peattor Vessel Materials."

The heatup and cooldown limit curves cE ' b 6Hil ow 4.M-2 include F -dim adjustments for this shif t in s

RTHDT at the end of M EFPY as well as adjustments _for oqs_sible errors in the pressure and temperature sensing instruments.

QUwAh h Tc. p p g a ter Allowable pressure-temperature relationships for various hestup and cooldown rates are calculated using methods derived f rom Appendix G in Section III of the ASME Boiler and Prressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAF-Qg.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures, a sem1 elliptical surface defect with a depth of one-quarter of the wall thickness. T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities nt inservice inspection techniques. Therefore, the reactor operation limit curves develooed for this rc'erence crack are conservative and provide sufficient safety margins for protection against nonductile failurt To assure that the radiation embrittlement ef fects are accounted for in the calculation of the limit curves, the most limiting value of theqnil-ductility ref erence temperaturehRTNDT,)1s used,and-thh includes the$ radiation-induced shif tx (ARTHDTx)chrresponding to the end of the perfod for which heatup and cooldown curNes are generated, w _ mdw.

s bND ad Mt R.Lu 10580/10590 90 0657A Amendment lH-8-M-

+-

The ASME approach for calculating the allowable limit curves for various heatup and cooldosn. rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time i.- ring heatup or cooldown cannot be greater than the reference / stress intensity factor, Kg7 for the metal temperavsre at that time. KIR is obtained from the reference fracture toughness curve, defined n Appendix G to the ASME Code.

The KIR curve is given by the equation:

KIR = 2u.78 + 1.223 exp [0.0145(T-RINDT + 160)]

(1)

Where:

K g is the reference stress intensity factor as a function of the metal temperature ? and the metal i

nil-ductility reference temperature RT NDT.

Thus, the governing equation f or the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

CKgg+Kg[$K IR (2)

Where:

Kg=

the stress intensity factor caused by membrane (pressure) stress, i

Ky/I-thg stress intensity factor caused by the thernel gradients, g_hw

&ess %:%6%

KIR =

n::: t provided by the Code as a function of temperature relative to the RTNGT of the material.

C

' 0 for level A and B service limits, and C

1.5 for inservice hydrostatic and lea test operations.

=

fas@

At any time during the heatup and cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw,. the appropriate value forrRTNDT, and the reference f racture toughness curve. The thermal stresses resulting f rom temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K f, for the reference flaw is computed. From Equation (2) the pressure stress i

,_ intensity factors.are obtained and, from these, the allowable pressures are calculated.

,4mrba,',

' ' ' ' ' definerlimits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain-pr_ essure-temperature ranges.

The leak test limit curve shown on the heatup curvesgf. PTL th f.: :.) represent minimum temperature requirements at the leak test pressure specified by ASME Section III and the NRC Standard Review Plan NUREG-0800.

y Qh PrtD Allowable combinations of pressure and temperatureFtor specified temperature change rates are below and to the right of the limit lines shown.

interpolation.

Limit lines for cooldown rates between those presented may be obtained by 10580/10590 No.

91 0657A Amendment tet-&-tl l -

f REFERENCES 1.

ASME' 8011er and Pressure Vessel Code Section III,1976 Summer Addenda.

e 2.

WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves", April 1975.

3.

ASME 8011er and Pressure Vessel Code,Section III. N-331.

4.

ASME Boller and Pressure Vessel Code,Section III, N-415.

5.

FSAR, Chapter 4.3.

6.

WCAP-8724, "ASME III, Appendix G Analysis of the Commonwealth Edison Company Zion Unit 1 Reactor Vessel".

7.

WCAP-8727 "ASME III, Appendix G Analysis of the Commonwealth Edison Company Zion Unit 2 Reactor Vessel".

8.

WCAP-10677, " Adjoint Flux Program for Zion Units 1 and 2".

9.

Regulatory Guide 1.99 Revision 2.

e 10.

Code of Federal Regulations,'10CFR50 Appendix h, " Fracture Toughness Requirements."

r.4 LOG 11.

WCAP-1494+, "Heatup and Cooldown Limit Curves for 05:hentop;.r.:He~

f.
  • S C:..
M. CLvii b.., Zion Dnits 1 and 2 Rese4er-

"Z -

Vettel",Ji ev. E 12.

WCAP-109 ', " Zion Units 1 a'nd 2 Reactor Vessel Fluence and RTPTS Evaluations",T<<- L i,]%o.

L 13.

CWE-86-563, " Low Temperature Overpressure Protection System Setpoing Analysis", August f',.1986.

6

-14.

CWE-865-588, Low Temperature Overpressure Protection System Setpoing Analysis Extens16.s" October 27, 1986.

4 w.

I 10580/10590 93a Amendment 10! " ??

0657A

}

Bas:s:

Low Temperature Overpressure Protection 9

/

320*F 3 7".2.G There are 3 mear. of pro 6ecting the RCS from overpressurization by a pressure transient at low temperatures (below E50*PI.

The first type of protection is ensured by the operation and surveillance of the power 4.fl.2.G operated relief valves with a lif t setting of 435 psig. A single power operated relief valve (PORV) will relieve a pressure transient caused by 1) a mass addition into a solid RCS from a charging pump or 2) a heat input based on a reactor coolant pump being started in an idle RCS and circulating water into a steam generator whose temperature is 50'F greater than the RCS temperature.

(1)

The second means of protection is ensured by a PORV being open.

It will have the same relieving capabilities as mentioned above.

The third means of protection limits the pressurizer level to 25% and the pressur*zer pressure to 100 psig.

A pressure transient caused by the inadvertent mass addition from a charging pump running for 10 minutes will be relieved by the large gas volume and low pressure present in the pressurizer as mentioned above.

Maintaining the pressurizer level below 35% will also make the hi pressurizer level deviation alarm available to the operator during a mass addition accident.

g In the event that a single PORV becomes inoperable, the repair period of 7 days is based on allowing sufficient time to effect repairs using safe and proper procedures and upon the operability of the redundant PORV.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period to reach tne restrictive conditions in the pressurizer provides suf ficient time to meet these conditions.

In the event that both PORV's btcome inoperable, the condition is more serious than for a single inoperable PORV, therefore every attempt should be made to depressurize the RCS in a controlled manner as rapidly as possible.

The 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> time period to reach the restrictive conditions in the pressurizer represents a reasonable amount of time to meet these conditions under an expedited circumstance.

The Low Temperature Overpressure Protection System must be tested' on a periodic bases consistent with the need for its use.

system during cooldown and startup.A CHANNEL FUNCTIONAL TEST shall be perforned prior to enabling the overpress The limitations and surveillance requirements on the ECCS equipment provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or the limiting conditions placed on the pressurizer.

The restrictions for startup of a RCP limits the heat input accident to within the relieving capabilities of a single PORV.

(1)

Pressure Mitigating Systems Transient Analysis Results July 1977 Westinghouse Owners Group on RCS Overpressurization.

712470/12470 94 Ameeiment Nos 110 0 00

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l

4.3.4.0.

Materials Irradiation Surveillance Specimen 3.3.4 Inspection. (per unti) i Specimen cap 10-5: ; ed 5 -the-reactor a

.t ve'-- m = = = ~ e,= m'+

i su s j,,,_p,m,_%

witMee= t *:9 th; ref uel hg pe -h

-e* G, -

c 6 3.s _.

a g, mt i---:d htely ;rc;ec h; er f;11 1%_the

,j E4 feet-tve-Fttil Tm. Veers =1Em} af ;,r.R e

I 11ie=ss1feMows -

f"f" ' *5

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g io cra, so ArredN CASSUlE MII".0"?".E SC"E00LE i

I.RA. R T. T 1

t

/

CAP.SULE CAPSULE

/

DESIGNATION REMOVAL TI - /(EfPV)

T' REMOVED

.16)

U REMOVED 3.52)

X REMOV (5.17)

[

Y 8.5 Z

32 t

'WSV STANDBY 4

UH1\\2 l

CAPSULE APSULE i

DESIGNAT REMMALTIME(EFPY)

I U

REMOVED (1.27)

REMOVE (3.56) i Y

8.5 K

13 f

W.S,V,Z STANDBY i

l 105 Amendment Mos. ;^e - ^4 11730/11740 R... _

Basis

^

4.3.4 The surveillance inspection program has been developed to comply with Section XI of the ASME Boller and Pressure Vessel Code and applicable addenda as required by 10CFR50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10CFR50, Section 50.55a(g)(6)(1). The design of the plant, state of non-destructive testing technology, and access to areas to bc inspected require such relief.

The Reactor Vessel Material Surveillance Program is designed to evaluate the effects of radiation on the fracture toughness of reactor vessel steel based..on the transition temperature approach and the fracture mechanics approach.

10CFR50, Appendix H, paragraph II B.1 requires that the reactor ressel material surveillance program shall meet the requirements of ASTM E185-82 such that the surveillance capsules represent end-of-life fluences at the reactor vessel surface and 1/4T wall thicknesses.

.r;,i__,_;;;;w';;

c.s cc..a.cd

.,4:r * : d;;r 9;;. 00-

-1.J-03.

1 1

I 1,19 Amendment Nos. 100 0 ^;-

11730/11740

- L n344.

8-

~

pp 364 b

=v 4

6.6.1.G PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 4j.

The reactor coolant system pressure and temperatu e limits. including heatup and cooldown rates, criticality, and hydrostatic and t

C-leak test limits, shall be estab11shed and 1

documented in the PTLR. The reactor vessel pressure and temperature limits and the heatup and cooldown rates are addressed in

?

Specification 3.3.2.A.

The analytical methods used to determine the pressure and temperature limits including the heatup and cooldown rates t

shall be those previously reviewed and approved by the NRC as described in NCAP-13406, "Heatup and Cooldown Limit Curves for Normal Operation for Zion Units 1 & 2", dated JulyL1992 and n

approved by the NRC SER dated'_

The reactor. vessel pressure and temperature limits, including those for heatup and cooldown rates, shall be determinedso that all applicable limits (e.g., heatup limits, cooldown limits,'and inservice leak and hydrostatic testing limits) of L he analysis are met.

The PTLR, including t

revisions or supplements thereto, shall be provided upon issuansa for each reactor vessel

' fluency period.

1' r

k.

.c y

y 3

'6m 1

--ia-

,9 w

=

q s 'a y a

9

,y,

.g g

r

-w._,a

,. _y

. -. ~... -

. _. =

4 5

d ENCLOSURE 4 i

PROPOSED CHANGES TYPED PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES 1

1 DPR-39 AND DPR-48 FOR LICENSE AMENDMENT REQUEST NO. 92-03 PAGES.J10DIflED I

1 86 vii 87 viti 88 5

89 79 90 80 91 83

.93a 83a 94 84 105 85 119 EAGES__RELEIED None l

l PAGES AQDEQ 1

Sa 1

316b i

l l

'ZNLD/2110/27

TAL OF CONTENTS 1.0 DEFINITIONS Page i 1.1 Action 1.26 Offsite Dose Calculation Manual I

1.2 Actuation Device 1.27 Operable - Pperat?1i ty 1.3

Actuated Equipment 1.4 Actuation Logic Test Page 5 1.28 Operating 1.5 Axial Flux Difference 1.29 9peraicng Cycle 1.6-Batch Release 1.30 Operationil Hede - Mode 1.7 Channel Ca11bration Instrument 1.31 Physics Tests 1.8 Channel Check 1.31a Pressure and Temperature Limits Report l

}

1.32 Pressure Boundary Letkage l

Page 2 1.9 Channel functional Test 1.33 Process Control Program (PCP) 1.10 Composite Sample 1.34 Protection Logic Channel s

1.11 Containment Integrity Page 5a 1.35 Protection System l

1.12 Continuous Release 1.13 Controlled Leakage Page 6 1.35 Purge - Purging 1.14 Core Alteration 1.37 Quardrant Power Tilt Ratio Page 2a 1.14a Core Onerating Limits Report 1.38 Rated Thermal Power 1.39 Reacter Pressure Page 3 1.15 Defined Terms 1.40 Pefueling Octage 1.16 Degree of Redundancy 1.41 Reportable Event 4

1.17 Dose Equivalent I-131 1.42 Shutdown Margin 1.18 E - Average Disintegration Energy T.43 Site Boundary 1.19 Gaseous Radwaste Treatment System 1.44 Solidification 4

1.20 Identified Leakage Page 6a 1.45 Source Check Page 4 1.21 Instrument Channel 1.45 Surveillance Frequency Notation 1.22 Leakage 1.47 Thermal Power 4

1.23 Master Relay Test 1.48 Unidentified Leakage 1

1.24 Member (s) of the Public-1.49 Unrestricted Area 1.25 Off-Site'AC Power 1.50 Ventilation Exhaust Treatment System Sources 1.51 Venting SAFETY LIMITS LIMITING SAFETY SYSTEM SETPOINT 1.1 Reactor Core 2.1 Protective Instrumentation Setpoints Bases 2.2 Protective Equipment Setpoints l

1.2 Reactor Coolant System Pressure Bases ZOSR-242 i

Amendment Nos.

i

F1aur _

Eage Deleted 84 Deleted 85 Deletad 86 Deleted 87 3.3.5-1 rase Equivalent I-131 RC Limit versus Percent of Rated Thermal Power 124c Deleted 131a 3.11-1 Restricted Area Boundary 226 6.1 1 Minimum Shift Crew Cm position 327 LIST OF FIGURES (Continued)

ZOSR-242 v11 Amend.nent Nos.

Iable Eage 1.1 Operational Modes-6b 1.2 Surveillance Frequency Notation 6c j

3.1-1 Reactor Protection System-Limiting Operation Conditions and Setpoints 30 3.1-2 Deleted 33 Deleted 68 Deleted 89 3.3.3-1 Reactor Coolant Pressure Isolation Valves 97b 3.3.4-1 In Service Inspection Program 106 l

3 3.5-!

Reactor Coolant Systems and Chemistry Specifications 122 3.4-1 Engineered Safeguards Actuation System-Limiting conditions on 129 Operation and Setpoints 3.4-2

' Deleted 132 3.7-1 Neutron Flux High Trip Points with Steam' Generator Safety Valves 160a 4

Inoperable - Forr loop Operations 3.7-2 Deleted 160b 3.8.9-1 Accident Monitor'ang Instrumentation 192a 3.9-3a Deleted 205 3.9-3b Deleted 206 3.9-3c Deleted 207 3.9-3d Deleted 207b LIST OF TABLES-ZOSR-242 vill Amendment Nos.

1.0 DEFINITIONS 1.28 QEERATING 1.32 ERESSURE BOUNDARY LEAKAGE OPERATING is defined as performing the intended PRESSURE BOUNDARY LEAKAGE shall be leakage function in the intended manner.

(except steam generator tube leakage) through a non-isolable fault in the Reactor Coolant System 1.29 OPERATING CYCLE component body, pipe wall, or vessel wall.

The OPERATING CYCLE shall be the interval between 1.33 ERQCESS CONTROL PROGRAM (PCEl the end of one major refueling outage and the end of the next subsequent major refueling outage per The PROCESS CONTROL PROGRAM (PCP) shall contain i

unit.

the current formulas, sampling, analyses, tests and determinations to be made to eneure that the l.30 OPERATIONAL MODE - M0QE processing and packaging of solid radioactive wastes will be accomplished in such a way as to An OPERATIONAL MODE (i.e. MODE) shall correspond assure compliance with 10 CFR parts 20, 61 and to any one inclusive combination of core 71, and Federal and State regulations and other reactivity condition, power level, and average requirements governing the shipment and disposal reactor coolant temperature specified in Table of radioactive waste.

1.1, when fuel assemblies are present in the reactor vessel.

1.34 PROTECTION LOGIC CHANNEL 1.31 PHYSICS TESTS A PROTECTION LOGIC CHANNEL shall be an arrangement of relays, contacts or other PHYSICS TESTS shall be those tests performed to components which operate in response to measure the fundamental nuclear characteristics INSTRUMENT CHANNEL outputs to produce a decision of the reactor. core and related instrumentation output.

The decision output is the initiation of 1l l and 1) described in Chapter 14.0 of the UFSAR, 2) a protective action signal. Ac the system level, authorized under the provisions of 10 CFR 50.59, the decision output is the operation of a or 3)'otherwise approved by the Commission.

sufficient numer of ACTUATION DEVICES and the associated ACTUATED EQUIPMENT as required to 1.31a PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) place or restore the Nuclear Steam Supply System to a design safe state. The channel is deemed to The PTLR is the unit specific document that include the ACTUATION DEVICES.

provides the' reactor vessel pressure and temperature limits, including haitup and cooldown rates, for the current reactor vessel fluence period in accordance with Specification 6.6.1.G.

Plant operation within these' operating limits is t

addressed in Specification 3.3.2.A.

ZOSR-242 5

Amendment Nos.

~

r 1.0 DEFINITIONS 1.3S PROTECTION SYSTEM 1

The PROTECTION SYSTEM shall consist of both the Reactor Protection System and the Engineered Safeguarus System. The PROTECTION SYSTEM encompasses-all electric and mechanical devices and circuitrv (from sensors through ACTUATION

. DEVICES) which are required to operate in order to place or restore the Nuclear Steam Supply System to a design safe state.

i I

'ZOSR-242 Sa Amendment Nos.

F LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2 PRESSURIZATION AND SYSTEM INTEGRITY 4.3.2 PRESSURIZATION AND SYSTEM INTEGRITY A. Heatup and'Cooldown A. The reactor coolant temperature and pressure shall be determined to be within the limits The Reactor Cool.nt System (with the exception specified in the PTLR at least onct per 30 l

of the pressurizer) temperature and pressure minutes during system heatup, cooit wn, and and heatup and cooldown rates shall be inservice. leak and hydrostatic testing maintained within the limits specified in the operations.

Pressure and Temperature Limits Report (PTLR) during heatup, cooldown and inservice leak and j

hydrostatic testing.

' APPLICABILITY: At all times.

ACTION:

With any of the limits specified in the PTLR exceeded, restore the temperature and/or pressure to within the 11rdit within 30 minutes; perform an engineering. evaluation to determine the j

effects of the out of limit. condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation er be in'a.least MODE 3 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 RCS TAVG and pressure to less than 200*F and SM psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

ZOSR-242 79 Amendment Nos

LIMITING CONDITION FOR OPERATION-SURVEILLANCE REQUIREMENT 3.3.2 (Continued) 4-3.2

' B.

The Reactor Coolant System pressure and B.

Not Applicable temperature limits specified in the PTLR shall be recalculated periodically as required, based on results from the material surveillance program.

C.

The secondary side of the steam generator C.

Not Appilcable must not be pressurized'above 200 psig if th?

temperature of the primary and secondary coolant is below 70*F.

D.

The pressurtzer heatup rate shall not exceed D.

Not Applicable 100*F/hr and the pressurizer cooldown rate not exceed 200*F/hr. The spray shall not be used if the temperature difference between 1

the pressurizer and the spray fluid is greater than 320*F.

E.

Hydrostatic Testing E.

Not Appilcable 1.

System inservice leak and hydrotests shall'be performed in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI, and applicable i.

addenda; except'as stated in

~

i Specification 4.3.4 C.1.

J ZOSR-242 80 Amendment Nos.

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

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2.G.

Low Temperature Overpressure Protection 4.3.2.G.

Low Temperature Overpressure Protection (Continued)

(Continued) 4.

Verifying each PORV's isolation valve is open at least once per shift when this method is being used for low temperature overpressure protection.

5.

Testing pursuant to Specification 4.0.5.

b.

The Reactor Coolant System (RCS; b.

The RCS pressure shall be verified to be i

pressure shall be less than 100 less than 100 psig, and pressurizer level i

psig, and the pressurizer level shall be verified to be less than 25% at less than 251, or least once per shift, when this method is being used for low temperature overpressure protection.

c.

The RCS is depressurized and one c.

Verifying one PORV and it's isolation PORV and it's isolation valve are valve are open at least once per shift, l

open.

when this method is belag used for low temperature overpressure protection.

j i

1 2.

a.

A maximum of one* charging pump, 2.

At least two of the three charging pumps, and j'

shall be capable of injection into all safety injection pumps and accumulators, the RCS.

shall be verified to be incapable of injecting j

into the RCS prior to entering a condition in i

b.

No safety injection pumps shall be which they are required to be incapable of capable of injection into the RCS.

injection into the RCS, and at least once per shift thereafter while they are required to be c..

No accumulators shall be capable of incapable of injection into the RCS.

l injection into the RCS.

For short durations of time during pump switchover, two charging pumps may be capable of

. injection into the RCS for the purpose of l

i maintafning seal injection flow to the reactor coolant pumps.

I i

ZOSR-242 83 Amendment Nos.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.3.2.G.

Low Temperature Overpressure Protection 4.3.2.G.

Low Temperature Overpressure Protection (Continued)

(Continued)

-3.

When starting a reactor coolant pump, 3.

Not applicable.

when no reactor coolant pumps are running, the temperature in the secondary side of the steam generator in the loop in which the reactor coolant pump is to be started shall be less than 50*F higher than the RCS temperature.

APPLICABILITY:

Mode 4 when the temperature of any RCS cold leg is less than or equal l

to 320*F, MODE 5 and H00E 6 with the reactor vessel head on.

ACIIQN:

a.

With one PORV inoperable, restore j

l the inoperable PORV to OPERABLE status within 7 days, or within the next 24 hou 5 either I

)

Depressurize the RCS to less than.100 psig and lower pressurizer: level to less than 251, or Depressurize the RCS and open at least one PORV and it's block valve.

ZOSR-242 83a Amendment Nos.

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ZOSR-242 85 Amendment Nos.

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OSR-242 86 '-

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4 ZOSR-242 87 Amendment Nos.

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ZOSR-242 89 Amendment Nos.

Bases 3.3.2 & 4.3.2 FRACTURE TOUGHNESS PROPERTIES The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G, and 10CFR50 Appendix G.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan.

These properties are then evaluated in accordance with Appendix G of the 1986 Edition of Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in HCAP-13406, "Heatup and Cooldown Limit Curves for Normal Operation for Zion Units 1 & 2", July 1992.

Heatup and cooldown limit curves are calculated using the most limiting value of the adjusted nil-ductility reference temperature (adjusted RTNDT) at the end of 14 effective full power years (EFPY) of service life.

The 14 EFPY service life period is. chosen such that the limiting adjusted RTNDT at the 1/4T location in the core region is greater than the RTNDT of the limiting unirradiated material.

This limiting adjusted RTNDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are I

shown in WCAP-13406 "Heatup and Cooldown Limit Curves for Normal Operation for Zion Units 1 & 2".

Reactor operation and resultant fast neutron (E greater than 1 MeV) Irradiation can cause an increase in the RTNQT-Adjusted reference temperatures, based upon the fluence, copper content, and nickel content of the material in l

question or based on credible surveillance data, can be calculated using Regulatory Guide 1.99, Revision 2,

" Effects of Residual Elements on' Predicted Radiation _ Damage to Reactor Vessel Materials." The heatup and cooldown limit curves specified in the Pressure and Temperature Limits Report (PTLR) include adjustments for this shift in RTNDT at the end of 14 EFPY as well as adjustments for possible errors in the pressure and temperature. sensing instruments.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 l

CFR Part 50, and these methods are discussed in detail in HCAP-13406.

The general method for calculating-heatup and cooldown limit curves is based upon the principles of the ' linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures, a semielliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME.Section.III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most lim

  • ting value of the adjusted nil-ductility reference temperature (adjusted RTNDT) is used, which includes the initial RTNDT.

the radiation-induced shift (ARTNDT) corresponding to the end of the period for which heatup and cooldown curves are generated, and a margin factor.

ZOSR-242 90 Amendment Nos.

i

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and pressure stresses at any time during heatup or 1

l cooldown cannot be greater than the reference stress intensity factor, KIR for the metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The KIR curve is.given by the equation:

KIR = 26.78 & l.223 exp [0.0145(T-RTNDT + 160)]

(1)

Where: Kyg is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperatiire RTNDT.

Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

l (2)

CKg+ KIT IKIR 1

Where:

KIM =

the stress intensity factor caused by membrane (pressure) stress, IT -

the stress intensity factor caused by the thermal gradients, K

IR =

reference stress intensity factor provided by the Code as a function of temperature relative K

to the RTNDT of the material, 4

2.0 for level A and-B service limits, and C

1.5 for inservice hydrostatic and leak test operations.

C At any time during the heatup and cooldown transient, KIR is determined by the metal temperature at the tip of the.

I postulated flaw, the appropriate value for adjusted RTNDT, and the reference fracture toughness curve.

The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the I

corresponding thermal stress intensity factor, KIT, for the reference flaw is computed.

From Equation (2) the pressure-stress intensity factors are obtained and, from these, the allowable pressures are calculated.

I The PTLR defines limits to assure prevention of non-ductile failure only.

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

I The leak test limit curve shown on the heatup curves in the PTLR represent minimum temperature requirements at the leak test pressure specified by ASME Section III and the NRC Standard Review Plan'NUREG-0800.

I Allowable combinations of pressure and temperature in the PTLR for specified temperature change rates are below and to the right of the limit lines shown. ' Limit lines for cooldown rates between those presented may be obtained by interpolation.

ZOSR-242 91 Amendment Nos.

REFERENCES 1.

'ASME Bo11er'and Pressure Vessel Code,Section III, 1976 Summer Addenda.

2.

WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves", April 1975.

3.

ASME Boller and Pressure Vessel Code,Section III, N-331.

4.

ASME Boiler and Pressure Vessel Code,Section III, N-415.

5.

FSAR, Chapter 4.3.

6.

HCAP-8724, "ASME III. Appendix G Analysis of the Commonwealth Edison Company Zion Unit 1 Reactor Vessel".

7.

WCAP-8727, "ASME III, Appendix G Analysis of the Commonwealth Edison Company Zion Unit 2 Reactor Vessel".

8.

NCAP-10677, " Adjoint Flux Program for Zion Units 1 and 2".

9.

Regulatory Guide 1.99 Revision 2.

10. Code of Federal Regulations, 10CFR50 Appendix G, " Fracture Toughness Requirements."

11.

HCAP-13406, "Heatup and Cooldown Limit Curves for Normal Operation for Zion Units 1 & 2".. July 1992.

12. WCAP-10962, Rev. 2, " Zion Units.1 and 2 Reactor Vessel Fluence and RTPTS Evaluations",. December'1990.
13. CHE-86-563, " Low Temperature Overpressure Protection System Setpoint Analysis", August 25, 1986.

l 14.

CHE-865-588, " Low Temperature Overpressure Protection System Setpoint Analysis Extension", October 27., 1986.

1 1

e 4

i

}

ZOSR-242 93a Amendment Nos.

Bases:

Loe Temperature Overpressure Protection 3.3.2.G There are 3 means of protecting the RCS from overpressurization by a pressure transient at low temperatures (below 320*F). The first type of protection is ensured by the operation and surveillance of the power 4.3.2.G operated relief valves'with a lift setting of 435 psig. A single power operated relief valve (PORV) will relieve a pressure transient caused by 1) a mass addition into a solid RCS from a charging pump or 2) a heat input based on a reactor coolant pump being started in an idle RCS and circulating water into a steam e

generator whose temperature is 50*F greater than the RCS temperature.

(1) i The second means of protection is ensured by a PORV being open.

It will have the same relieving capabilities as mentioned above.

The third means of protection limits the pressurizer level to 25% and the pressurizer pressure to 100 psig.

A pressure transient caused by the inadvertent mass addition from a charging pump running for 10 minutes will be relieved by the large gas volume and low pressure present in the pressurizer as mentioned above.

l Maintaining the pressurizer level below 25% will also make the high pressurizer level deviation alarm available to the operator during a mass addition accident.

3 In the event that a single PORV becomes inoperable, the repair period of 7 days is based on allowing sufficient time to effect repairs using safe and proper procedures and upon the operability of the redundant PORV. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period to reach the restrictive conditions in the pressurtzer provides sufficient time te meet these conditions.

i In the event that both PORV's become inoperable, the condition is more serious than for a single inoperable PORV, therefore every attempt should be made to depressurize the RCS in a controlled manner as rapidly as possible. The 16-hour time period to reach the restrictive conditions in the pressurizer represents a reasonable amount of time to meet these conditions under an expedited circumstance.

The Low' Temperature Overpressure Protection System must be tested on a periodic bases consistent with the need for its use. A CHANNEL FUNCTIONAL TEST shall be performed prior to enabling the overpressure protection system during cooldown and startup.

The limitations and surveillance requirements on the ECCS equipment provides assurance that a mass addition pressure transient can te relieved by the operation of a single PORV o" the limiting conditions placed on the pressurizer.

The restrictions for startup of a RCP limits the heat input accident to within the relieving capabilities of a single PORV.

(1)

Pressure Mitigating Systems Transient Analysis Results July 1977-Hestinghouse Owners Group on RCS Overpressurization.

ZOSR-242 94 Amendment Nos.

LIMITING CONDITION FOR OPERATION

. SURVEILLANCE REQUIREMENT

'3.3.4 4.3.4.D.

Materials Irradiat on Surveillance Specimen i

Inspection. (per ualt)

Specimen capsules shall be removed and examined to determire changes in their material properties, as required by 10CFR50 Appendix H.

a.

i h

l-1 4

ZOSR-242 105 Amendment Ncs.

Basis 6.3.4 The surveillance inspection program has been developed to comply with Sectico XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by.10CFR50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10CFR50, Section 50.55a(g)(6)(1). The design of the plant, state of non-destructive testing technology, and access to areas to be inspected require such relief.

The Reactor Vessel Material Surveillance Program is designed to evaluate the effects of radiation on the fracture toughness of reactor vessel steel based on the transition temperature approach and the fracture mechanics approach.

10CFR50, Appendix H, paragraph II B.1 requires that the reactor vessel material surveillance program shalf meet the requirements of ASTM E185-82 such that the surveillance capsules represent end-of-life fluences at the reactor vessel surface and 1/4T wall thicknesses.

4 i

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ZOSR-242 119 Amendment Nos.

L

6.6.1.G PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, criticality, and hydrostatic and

' leak test limits, shall be established and

. documented in'the PTLR.

The reactor vessel pressure and temperature limits and the heatup and cooldown rates are-addressed in Specification 3.3.2.A.

The analytical methods used to determine the pressure and temperature limits including the heatup and cooldown rates shall be those previously reviewed and approved by the NRC as described in HCAP-13406, "Heatup and Cooldown Limit Curves for Normal Operation

'for Zion Units 1 & 2", dated July 1992 and approved by the NRC SER dated The reactor vessel pressure and temperature limits, including those for'heatup and cooldown rates, shall be determined so that all applicable

-limits (e.g., heatup limits, cooldown limits, and inservice leak and hydrostatic testing

~ limits) of the analysis are met.

The PTLR, including iavisions or supplements thereto, shall be provided upon issuance for.each reactor vessel fluency period.

1 ZOSR-242 316b Amendment Nos.

1 ENCLOSURE 5 SUPPORTING DOCUMENTATION (1) HCAP-13406, "Heatup and Cooldown Limit Curves for' Normal Operation for Zion Units 1 & 2" (2) HCAP-10962, " Zion Units 1 and 2 Reactor Vessel Fluence and RTPTS Evaluations", Revision 2 (3) LTOP Enable Temperature Methodology 20SR-242(50)

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