ML20042F794

Proposed Tech Spec 3/4.4.6, Pressure/Temp Limits

ML20042F794
Person / Time
Site:	Clinton
Issue date:	04/27/1990
From:	ILLINOIS POWER CO.
To:
Shared Package
ML20042F793	List: JSP-328-90, Application for Amend to License NPF-62,changing Tech Spec 3/4.4.6.1, Pressure/Temp Limits, as Result of Application of Reg Guide 1.99,Rev 2 ML20042F794
References
NUDOCS 9005090398
Download: ML20042F794 (13)

Text

{{#Wiki_filter:i U-601633. LS-89-012 Page 11 6f 23 _ REACTOR COOLANT SYSTEM 3I4.4.6 PRESSURE / TEMPERATURE LIMITS LIMITING CONDITION FOR OPERATION ~ 3.4.6.1 The reactor. vessel pressure and metal temperature.shall be i accordance with the limit lines shown on 6.1-1 (1) curv. W c' fcr hydrostatic or leak testing; (2) curv _ _ r " for heatup by sch-nuclear means, cooWwnle11 ewing a nuclear shutdowri drne Iow power PHYSICS TESTS; and v for operations with a critical core other than low power l 1 a. The maximum rate of change of reactor vessel steam space coolant tempera-ture during normal heatup or cooldown shall be limited to 100*F in any I hour, b. A maximum metal temperature change of 120'F in any 1 hour period during inservice hydrostatic and ' leak testing operations above the heatup and cooldown limit curves, and c. The reactor vessel flange and head flange metal temperature shall be > 70*F when reactor vessel head bolting studs are under full tension. 9 i APPLICABILITY: At all times. ACTION: With any cf the above limits exceeded, restere the temperature and/or pressure to vithin the limits witqin 30 minutec; perform an engi.cering evaluation to determine the eff ects of the cut-of limt condition or. the structural integrity of the reactot cociant system; determine that t% ieactor coolant system remair;s ~ at.ceptsble for continued operations or to in at leact. HOT SHUTDOWN within

12 hours and in COLD SHUTD0'A within the following 24 hours.

l i SURVIILLANCE RJ,Qu1REMENTS 4.4.6.1.1 During system heatup, cooldown and I. service leak and hydrostatic testing operations, the reactor vessel pressure and metal temperature of the reactor vessel flange surfaces, bottom head outside surface and bottom head inside surface, as measured by the bottom head drain temperature, shall be l determined to be within the operating limits defined by Figure 3.4.6.1-1 at least once per 30 minutes, j 4.4.6.1.2 The reactor steam space coolant temperature shall be de.termined to be within the heatup and cooldown limits of 100'F in any 1 hour at least once I per 30 minutes. CLINT.ON - UNIT 1 3/4 4-22 1 9005090398 900427 FDR ADOCK 05000461 p FDC

U-601633 LS-89-013 Page-12 of"23 REACTOR COOLANT SYSTEM PRESSURE / TEMPERATURE LIMITS SURVEILLANCE REQUIREMENTS (Continued)

4. 4. 6.1. 3 The reactor coolarit system temperature and pressure'shall be deter-mined to be_to the_ right of the criticality limit line of Figure 3.4.6.1-1 curv{ TC^; C' Z : 3riMwithin 15 minutes prior to the withdrawal of l

contr6T P ds to tr1_ N to criticality _ __ _cctegal properIIe +o OnliFiCtherge5jn_ reackcPyssqre Ve55el 4.4 M.4 Ine reEctorvFssel matenal spectifiens shat 1 De remove n3 Examinec = i as a function of time and THERMAL POWER as required by 10 CFR 50, Appendix _H, _ _ _ i ac danca with l schedule in Table 4.4.6.1-1. The results of these wirdirstien l .r:::: _:5 -- h el' re hall be used to adjust the curves of Figure 3.4.6. ~ - t Et,CTE b l ' 4.4.6.1.5 A. 4 rt-- ?!E "% ca-ciaea ch=1' M -^-^" d e + + k* f i " + aa== ...+.... s t m ..a 6.<m.. + w.. a, ;.....,. < 4. 4 4. s..... 4.. ..a

~.;L K&4 ; ~;.jj.~. p;.,fsf ?;,'.j'?4,.Mje f'j',gg{j'.' fp {g.. a_

pe-r 1 r -! - d ut-d te -^d!fy th* Ff gere 9?/'.'.5-L The r, rett ef thr,e- .. c a.,m..... e. e... s..- +&- s <3._.._.,. a.._._ <-.+4.., . w. i s w.._..._ a_.,. .,a 4._., + 4.4.6.1.6 The reactor vessel flange and head flange temperature shall be verified to be 170'F when vessel head bolting studs are under full tension: InOPERATI0HALCONDITION4'when'reactorkoolantsystemtemperatureis: a. O b 1. < 90'F, at least once per li hcut$. r-2. 5 80*F, at least once per 30 minutts. b.' Wittk 30 minutes prior to c.nd t.t least once per 30 minutes cdring .tensMning of the reactor vussel hetd bolttog stud +, except 10 percent cf l the bolting studs mey he fully ternioned at t 24'? but $,70'F. -~ l l 4 9 ( 3/4 4-23 CLINTON.- UNIT 1 i

~ f N h b U-601633 I ts.89-012 - Page 13 Of 33-i 1 1 16e I F i j. A A' B B' C C' gano l f I ). COREBELT NE l AFTERB0*8. tFT l l j I bd I l R,! I f I I I 1-g im0 f g f p i E Vi$$tL / f { Ot5 CONTINUITY tiuit$ / f f I I I I l g '" 1 I l C I J/ / l t p i "4 - t. .j / ~Non CUnva$ 5 . C 4Ru Rio:Cir o . la MO TO A6 PLY AS Tit ITSF0fif0vtAR$ y/ { / ta tsPu os ePtRAhot \\ .M. '1 l 4 e t i . NIW SOCF R$D i I g APFthCtk. G LtMIT$ (1963s i j j $v$ttM MYDROTf 5T L MiT l l~ i (- so'o A. j l wtTt* 6 UEL IN VES$t L i t 1 E ! 3f!p*1 E NON NUCLE AR Mt ATING L tT I g m== A0iTV' C. NUCLE AR'ICORE CRITIC ALI ^~ Liust BA$tD ON GE BWR Lausf B 700 - 70*F / LICEN5tNG TOPICAL R! PORT g. / NE D0'21778 A 1 /C 1 A'. B'. C'. CORE BE LT LIN LIMITS At TER /C-l' / AN A$$UMtD80 F TEMP $HIF T f ROM AN INITI AL wtLD RTNDT 08 30-F. $tt NOTE. I I O o tw am 3m 4x $00 00 MINIMUM Rt ACTOR V L$$tL MET AL TEMPERATURE.t'Fl Figure 3.4.6.1-1 Minimum Reactor Vessel Metal Temperature Versus l Reactor. Vessel Pressure l CLINTON - UNIT 1 3/4 4-24 l

U-601633 LS-89-012 i i' Page 14 of 23 1600 CURVE A EFPY 2 B C g 1400 1200 5 <r. !Ne,1000 s R d 12 W l w 800 h-0 AND C - CORE BELTLINE WITH ASSUMED 130*F - g SHIFT FROM AN INITIAL WELD RTNOT OF - 30*F h CURVES A. B AND C ARE VAllD FOR 12 EFPY OF OPERATION M g 600 CifRYG A INCLUDES BELTLINE f m {' ART W1.UES SHOWN BELM: l EFPY ArJ (*F) 4 58 400 o ? 03 J-1* '. 00 %12 pig n- .w a - SYSiEM HYDRCTTST LnMit WITH FUEL If4 VESStL B - NON NUCLEAR HEATING 200 LIMIT BOLTUP C ~ NUCLEAR (CORE CRITICAL) 70'F LIMIT l l l f f O 0 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) Figure 3.4.6.1-1. Reactor Vessel Pressure Versus Minimum Reactor Vessel Metal Temperature CLINTON - UNIT 1 i 1 3/4 4-24 4

'IABLE446Q - ~ n C REACTOR VESSEL MATERIAL. SURYEILtAMCE PROGRAM-WITHDRAWAL SCHEDULE-g CAPSULE ESSEL LEAD WITHDRAWAL TIME. e NUMBER LOCATION - f FACTOR at-tY-10. (EFPY) 5 1. Capsule 1 3' . 09- 0.M 10 e i 4 , 2. Capsule 2 177* r -G O.(pT 20 L 3. Capsule 3 193* -&e9 O.fo~l Spare s 3 6 s e l e 9 e e e 2C7 %48 ~ ?;- "S0-O PJ BJ Lt O e ,y- + ,a ye e -, e. -ds.,,.w-a, -s --c ~- ,S-,"- v. x -m< - = - .c----~-- - - -. - - -- - - -,--

q ~ U-601633 a.. LS-89-012 ~ Page 16 of 23 I i INDEX-B'AS ES' .SECTION PAGE INSTRUMENTATION (Continued) Radioactive Liquid Effluent Monitoring Instrumentation.... B 3/4 3-8 Radioactive Gaseous Effluent Monitoring Instrumentation... B 3/4 3-8 3/4.3.8 TURBINE OVERSPEED PROTECTION SYSTEM....................... B 3/4 3-8 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION................... B 3/4 3-8 3/4 3.10 NUCLEAR SYSTEM PROTECTION SYSTEM - SELF TEST SYSTEM....... B 3/4 3-9 Bases Figure B 3/4.3-1 Reactor Vessel Water Leve1.................. B 3/4 3-10 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 + RECIRCUl.ATION SYSTEM.'...................'.................. B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES.......'................................ B 3/4 4-3 3/4.4.3 REACTOR C0'0LANT SYSTEM LEAKAGE Leakage Detection-Systems...... B 3/4 4-3 Ope rati onal Le a kage....................................... B 3/4,4-3 1 3/4.4.4 CHEMISTRY................................................. B 3/4 4-4. 3/4.4.5 SPECIFIC ACTIVITY......................................... B 3/4 4-4 3/4.4.6 PRESSURE / TEMPERATURE LIMITS............................... B 3/4 4-5 .3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.......................... B 3/4 4-6 '3/4.4.8 STRUCTURAL INTEGRITY.....................'................. B 3/,4 4-7 3/4.4.9 RESIDUAL HEAT REM 0 VAL..................................... B 3/4 4-7 Bases Table B 3/4.4.6-1 Reactor Vessel Toughness Values............ B 3/4 4-8 Wj [. = Bases Figure B 3/4.4.6-1 Fast Neutron Fluence (E>1 Mev).at as a Function of Servi ce Li fe................................ B 3/4 4-10 . Figure B 3/ 4.6- -P.redicted-Adjustment Of P.:f:r:5:: T::p r:ture, ,, :: : Function of F%ence Ind-Capper Centent per

• 4

-Regttleteri Ceue 1. 33..'DC4ATfD ;........T..b ra r.-... B 3/4 4-11 CLINTON - UNIT 1 xviii

n 1 g U-601633 ~ 1.S-89-012. No CHMM&S ~PRombeb rose n a a _ REACTOR COOLANT' SYSTEM FoR C6NTI NUITY ONLY BASES 3/4. 4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour thyroid and whole body doses resulting from a main steam line failure out-i side the containment during steady state operation will not exceed small frac-j tions of the dose guidelines of 10 CFR 200. The values for the limits on speci-fic activity represent-interim limits based upon a parametric evaluation by the 3 NRC of typical site locations. These values are conservative in that specific site parameters, such as site boundary location and meteorol.ogical conditions, were not considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 micro-curies per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 microcuries s i per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon ) which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 0.2 microcuries per gram DOSE EQUIVALENT I-131 but 1ess than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 must, be restricted to no more than 800 hours per year, approximately 10 percent of the 1 i unit's yearly operating time, since these activity levels increase the 2 hour thyroid dose at the site boundary by a factor of up to 20 following a postu-l lated steam line rupture. The reporting of cumulative operating time over 500 hours in any 6 month consecutive period with greater than 0.2 microcuries j per gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission eval tion of the circumstances prior to reaching the 800 hour limit. I i Information obtained on iodine spiking will be used to assess the parameters L associated with spiking phenomena. A reduction in frequency of isotopic analy-i L sis following power l changes may be permissible if justified by the data obtained. Closing the main steam line isolation valves prevents the release of activity to the environs should a stcam line rupture occur outside containment. i The. surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to l take corrective action. 1 3/4.4.6 ' PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by norme? load transients, reactor trips, and start-t up.and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown the rates of temperature and pressure changes are limited so that the i maximum s,pecified heatup and cooldown rates are consistent with the design ( i assumptions and satisfy the stress limits for cyclic operation. The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture p toughness requiremente of 10 CFR 50 Appendix G and ASME Code Section III, c Appendix G. The cuiv., are based on the RT and stress intensity factor NOT information for the reactor vessel components. Fracture toughness limits and the basis of compliance are more fully discussed in FSAR subsection 5.3.1.5 j entitled " Fracture Toughness." CLINTON - UNIT 1 8 3/4 4-5 Amendment No.18 I

M<m .* gr. - f

• cU 601633 LS-89-012 Page 18 of 23' h

i I

REACTOR COOLANT SYSTEM I

BASE'S \\ i 3/4.4.6 PRESSURE / TEMPERATURE LIMITS (Continued) LThe reactor vessel materials have been tested to determine thei,r initial RTNDT* The results of these tests are shown in Table B 3/4.4.6-1. Reactor operation

and resultant fast neutron (E greater than 1 HeV) irradiation will cause an increase in the RTHDT.of the core beltline reg refore, an adjusted reference. temperature, based upon the fluence,ga content and esti

,ej re icte'd using2 E:~? '{ . M;;r: _ ge materialig"u ongL Re uktc7{1etETde l,41 a ~Radi4+4ers coppeg ...... enf9;;r: E.. 2 14te. l m tement eUR Ma 4 a , cugk B** end C** ~ ConJHion$ ot I e pres ure emperature ur l. f r them,_._Jn_-hrrne: lhe actuaT '12.E4eoHve j L includes an assumed shift in RTHDT Oil Power l shift in RTNDT of the vessel material will be established periodically during N # ' ' operation by removing and evaluating, in accordance with ASTM E18 and l j, 10.CFR 50, Appendix H, irradiated reactor vessel. mater,tal specimens nstalled near the inside wall of the reactor vessel in the core area. The frradiated ___Qgie 1 specimens can be used to ict reactor vessel material transition temperature shift. Flux wires which oved.after the first fuel cycle and#at later qq' ~l intervals with the surve ce sp'ecimens are analyzed and provide an improved gp=g neutron fluence estimate for the reactor 3etse_1.httis then sdt _modifv Bases I cure B 3/4.4.6-1 W --- % t Fel:E r ve:W Me?T tr-ritten, n l L C M C d ftd.The operatin re7.4 1 1-1(fir __be- ~ Preda+; ens, an m ' :,, of reactor i 4as 1EmauYred on the basis of the specimen da co ihe teamm n =+ ions e9 'Rektery veMet un 4..'3-Urd??!I1c-1 4 .-wGolde i.% material I c erature limit lines snownM,RuaionL. K curves iguiYs'T.W. w,ng;4;on pressure- -G'r nd CC A', for reactor critica1'ity and for inservice leak and hyd c' +einprWre. e ting have tiein provided.to assure compliance with the minimum temperatcre $g., g Df . requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for Re5ol"+eN inservice leak and hydrostatic testing. Gude14 L The number of reactor vesse) irradiation surveillance capsules and the frequen-g; g* l cies for removing and testing the specimens in these capsules are provided in l Table 4.4.6.1-1 to assure compliance, with the requirements of Appendix H to a$cers l. 10 CFR 50. GndtdM [ 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main' steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment; however, single failure considerations require that two Valves be.0PERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges. l B 3/4 4-6 CLINT,0N - UNIT 1

- f' BASES Tl B 3/4.4.6 c.o n C REACTOR VESSEL TOUGHNESS VAi.UES'. 5 E ~ MIN.' EOL a LIM HEAT #-SLAB # -{gjTITtG -MA

• MAX ^ E0L

-2 UPPER SHELF g INE WELD SEAM I.D. OR tpt; RTNDT(F W T ('F) (FT-LBS)' RTNOT (*F)'- Q I. COMPONENT OR MAT't-TYPE HEAT #/ LOT # CU(%) f(%) s I[-64-B5 4 9 PLATE SA-533 GR.B. C~4380-2 0.07 )-0;414 #2 (49 CL.1 t,0 43 WELD N/A 76492/ 0.10 -0 MG. IM h5 ~7F 50IM [- 8- - L430827AE-1,og These values are given only for the benefit of calculating the end-of-life (EOL)9 NDT' h_ _3 RT NOTE: * = HEAT # - SUB # HEAT # - SLAB # NON-BELTLINE MT'l TYPE OR OR . HI ST S ING I. COMPONENT WELD $ TEAM I.D. HEAT #/ LOT # NDT ws [ SHELL RING SA-533 GR.B CL.1 C4240-2 -10 A2758-1 m' BOTTOM HEAD DOME A2757-1 " C4027-1 +10 BOTTOM HEAD TORUS TOP HEAD DOME C4374 -40 TOP HEAD TORUS-A2879-2 -10 TOP HEAD FLANGE SA-508, CL. 2 CCZ 41-5478 -40 SER 915 VESSEL FLANGE CWS 51-5218 0 SER 878 FEEDWATER N0ZZLE Q2AL10W -20 5 N, $ w h -G40 Om. Q + 9 cwe ag. a p. Asi-em.'_w-.*_im- ,.ui.y%., _-au.s,-sme-masse,iy.a-wws,. h g,-.,.

.,x BASES T BLc B 3/4.4.6-1 (Continued) h. j O REACTOR VESSEL TOUGHNESS VALUES i 5 e. h HEAT # -~ SLAB ~# HEAT # - SLAB'# fI / E NON-BELTLINE MT'L TYPE OR OR ~ HIGHEST STARTING; Q I. COMPONENT - ifELD STEAM I.D. HEAT #/ LOT # . RTNOT ( F)

4. ~

~m, H WELD PER GE PURCH. SPEC. NO CNV AVAILABLE O (PER PURCIC A SPEC.. REQUIREMENTS) 1 l CLOSURE STUDS SA-540 GR. B23 11312- +10 - I OI j $ h. ) ~- d' < ft. ^ lQ = .i h' R - /) o, l-- L $ li 6 l-I -i l. l 4 1 J> O Z %~ 'i-C ~ 2 5; ? " i @l. I ESO O u; I f e -'s + - r 's~ + - - ~ <- -m w~' ~ ' < ' ~v--! o -'*=r-e n -v,- ~w* m~u wm n s o e"~k

Te lace w;+h Insec+ page. U-601633 v LS-89-012 Page.21 of 23 r / \\ / \\ / j ) / \\ ) i \\ / / / 1 f \\ 6 I l s e e 3 \\ / / \\ f ) N / / \\ / I s' / 3 ./ / \\ / / e 4 } { s / a a \\ f f M \\ / ,r x / / / / T / ( "A \\ / F e y / c.53 .r \\ ) 't. / A } ,r e / T F } \\ f ,r s / / 1 lg f \\ ) / \\ ) \\ f A s 5 / .r s 1 / J' T i.- r f \\ / ,/ s f s \\, f / g \\ / / / .r / \\ / ) ,r \\ r J \\ A' r Y o 40 o to-20 so senvict ure tv ri+i Bases Figure B 3/4.4.6-1** Fast Neutron Fluence (E)1 Nav) at k T as a Function of Service Life

• c
• At 90% of RATED THERMAL POWER and 90% availability
  • Initial Figure; to be modified per Specification 4.4.6.1.5 CLINTON - UNIT 1 B 3/4 4-10 d'

v- \\; ~ ,C U-601633 "~ LS-89-012 L. ~- - Page 22:of' 23 10 i 7;;- 9 i 8 7 I NOMIN AL 1.0. 4 1 2 L i i t l A l 6 co o x N '1 ? 5 i 'w l g l o y-d- l L. 5: 4 91 a: 'g W ,i: i l-3 L j L 1 { 2 l. !i

l. '

i-1 l 1 1 1 i l f t O 0 10 20 30 40 50 SERVICE UFE (years') i Bases Figure B 3/4.4.6-1. Fast Neutron Fluence (E> 1 MeV) at I.D. Surface as a Function of Service Life ' At 90% Rated Thermal Power and 90% Availability CLINTON - UNIT 1 B 3/4 4-10

L 4, _-

t i n C za looc 2 [40 + IOOO(%Cu-0.08) + SOOO(%P O.008)] [f/10 ]/2 19 1 y e c A= t- %Pe0L0t2 c y z Q 035%ce ~ eco 0.30%Cu " g O.26%Ce ~ 'A 2 N . %c= a uurr frj soo - O.tS*/ Cu c,y 200 o,too/ co q W( b ,d, NOTE: DASHED UNES REPRESENT GE RECOMMENDED EX1RAPOLAT10N3 ' 1' 100 - ,c, ]" TO 20*E FOR BWR OPERATION. ~ '~ > )b! h so p. l / / /- l / / 1b // / / / / COWER LIMIT' 40 fl // / / / 7 co.o.oe lE 3o / /^ l / / ejo P4.OO8 t-4 sfl l l}kI l $l ll l l I 1 I Yl q go 2 : 10 4 a 10 " 2 4 6 8 to 2 4 e a 10 ' - 2 l4 s t f

• FLUENCE,ndm' IE)IIAsVI s

Bases F dre B 3/4.4.6-2 Predicted Adjustment of Reference Temperature, "A", as a Function of luence and Copper Content per NRC Regulatory Guide 1.99 (Adjustments.<S0'F are hown i in This Figure by Extrapolating the Regulatory Guide 1.99 Curves To Accou ygc for Lower BWR Fluences). gy4

• 3S U s S.

I oC" .m 0 .~ Lh .}}