ML19224B533

Discusses Reactor Trip on 780423.Findings Concern Reactor Coolant Pumps,Control Rod Drive Mechanisms,Fuel Components & Reactor Vessel Transient Results

ML19224B533
Person / Time
Site:	Crane
Issue date:	05/02/1978
From:	Rogers L BABCOCK & WILCOX CO.
To:	Geoffrey Miller METROPOLITAN EDISON CO.
References
SOM-II-140, TM-0237, TM-237, NUDOCS 7906150263
Download: ML19224B533 (8)

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DOCDENT NO:

COPY MADE ON OF DOCDEtiT PROVIDED BY METROPOLITAN EDISON COMITiY.

Wilda R. Mullinix, NRC 79061502G3 79061802G3

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u n C.: i! U li v V~^r Po..er Generat:cn Group P.O. Oct 1260, Lyr.chburg, Va. 24 505

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Te'e,chcre: (804) 334 5111 w y 2, 1978

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Mr. G. P. M ill er Stat ion Superintendent Metropolitan Edison Cc=pany Pact Office Box 480 Middletown, PA 17057 Subj ec t : Reactor Trip /E.S./Cccidow". Incident of 23 April 1978 Dea-Gary:

I.7 accordance with.~1ctu r eqi.r s: for the evaluation of specific areas of conc ern follcains the subj ect plan; transient, EV.I has cc=pleted a pre-

' M w;. review of the effects cf t " c"'_C ied data.

The findings con-cerning the reac tor coolant prp, the control rol drive mechanisms, fuel c or:po n ent s, the 9eactor Cooli.; Systen water chemistry, GTSG transien; result s, reactor vessel trans ant result s, and reactor ecolari piping, A

including pressurizer, are as follc.s:

I.

Reactor Ccolant Praps Durine the first 22 minutes of this tre*sient. t hr e e

.n=c. c ver e c.rerat inz,

o two in "B" loop and one in "A" Iccp. The ver st co ndition, fror an :iPS3 standpoint, 'rould have affected the sin;1e pump in the "A" loop.

This p=p was flowing approxi ately 117,000 gp at the time the inciden' and would have require'i a rini== cystaa presst.re of abc.-

40 poig to prevera cavitation ir the impeller at a cold leg temperature of L6h 7.

A t, the indication is ths, the systen pressure only dropped to 752 psig,,te do not believe the p rps operated nder cavitating conditi0ns.

'de no t e t hat oft'r 22 ":i..;tas, single pu.p cperation in cach leap.zas in i t ic.t ed, however, by t..is tire the Reac;cr Ccclant Systr_m pressure was tack up to 21h0 pcig, r.ich would provide adequate li?Sh avr.i2 able to the puups.

Fro:t /crbal cbcervat. icns, we under.tand that injection uater aac.;a i n-tair.ed and there was no noted chance in shaf t s i srat ion.

We vculd like tJ det add.tional inforaz ien, cuch as seal cavity preasure reopense and se:tl lcakage during t:.is tranciu.it to furt her our knowlcdce ac to how these ceals respcnd tc such plan; transienta.

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G. P. Miller 5/2/78 As this was a very quick temperature and prersure ramp, 125 F in 3 minutes, ILLS psi in 3 minutes, we cannct make,ny statement as to the effect of long-terr r tigue life on the punp casing and cover.

a These transients vould h1.

o be evaluat ed by the '. ec hanic s Research Institute, the consultant that performed t'. e stress analysis on these pumps. We recently obts_:a ' a quotation for SE,000 for a similar analysis on the CMUD pumps.

If we are cd;ised to proc eed with this analysis, we vill pursue obtaining a e

quc a

)n for the TMI-2 pumps.

E R ecccm endat io n s BLW recommends continued operation of the reactor coolant eunt. s,

Startup and power escalation data pertaining to the reactor coolant pump seals and pump vibration data should be obtained and cc pared with baseline referenc e data. This data should be forwarded to ElW for fin 11 recc cendations and confirmation of c:a assessment.

We would recensend pursuing the analysis described above, however, this would not delay the pres ent operation of these pt ps'.

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.II.

CRLM's Confir:1 tion is requested that the safety reds which were.,i t hdrawn during the transient were driven, not tripped, bac:. into place.

Reccmmendations Eased en the above, and the similarities of this transient to the recently analyzed EMUD trsnsient (Marc h 20,1978 ), we do not feel there are any significant concerns regarding long-tern damage to the CRCM's or their ability to continue to perform as designed.

Final calculations to su.r. cart these findin~s are antici ated b.y a

e June 1, 1978.

In addition, the normal drive venting procedsre must be followed prior to returning the C?:M's to service.

I.I.

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It appears that the cecidsvn linit for BOL clad ccrpressica found in Limit / Precaution Curve 1.0, 05.2 was violated by as nuch' as 250 psic during the accidental derressurication, However, this limit represents a worst case envelope and dccs not realistically reflect the conditions encounterel in the TMI-2 ransient. A specific analysis of the TMI-2 conditions using the supplied reactor ecolant t emperature/precstre j

infarnation and the TACO code (Version 18) indicates that in ac tuality

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Miller 5/2/78 the fuel rod cladding never experienced a tensile force. This conclusion is based on a conservatl.e analysis. and all information available to date.

Further anslysis detuil.s can be found in cal-culational file 32-9072-03.

R econnend at ions Although the limits / precaution curve was violated, a trancient specific analysis indicates the fuel design criteria r.s not; therefore, PA'J

.does not feel there is an-/ concern regarding the ability of the fuel to continue to perform as designed.

All of the above findings arc ' cased en the infor ation p.ovided frem the site to Engineering through '.~uclear Servic es.

IV.

RCS Uater Chenistry A review has been cade of the o;<. rating and chenistry data associated with the chlcride con'=

'-e-ion of the Reactor Coolant Systen on

23. April, and it is cu cpinion that the high chlorides vill have no deleterious effect on the structural integrity of the Reactor Coolant Systen or associated sys =_ 3 and equip ent.

The efore, the Reactor Coolant Systen remains a::eptable for continuei operation (heatup and startup).

This letter cons;icutes the req _. red engineering evaluation in accordance.ith Plant Technical S,:ecification (3/4 h-17).

This evalu1tica is based :n the following condi icns:

C hlor id es

3. 0 p,n (nax. )

Oxygen

0. 0 ppm as indicatcd by chemistry analysis Sodium

$130 pps and the Reactor Cuolant Systen at Mct Stardby Conditions.

This evaluation is specifically based o 1 the basic pH assceiated with the sodium hydrcxide contini.ation and the presence of lov oxygen lev el s.

R ec om e n?.at io ns In agreenent with the directions cutlined for a recovery prcgram followint; a similar chenistry c~'

• ion of the plant, EiW vould expect a concentrated effort to eliminate the current Reacter Coolant System, D

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.r BWS, and auxiliary systens and piping contamination.

The Reactor l

Ccolt.. System sediu Icvels must be reduced below 2 ppm prior to

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achieving criticality in the re:1ctor with chlcrides as lov as pos-sible, 2nd because cf the current plant mair.tenance work, all control rod n.e rechanis s vi'l require full vern.irg before return to ope 1:icns.

It cust b e s tr es c e.1, however, that Met-Ed shculd ensure that se dit: levels c' e below 10 ppm in the control rod drive mechanism ec p: ents during these.enting operatienr, with chlorides similarly red:ei.

The 11." should have all of the conteninants reduced to the lowest achirrable cle2nup ceniitions. Significant flushing op rations on all :f the interconnecting pipin should be acecr.2'alished to eliminate o

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_o..,on 4 n c o.,. o c.m, v.

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'Las reviewed the data sent concernine the rapid ecoli:v which occ"--

- QI-2 on 23 A ril.

Eased on the information a

e rec eiv:-i ':y telephone ani telecopy, it appears the hot leg reactor eccla.

te perature d ; ped frcn 592 F to 46C F in 3-1/2 minutes; and the n' ' in te peratre Of ' ash ? vas reached 6-1/? ninutes af ter the trip.

~'re reactcr 00:12 ; temperature remained relatively constant at this :e:pa 2ture, fer at least 3 hou s, before a gradual eccido e was initia ei.

It is cc se.m ative to assume tha, the maximua tube to sisZ t=;peratu e difference is equal to the m'.xinua temperatu e g7 _.-.

n..

This is less than the l' : 7 tube to shell te peratre difference that ns W y ei for the ra_:ii cocldown at 5:CD in " arch and based on that analysis is acc eptable.

The rapid t=- peratre de:resse vill cause local ther=al stresses which vil.1 have to be eval atei in order to determine the effect of this tracci=.t c the fatigue life of the vessel.

Even though this transient vas 2:re rapid than the I'C'] cooldoun in March, the ' -tal te.perature i

drep vis cignificantly lass ; and E&'4 feels.he net effect of the T!I-2 i

cocid:vn v:.ll be less severe than the SWD ccoldcun currently being a., e,.._.-s_-.

i Attached is a graph shcrins the temperature dictribution in the support sk-2t turin.; the critic' transient timec consicered in the CTSG Stress Repor.

Since the re1c t.. coolant temperature dii not drcp below LSh F durin; the initial eccli;vn, this trancient chauld not produce a temper-atu e iradient any core svere than the condition snalyced in the strecs report.

Therefore, the effect of thic trancient on the support ckirt vou.ld be the same at a liPF/hr cooldovn.

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t A further ratisae justificatio.1 of this transient is that it is less severe than the O'UD rapid ccoldown. The EU.lD rapid cooldcun transient is in the process of being analycad by utilizing simplified,

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conservative analysis methods. This analysis is in the final stages and the increase in cunulative factors t' the more critical locations is small.

VII.

Reactor Ccolant Pirir.c and Pressuriner The transient data shown in the table below has been reviewed for preliminary impact On the operacing life of the TMI-2 reactor ccolant piping and pressurizer. The results of this review show there vould be no significant increase in the cumulative usage facter for any portion of the reactor ecolant piping or press" 4 ~a However, a detailed analysis would r.eed to be perforced at a later date to doc-unent th4.s transient's actual effects for purposes of the plant's life history CUF.

Most of the reactor ecclant piping has, at present, a low CUF as does the pressuriner. Therefore, the increase in CUF due to this one time cccurring transient wculd not cause any portion of the reactor coolant piping cr pressuriser to have a CUF greaper thaa allowable for the life of the plant.

Thus, there are no known consequences of this abnorral transient on the reactor ecolant piping and pra:surizer which would prevent startup of the plant at this time.

TEiPI3ATURE T u.L CUmr =_v.

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v p = = c u =.. =

1m o.o 592 57 5 2200 0.2 592 575 0.8 520 455 1.0 1750 2.5 820 2.7 455 3.0 760 5.0 980 8.0 1570 11.0 2200 13.0 21 Go 16.0 2200 28.o 458 h58 115.0 h80 h80 2200 A l 0

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All of the sho ce preliminary etaluation results are transmitted to you in crier to cupport your presentation of an initial reycrt.

EMI assu e: that a full-detailed report requirinc a longer time span to prt;sre will be presented.

In support of that prenice, E&'.i is ccr.:icuing to asceable the nececc2ry information to provide Met-Ei riti the detailed evaluations of the transient on the US3 cupplici epit ent.

The Site Office vill be able to 6 ve you an i

estL:ste of the detailed repcrt deliver;. date within 1 short pericd.

If you h1ee : - further questians, plesce do not hesitate to cor. tact ne.

Very truly yourc,

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