ML20024B277

Responds to 771006 Telcon Requesting More Info Re Basis for Vendor Conclusions on 770924 Depressurization Event

ML20024B277
Person / Time
Site:	Davis Besse, Crane
Issue date:	10/07/1977
From:	Lauer J, Lazar A BABCOCK & WILCOX CO.
To:	Domeck C TOLEDO EDISON CO.
References
TASK-*, TASK-03, TASK-06, TASK-07, TASK-3, TASK-6, TASK-7, TASK-GB BWT-1579, GPU-2412, NUDOCS 8307080200
Download: ML20024B277 (2)

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• 303 Itadisen Avenue J. C. Lewis 6 r

Toledo, Ohio 43552 D. J. Delacroix '

E. C. riovak/2c .

Subject:

Toledo Edison Cc: pany ~

E DEPRESSURIZAT:C:: EVENT OF SEPTC;EER 24, 1977 E Davis-Besse Unit 1 t 1155-14 c.

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Dear fir. Domeck:

Our letter E!.T-1578 dated Octcher 5,1977 advised that Edf is revieaing the available data regarding the de:ressurization event of Septe.-ber 24 and we consider that there E, s has been no degradatica of safety in the plant. By telecon of October 6, you have F-advised that I;RC.would like c=re informatica regarding the basis for our conclusions. .

n in which  !

The ce=;:enedts .are desig.ed for ferty cycles of a d:presguri:stien transie't the pressure drops 1/03 pst and the te :erature drops 62 F in fifteen minutes. In (.

this actual transient the pressure dro;;cd 1250 psi and the tc perature dropped 45'F

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in 7.5 cinutes. The stresses due to pressure are net sensithe to tice in these load- -

ing ranges. Since the pressure change of the actual transient is less than the g pressure change of the generali:ed transient, the stress effect cue to pressure would '.

be less than calculated for tk.e generali:ed tra.sient. Although the rate of tgeperature  ;

change is higher in the actual transient', the overall te ;erature change is 17 less. t These two differences tend to offset each cther such that the resulting stresses for the actual transient were no worse than tne calculated stresses for the design transient.

i-One steam generator a;sparently boiled dry during the depressurization event because the auxiliary feed,ater turbine failed to cern up to speed. The design transients L include twenty cycles in which feed.<ater fice is icst to one generator and the  :

generator is evaporated to a dry pressuri:ed c=ndition. The introduction of fee &ater G into a dry stea= generator is a design c:nditica and will have no harmful effects.  ?

The major concern while the steam generator is dr/ is variatien in the tube-to-shell temperature differences. In the actual transient the steam generator was dry for a ,

short period of time and the generator re sined pressuri cd. During this timg of ap- -

- proxicately 13 minutes, the reactor coolant tee;:erature drep;cd a total of 50 . This tamperature dn:p was not sufficient to cause excessive stress or permit defonnation  :

of the tubes, and is within the established design lic:its.  : E F

The actual stresses were no wc se than the calculated stresses frem the ' design transients f and consequently the fatigue usage resulting frem the actusi transient is no worse than ,

that for the design transients. The predicted fatigue usage for this transient is the F.

sane as that of one design cycle of rapid d:pressuri:ation and one design cycle of .'.

I startup of a dry steam generator. There is no change in the csiculated.fstigue life L of the cor.ponents. Since the stresses and defor:.:tions resulting from design transient analysis are acceptable, there is no reason to cxpect oventressing or material defonnatic -

in the RC system due to the actual transient. He do not consider it necessary to e lhe Babceck & W.tcos Compwiy f (St.Wehed 19G7 c- .,

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Byr-1579 Page 2 NI Getober 7, 1977 b;; O r:

conduct a detailed inspection of hangers and restraints for evidence of damage or deformation. ,?

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The reactor coolant pu os were all operated at er near saturation pressure (A2 and B1; for about one cinute and Al and 82 for about 45 minutes). There is so=e risk o -

cavitaticn dry bearir.gs damage and resultingto thedaimpe11crs age. and also a risk ths.t saturated steam would cause -

L In addition, radici offsets due to cavitation may dacage the seals.  ;.

flow for about 1-3/4 minutes after containr.ent isolation valves were clo  !!e have 9

reviewed dacage is these stall. conditicas with the pu: p canufacturer and concluded that the risk of Disassedly and inspectica of the seals, bearings, and ir:sollers )

would not provide 1 ggt assurance that thay will o;erate pre;erly. Therefore we ,,

have recc::nended that the cu ;s be instru ented to censure shaft vibration, s,eal p cavity pressures, RC pressure, standpi;e leakage, and seal injection flea and targerature. b Mode 5 and the cbserved ; ara sters shcu no irif  !?e caticn.of expect to damage.Each .-

have st=fler test ru s in :*cte 3 v. hen che PJ orassure is above 1300 psi. If these L

I be safely operated as designed. runs also sh:w no indication of da: age, B&'.1  ; wou l,

B&!! has evaluated the 9/24 fr.cident with regard to its effects upo'n fuel perforcance

.This conclusten is based upon the following consideraticas:and has D: co h.

Prior to the subject transient the react'or had been operating for ap- I prwxicately cne week at a caxie.:n of 151 of ratad pcuer; fr-ediatel prior to trip the pcuer level was n.prc.s!:.stely 10t of rated pouer,y -,

F therefore, the heat gaseraticn in- the core (decay heat) during the depressurizatien transient was extre..ely 1ca and significantly less .

than that produced by the reactor coolant pu ps. '

The core burnup'on 9/24 was approximately.1 EFPD.  ;'

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, During the transient the caxicus fuel red in' tert:a1 pressureits bee n .

conservatively esticated to have been no : ore than 300 pst greater than L the minicu= RC systen oressure; the raxt:,;:: fuel red cladding tec erature  %

was 550F.

The tensile stresses icrosed on th: cladding as a resu;lt of (

the 300 psi pressure differential existed for less than ene hour. For g cladding with low irractation exacsure ene.c4 to this temperature / pressure codination no defor:.ation or failure v.uld he predicted. :J G

Reactor coolant teccarature, pressure, and (1cu rate data obtained during I the course of this transient indicated that th3re was no significant heat generaticn in the reactor core; this data further indicates that no i-significant boiling occurred in the core.  :

Yery truly yours, '- '

r A. . azar nior Project Manager

. = l JAl./hj .s C us4- t.

J .1.auer, Project Manager

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