ML20148C273

Proposed marked-up Tech Specs Incorporating Rev to Heatup & Cooldown Limit Curves

ML20148C273
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Site:	Fort Calhoun
Issue date:	03/14/1988
From:	OMAHA PUBLIC POWER DISTRICT
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ML20148C249	List: ML20148C244 ML20148C265 ML20148C273
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NUDOCS 8803220317
Download: ML20148C273 (8)

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e AITAGMENT A 2.0 LIMITING CONDITICMS FOR OPEpATION 2.1 Reactor Lootant tvstem (Continueo)

( 2.1.2 Heatuo ano Cooteown Rate (Continued)

(a) The curve in Figure ?-3 chall be used to predict the increase in transition temperature caseo cn integrated fast neutron flux. If measurements on the irraciation specimens indicate a deviation from this curve, a new curve snail be constructed.

(b) The limit line on the figures snall be upda+.ed for a new integrated power period as follows: the total integrated reactor themal power from startuo to the end of the new

erto :c'11 be converted to an eouivalent integrated fast neutrt- a.y.osure (E>1 MeV). For this plant, based ipon i

surveillance materials tests, weld chemical composition data, and the effect of a reducco vessel fluence rate proviced by core load designs beginning with fuel Cycle 8,

• ne orecictea surface fluence at the initial reactor vessel l

eitline weid - ial for a0 years at 1500 "Wt and an 80%

uction l eo _ '

'eso-fact:r-- : 23)OM n/cm2 .' The flu (2h .u the ;ivercMidlations ycle 1 9' aver 30 4.3.

anci c ycle s a9e7.a:eJmuthal flux distribution (pMa, sed opQlots enerate usin l

predicted transition temp'Ersture shif t to the end of the new ceriod snali then be obtained from Figure 2-3.

(c) The limit lines in Figures 2-1A and 2-1B shall be moved parallel to the temcerature axis (horizontal) in the

( direction of increasing temperature a distance equivalent to the transition temperature shift during the period since the curves were last constructed. The boltup temperature limit line shall remain at 82*F as it is set by the NDTT of the reactor vessel flange and not subject to fast neutron flux. The lowest service temperature shall remain at 182*F \

because components related to' thic temperature are also not subject to fast neutron flux.

(d) The Technical Specification 2.3(3) shall be revised each time the curves of Figures 2-1 A and 2-1B are revised.

Basis All components in the reactor coolant system are designed to withstand theeffectsofcyclic(;gadsduetoreactorcoolantsystemtemperature and pressure changes. 1 These cyclic loads are introduced by nomal unit load transients, reactor trips and startup and shutdown operation.

During unit startup and shutdown. the rates of temperature and pressure changes are iimiteo. The design number of cycles for neatup and cool-down is hsed upon a rate of 100*F .in any one hour period and for cyclic operation.

L 8803220317 s80314 2-4 Amendment No. 22.A7.14.74.77.100 PDR ADOCK 05000205 P PDR .

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e 2.0 t.IMITING CONDITIONS FOR OPEPATIO!I yx 2.1 Reactor coolant system (continued) 2.1.2 Heatuo anc Coolcown Rate (Continued)

The maximum allowaole reactor coolant syr,wn pressure at any temperature is based uoon the stress limitations for brittle fracture considerations.

These limitatiens are cerivea by using the rules contained in Section Illd of the ASME Code including Apr,r.noix G, Protection Against Non-ductile Failure, ano the rules contuneo in 10 CFR 50, Aopendix G.

Fracture Tougnness Requirements. Dis ASME Code assumes that a crack 10-11/16 incnes long and 1-25/32 incnes deep exists on the inner surface of :nc vessel. Furthemore, coerating limits on pressure and temperature assure tnat the crack does nct r, row during heatups ano cooldowns.

The reactor vessel teltline mterial consists of six olates. The nil-ductility transition temoerY.ure (TNOT) of each plate was established t,y creo weignt tests. Char?y tests were then performed to determine at wnat temocrature tne olat<n exnibited 50 f t-lbs. absoroeo energy and 25 mils laterial exoansiot. for the lonoitudinal direction. NRC tecnnical

.osition MTE5-5-2 was uwa to establisn a reference temocrature for transverte c1rection t NT) of -12 F.

The mean RTHDT value 'or tha Fort Calhoun submerged arc vessel weldments was ceteminea to be -56*F with a standard deviation of 17'F. By apply-ing the shif t pred'*.tica methocology of the proposed. Regulatory Guide 1.99, Revision 2, a welr'. material adjusted reference temperature (RTNOT) was established at 10*F based on the mc n value plus two standard deviations.

The stancard deviation was detemined by using the root-mean-squares

( method to combi.u the margin of 28*F for uncertainty in the shift equation with the margin of 17'F for uncertainty in the initial RTNOT value.

Similar testir.g was not perfomed on all remaining material in the reactor coolant systen. However, sufficient impact testing was perfomed to meet appropriate qesign code requirements (3) and a conservative RTN0T of 50F has been as~:1blished.

As a result of fast neutron irradiation in the region of the core, there will be an increase in the TNDT with operation. The techniques used to predict this integrated fast neutron (E>1 MeV) fluxes of the reactor vessel ,

are descri.'ad_in_Section_3.4.6-obtheJSAR, except that the integrated ggg fast neutr;n flux (E>l MeV)_ishl0j n/cm , including tolerance at the 2

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~~Yhe 40 year ace-of=the: critical design life of the@eactor(ygssel vessel. DJ beltline weld material, over Since the neutron spectr.t and the flux measured at the samples and reactor vessel int de radius should be nearly identical, the measured transition shif t for i sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the differance in calculatet flux magnitude. The maximum exposure of the reactor vessel will i.e ot :ainea from the measured sample exposure oy applicattun of tne

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calibratec azimuthal neutron flux variation. The maximum integrated fast -

_ neutron (ILI MeV) exposure of the reactor vessel at the critical reactor tel bei iline location including tolerance is computed to be

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g , 2.9 101 n,'cm2 at the vessel inside surface for 40 years operation at v-2-5 knendment No. 22.A7.54.7A.77.100

2.0 LIMITING CC;10lTIO!15 FOR OPEPATI0tt 2.1 Reacter Coo tant System (Continued)

C 2.I.2 Heatuo ano Looldown Rate (Continued)'

1500 .'Nt ana 80% load f actor. The predicted shift at this location at the 1/4t ceptn from the inner surface is 332*F, including margin, and was calculatea using the snif t preciction ecuation of the procosed Regulatory Guide 1.99, Revision 2. The actual shift in T?iOT will be re-established periccically during the plant coeration by testing of reactor vessel material samples wnicn are irraciated cumulatively by securing them near the inside wall of the reactor vessel as described in Section 4.5.3 and Figure 4.5-1 of the USAR. To ccmoensate for any increase in the Tfl0T causec by irraciation, limits on the pressure-temoerature relationsnip are periodically enangea to stay within the stress limits curing heatuo and c:oldown. Analys)s of the second removed irradiated reactor vessel surveillance specimen W), comoineo with weld enemical ccmoosition data and recucco core loaqing_de.f ns initiated in Cycle 8, indicates that

'he-#4entdWind offl5 Effective Full Power Years (EFPY) at

.ege vyt m a g e w }(p i /cm 2on the inside surface of the reactor vessei. This resuits in a total snif t of the RTNOT of 35'F. :nclud1jn . s margin, locatten for the area of f;greatest as determined y ure g :ensitivity 2-3. (wela metal) at *.ne l' tOperation thrcu will result in less t. in @0 FPY.

14 The limit lines in Fi es 2 A and 2-1B are based on the following:

A. Heatup and Cooldown Curves - Frem Section III of the ASME Code.

( Appendix G-2215.

KIR

• 2 KIM

• KIT Kgg = Allowance stress intensity factor at temperatures related to RTilDT ( ASME III Figure G-2110.1).

K gg = Stress intensity factor for membranc stress (pressure).

The 2 represents a safety factor of 2 on pressure.

K IT = Stress intensity factor radial themal gradient.

The above equation is applied to the reacter vessel beltline.

For plant heatup the themal stress is opposite in sign from the pressure stress and consideration of a heatup rate would allow for a higher pressure. For heatup it is therefore conservative to consider an isothemal heatup or XIT = 0.

For plant cooldown themal and pressure stress are additive. l L 2-6 Amendment No. 22.A7.AA.74.77. loo

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• ' D ATTACllMENT a JUSTIFICATION DISCUSSION No SIGNIFICANT llAZARDS CONSIDERATIONS locu: entati:n ot the enemteal content at all Fort :..lhoun reactor vossei belt-i!na mater:als was compicted in L986 4 Reference 2) . The . lower ;nell longitudin-r al veld seams 3-410). :.nich consist or three stifferent weld wire heats. were ounc co oe limiting as a. resuit or high copper and nickel content. Since the

omo1 cation of these welds is unknown, the most limiting weld is used in the RTg,7 analysis. k' eld wire heat 12008. flux lot 2774, containing 0.23 w/o Cu ano t .95 2,o Ni was found to be limiting when using the shift prediction equa-t ion o f 10 CFR 5 0. 61. 7his weld was also assumed for use with Rec. Guide 1.99 3 rat t 7.ev -, wnen preparing the tacility License enange ro update rhe.heatup

, and ecoldown limit curves to 15.0 EFPY (Reference n. Further avestigation covealed .at weid wire neat '7204. ilux tot 1774 with 0.22 wo cu and l'.02 a n '. . i: : ore ..mitinc wnen ustnr *he Dratt Reg. Juide methouoior. Using the nom::try .ctor .nne. .:.a with  := seld wtre neat . n :.he b'e n itde i . 44 ca:t Nv ..quatian

.constraten nat *he ..xtst:nc 7echnicai reetiication

catu: .ino coo tdown . t: urves hterenco ,a are non-conservativo.

7he valid .ifetime of the >xisting 15.0 EFPY curves has been reanalv:ed using the :luence preciction equation developed in Reference 5 and the more limitin6 chemistry factor associated with the 3 410 weld seams. The Reference 5 fluence prediction equation for the longitudinal 3 410 weld is:

1. - (8.8x10 @ (0.68) + (EFPY 5.92)(4.8x1019) 4g g 2 32 This equation takes additional credit for the vessel flux reduction' program which was previously not taken in the Reference 3 submittal. The results of this analysis indicate that the current heatup and cooldown limit curves are valid to only 14.0 EFPY. Since Fort Calhoun Station has been operating for less than 10 EFPY. no challenge to the reactor coolant system or violation of Technical Specifications has occurred from using the existing curves.

This proposed amendment corrects the labels of the heacup and cooldown curves (Figures 21A and 2 1B) from 15.0 to 14.0 EFPY, Figure 2.3 is corrected to reflect the more conservative shift prediction equation associated with the limiting weld wire heat in the lower longitudinal weld seams. The predicted 40 year inte5 rated flux was uso revised to be consistent with the fluence predic-tion equation used in this assessment. Reference to 15.0 EFPY in the Basis Section is revised to 14.0 ETPY and the corresponding change from Cycle 16 to Cycle 15 is also made.

Si rni fienne !H:ards ennsidaracions The proposed amendment to the Teuhnical Specification does not involve an unre-viewed safety question because the operation of the Fort Calhoun Station in ac-cordance wtth this change would not:

'11 :ncrease the probability or occurrence or the consequences of an acci-dont or malfunction of equipment important to safety previously evaluat-ed in the safety analysis report. The proposed revision to the Techni-cal Specification heatup and cooldown limit curves imposes more conserva-tive limits on operation by' revising the valid operating life oc the ex-tseing curves from 15 EFPY to 14 EFPY. .There has been no challenge to ~ ,

the reactor coolant system associated with using the previous curves since the Fort Calhoun Station has currently been operating for less than 10 EFPY. Theretore this amendment would not increase the probabil-itv of occurrence or the consequences of an accident or malfunction of equtement !=portant to safety previously evaluated in the safety ans Ly-

-sis report. L

's :rca:e the nossibilie. :or an accident or malfunction or different vpe than any "valuated provtousiv in the safety analysts report. This imencme nt o n t '. revises :he label defining the lifetime in EFPY ot the Tecnnical Specification heatup and cooldown limit cu rve s . These curves are bounded by the extsting Safety Analysis Report. There are no anti-cipated changes to the current operating practices. Therefore, the pos-sibility of an accident or malfunction of a different type than any eval-uated previously in the safety analysis report would not be created.

(3) Reduce the margin of safety as defined in the basis for any Technical- ,

Specification. The revised operating life for the existing curves was determined using a more conservative chemistry factor along with the shift prediction equation, including the appropriate 2a margin, as presented in Regulatory Guide 1.99, Draft Rev. 2. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.

Based on the above considerations, OPPD does not believe that this amendment in-volves a significant hazards consideration.

References (1) Docket No. 50 285 (2) Lotter from OPPD (R. L. Andrews) to NRC (A. C. Thadani), Fort Calhoun Specific Wald Chemistry Data Reporting Requirements of 10 CFR 50.61, dated January 23, 1986 (LIC-86 024)

(3) Letter from OPPD (R. L. Andrews) to NRC (A. C. Thadant), Revision of Application fot Am admant of Heacup and Cuoldown Limit tutveu, dacca July 10, 1986 (LIC 86 309)

(4) Amendment No. 100 to Facility Operating License DPR 40, September 8, 1986 ,

(5) Letter from OPPD (R. L. Andrews) to NRC, Revised Pressurized Thermal Shock Analysis, dared December 21, 1987 (LIC 87 692)

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