ML19322C507
| ML19322C507 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 10/07/1977 |
| From: | Lauer J, Lazar A BABCOCK & WILCOX CO. |
| To: | Domeck C TOLEDO EDISON CO. |
| References | |
| TASK-TF, TASK-TMR BWT-1579, NUDOCS 8001170836 | |
| Download: ML19322C507 (2) | |
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'3abC00!G ilCOX eoc cenereuen creun P.o. Boa : 200. Lyn:nburg, Va. 24505 mmem m ansm October 7, 1977 bht-1579 File: T1.2 Mr. C. R. Domeck 12B Huclear Project Engineer Toledo Edison Comoany J. D. Lenardson Power, Engineering & Construction J. C. Lewis i
cc:
300 Madison Avenue D. J. DeLacroix Toledo, Ohio 43552 E. C. tiovak/2c
Subject:
ToTedo Edison Company 24, 1977 DEPRESS.URIZAT10!i EVEt{T OF SEPTEMBER n
Davis-Besse Unit 1
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Dear Mr. Domeck:
Our 'actter bht-1578 dated October 5,1977 advised that B&W is reviewing the avail dstr. regarding the depressurization event of Septer.rer 24 and we consider that the By telecon of October 6, you have has been no degradation of safety in the plant. advised that HRC would li x
The components are designed for forty cycles of a depresgurization transient in w In the pressure drops 1400 psi and the temperature dreps G2 F in fifteen minutes. o this actual transient the pressure dropped 1250 psi and the tem Since the pressure change of the actual transient is less than the in 7.5 minutes.
pressure cht.nge of the generalized transient, the stress effect due to pres ing ranges.
be less than. calculated for the generalized transier.t.
less.
change is higher in the actual transient, the overo.ll temperature charge is 17 These two differences tend to offset each other such that (One steam generator apparently boiled dry during the depressurization event b The design transients the auxiliary feeddater turbine failed to come up to peed.
include twenty cycles in which feedwater flow is le:st to one generator and theThe int generator is evaporated to a dry pressurized condition.
into a dry steam generator is a design condition a7d will have no harmful effects.
The major concern while the steam generator is dn is variation in the tube-to-sh was dry for a In the actual transient the steam generatotDuring this timg of temperature differences.
short period of time and the generator remained pmssurized.
This proxitately 13 minut6s, the reactor coolant temoem
,y of the tubes, and is within the established design. limits.
The actual stresses were no worse than the calcula ted stresses from the de and consequently the fatigue usage resulting frcm the act::zl transien't is no The predicted faticue uscge for this transient is the f
san 2 as that of one design cycle of rapid depress.n:ation and one design c that for the design transients.
li f-There is no cw we in the calculated f ati%
l Since the stresses and defoN.tions resulting f rom design 1.rmim.
startup of a dry steam genurator.
analysis are acceptable, there is no reason to c.wct overstressing or of the components.
We 40 not consider it accessag in in the RC system due to the actual transient.
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r-1 lho Dab:pck I. Wdret Corup.iny / E.bnbu', led inC7
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O Bl/T-1579 Labcock&Wilcox Page 2 October 7,1977 conduct a detailed inspection of hangers and restraints for evidence of damage or de for=ti on.
'The reactor coolant pumps were all operated at or near saturation pressure (A2 and 31 for about one minute and Al and B2 for about 45 minutes). There is some risk of cavitation darage to the impe11ers and aise a risk that saturated steam would cause In addition, radial offsets due to cavitation may dry bearings and res'ulting damage.
All four pumps suffered either loss of or erratic seal injection damage the seals.
We have flow for about 1-3/4 minutes after containment isolation valves were closed.
reviewed these conditions with the pump manufacturer and, concluded that the risk of damage is small.
Disassembly and inspection of the seals, bearings, and impe11ers would not provide' 100% assurance that they will operate properly.
Therefore, we have reco=nended that the pumps be instrumented to measure shaft vibration, seal cavity pressures, RC pressure, standpipe leakage, and seal injection fibw and Each pump has been run for two minutes with this instrumentation in temperature.
Mode 5 and the observed parameters show no indication of damage.
We expect to have similar test runs in Mode 3 when the P,C pressure is above 1300 psi.
If these runs also show no indication of damage, B&W would then reco:raend that the pumps may be safely operated as designed.
BLU has evaluated the 9/24 incident with regard to its effects upon fuel performance and has concluded that there are' no safety concerns with respect to the reactor fuel.
This conclusion is based upon the following considerations:
Prior to the subject transient the reactor had been operating for ap-proximatelf one week at a maximum of 15% of rated power; immediately prior to trip the power level was approxirately 100 of rated pr..er, therefore, the heat generation in the core (decay heat) during the depressurization transient was extremely low and significantly less than that produced by the reactor coolant pumps.
The core burnup on 9/24 was approximately 1 EFPD.
During the transient the raximum fuel rod internal pressure les bee n conservatively esticated to have been no more than 300 psi greater than the minimum RC system pressure; the maximum fuel rod cladding temperature was 550F. The tensile stresses imposed on the cladding as a result of For the 300 psi pressure differential existed for less than one hour.
cladding with low irradiation exposure exposed to this temperature / pressure combination no defomation or failure would be predicted.
Reactor coolant temperature, pressure, and flow rate data obtained during
.the course of this transient indicated that there wasino significant heat generation in the reactor core; this data further indicates that r.o significant boiling occurred in the core.
Very truly yours,
. pn g7 1{
A. h. Lazar, Senior Project Manager JAL/hj N
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J. A. Lauar, Project Manager
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