ML20076J907

Forwards Writeups Re Description of Event,Evaluation of Reactor Coolant Components,Evaluation of Rc Pumps & Evaluation of Fuel Pertaining to 770924 Depressurization Event

ML20076J907
Person / Time
Site:	Davis Besse, Three Mile Island
Issue date:	10/28/1977
From:	Lauer J, Lazar A BABCOCK & WILCOX CO.
To:	Domeck C TOLEDO EDISON CO.
References
TASK-*, TASK-03, TASK-3, TASK-GB BWT-1589, GPU-1021, NUDOCS 8307060607
Download: ML20076J907 (24)

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. j .. 3-Q n A /= f j _, ;t' i babcock &Wilcox e-er cem-- crco P.o. Box 1250, Lyn:hburg. Va. 24505 (ctober 28, 1977 Telephone: (804) 3a+5111 i " ' ' ' BVT-1589 i

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,' i File: T1.2/120

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cc: J. D. Lenardson w/a l J. C. Lewis

! . D. J. DeLacroix fir. C. R. Domeck --

. I P. P. Ana's/4c w/a Nuclear Project Engineer -

E fovak/lc w/a Toledo Edison Company Power Engineering & Construction C/;eM- aY M. C " // 1' /

300 liadison Avenue Toledo

• Ohio 43652 hM [* /

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Subject:

Toledo Edison Company 4 ' g , /Q c' t REPORT ON DEPRESSURIZATI 1 EVENT r O8I L. '

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Davis-Besse Unit 1 i

B&W Reference NSS-14 ce ._.g, , [ / %'rV S d b !. : ~;

Dear Mr. Domack:

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1 Cligarding By telecon of October the deprossurization event 10, you have24.

of September requested The NRC exit B&'.l intd input isial for eea Figof@gda.t.

October 7 summarized the necessary content of the report. risagra dtng B&'. n in the following areas in order to substantiate .the conclusior ~seH@p.y ma7.41vr,itg,-sp dated October 5 and 7: ve- o I . 4___-

10put Sa c.tir~ pD._4.1.12" '/7?

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@ A. Description of the event

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Evaluation of the reactoi coolant components s

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Evaluation of RC pumps -es: ~ T._ d  !

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_ct'.!J_J_.l y/3 bb) D. Evaluation of the fuel .9+ } _j 1._!d,- -

In order to expedite submittc1 of your report, we are sending GRtions.A,'C-and D-at i this time, as agreed in our telecon of October 24. We expect t5:fomard ,Sec-tion '

-B-by-November 7, and we will try to improve on this date. ].':~

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Section A describes the secuence of events as reconstructed frodi computer .alarnt print-r out, reactimeter plots, anc? control room recorders (Attachment'; '1,3. We -have httuttud pertinent recorder charts of T RC pressure, pressuri:er leN , (Attachments -A27#rian A4) and reactimeter plots of Rein,let temperature, RCS flow in h loop,; RC pressurO pressurizer level, and water level and outlet pressure of each sg, genpratori Uu. m n-ments A5 tnrough A13)

Section B will include evaluations of stresses in the pressure boundary, the depressurig tion transient, boiling the SG dry, jet impingement 'or the SG, and effect upon fatigue !

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The Babc ek & Wile a C mpany / Established 18G7 Hg & hmM &

M N MUI!DC..,. 8307060607 771028 -

PDR ADOCK 05000289 -

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October 28, 1977 BWT-1589 BabcocksAVilcox Section C explains the eyaluation which was performed to verify that thereThe was no significant damage to RC pump bearings, seals, or impellers (attachment Attachment C3 Cl).

transient as it affected the pumps is summarized in Attachment C2. The defines the instrumentation and operational checks applied to the pumps.

results of the operational checks are tabulated in Attachment C4.

Section D evaluates the effect upon the core to determine (11 whether steam was produced in the core (2) the maximum interna 1 Reactimeter fuel rod pressure, plots are and (3) whether maximum lift force exceeded the limit (Attachment D.1).

attached for reference Attachments D.2 through 0.6.

Very truly yours, A. H. Lazar Senior Project, Manager

/

JAL/hj gA.L'auer Project Manager Attachments i

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Ssquence of Events ,

The event started at time 21:34:20 on September 24, 1977. The plant was in Mode 1 with Power (MWT) = 263. .The turbine had been shutdown earlier in the ovening to repair a leak in the main steam line at an instrument connection between the turbine stop valves and the high pressure turbine. At this time a half trip of the Steam and Feedwater Rupture Control System (SFRCS) was initiated by an unknown cause. This trip shut the startup feedwater valve to #2 steam generator cnd stopped all feedwater to this generator (because of the low power level the main feedwater block valve was already shut, isolating the main feedwater control valve).

The low level alarm was reached in #2 steam generator at 21:34:44. Before thei cporator could identify and correct the problem, the low level in #2 steam generator produced a full trip of the SFRCS. This trip shut the main steam isolation valves end feedwater isolation valves in both steam geherators (time 21:35:18). STRCS clso started both auxiliary feeewater pumps. The number one pump performed as in-tended, however, number two auxilisry feedwater pump only came up to 2600 RPM, in-cufficient to feed its steam geaerator (#2).

The loss of feedwater, first to one and then both steam gensrators, caused an increase in primary water temperature, which resulted in an increase in pressurizer icvel and thus reactor coolant system pressure. At 2255 PSIG the pressurizer electro-matic relief valve received an open signal. During the next 40 : econds, it received nine different open and close signals. After one of those signals the valve stuck open. This provided a continuous 21" 1 vent path from the pressurizer to the quench teak. When pressurizer level got to 290", the operator ma,nually tripped the reactor (time 21:36:07). Enaugy escaping from the electromatic relief valve and three. main steam relief valves caused a rapid cooldown and depressurization of the reactor coolant systa=. Reactor coolant system pressure dropped to 1600 PSIG (time 21:37:17) initiating the Safety Features Actuation System (SFAS). This started high pressure injection and c4esed numernun eense6nment 4ew&ns6en vn&ves, including the queneh tank cooling lines.

With the electromatic relief valve c~till open and cooling water isolated to the quench tank, the quench tank rupture disc ruptured (time 21:40) relieving water / steam to the containment building. This discharge damaged a nearby ventilation duct, was deflected off this duct and directed onto #2 steam generator. The steam tors off cpproximately a 10' high x 20' circumferential section of insulation from #2 stea:a stnerator. The paint froa the then exposed area of the team generator was blasted . ,

tway. Oche h ea.;ie ? af ey-t t= ~ t i aa h- dI he steam in the containment #'[y.

include two fire alarms (one near RCP 2-2 and one near the pressurizer) and a single on high r channel RPS trip *1,W y,eactor building pressure (4 PSIG). .

Whenthe6EinsteaaIJreliefvalve/reseatedthedecreaseinreactorcoolantsystem temperature stopped and the high pressure injection pumps started to raise pressurizar level. At time 21:40:34 the operator stopped the high pressure injection pumps.- (The operators had been heavily involved before this time in regnining seal injection flow to the reactor coolant pumps. This flow had been stopped by the SFAS actuation. By -

21:39:40 the appropriate SFAS signals had been overriden and normal flows restored to the seals of the pumps). Reactor coolant system pressure continued to decrease until saturation pressure was reached and steam began to' form in the RCS (approximate time 2,1: 42) . This caused an insurge of water into the pressurizer and pressurizar level vent off scale high at 320 inches. During this level increase the operator, seeing )

I dverage reactor coolant system temperature and pressurizer level increasing, stopped cne reactor coolant pump in each loop (time 21'43:11).  :

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.."Dus to decrsasing pressure in #2 steam generator, th6 SFRCS system gave a lowpressup)l.oc.kpygitsignalat tims 21:48:33. Thisalertedtheopt{aterto the low fevel_and feed. condition of #2 steam generator. He blocked the low pressure trip (time 21:49:38), took manual control of the speed of 02 auxiliary feedwater i; pu=p and fed #2 generator (time 21:50). The operator saw the rapid addition of cold feedwater dropping the reactor coola.tc system temperature and stopped the feedwater addition to this generator.

At approximately 21:55 the operator shut the block valve for the electromatic relief valve on the pressurizer and stopped the venting of the reactor coolant syste=

to the quench tank. At 22:05 pressurizer level came back on scale. At 22:15 the operator started a second makeup pump to try and stop the pressurizer level decreace.

This additional cold water started the reactor coolant system on a slow decreasing temperature transient. At 22:17 pressurizer level reached the low level interlock and cut off the pressurizar heaters. At 22:23 the operator started a high pressure injection pump to try and stop the decreasing pressurizer level.

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• pressure in #2 steam generator again decreased to the point where the STRCS gave a low pressure block permit signal. The operator again blocked the trip and, through manual speed control of its auxiliary feedwater pump, restored level and pressure in #2 steam generator (time 22:25).

! With pressuriser level well on its way to recovering,the operator stopped the high pressure injection pump (time 22:27:44). At time 22:31 he restored RC makeup flow to normal. This stopped the slow decreasing RC temperature transient started at time 22:15. All plant parameters were now fully under control and the plant was brought to a steady state condition and a normal ' plant cooldown started.

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RC PUMPS I

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As a result of the September 24 abnormal system transient, the reactor coolant pumps experienced the conditions outlined in Attachment C.2. In order to demonstrate that there was no serious damage to the pumps, a series of operational j checks were performed as outlined in Attachment C.3. The results of the operatiorial checks are described in Attachment C.4.

B&W has reviewed the results of the operational checks and concluded that no detectable damage has occurred te the pump components. Bat! finds _the pumps to be serviceable for sustained full operational conditions with noimediateL ,

requirement for maintenance.

It should be noted that a step increase in vertical vibration of 2-2 pump was observed during the initial low pressure checkout runs. This indication was later assessed to be spurious instrument noise as a result of a loose connector on an instrument line. After the connector was tightened, vertical vibrations remained less than one quarter mil) peak-to-peak amplitude.

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I RC PUMPS SEPTEMBER 24 TRAtlSIEllT 0 DB-1 All four RC pumps were subjected to the following:

!l 0:00 . Reactor trip 1:10 SFAS trip

. 1:12 Seal return valves shut for 1:16 l 1:13 Seal injection valves shut for 1:52 i all four pumps operated for 1:15 with no seal e injection and no seal return flow during an.RCS' I de-pressurization

2

28 Seal return' valves open 3:05 Seal injection valves open

• 6:00 Steam formation ~

pressure oscillating near P for $30 to 45 minutes 36:07 Total seal injection flow 16w alargAT Pump 1-1:

7:04 Pump tripped 7:45 Shaft stopped 36:07 About one minute of low seal injection flow (near 2 gpm) flow imbalance starved seal injection 36:30 Seal return valve shut 1:12:55 Standpipe level high' 1:17:07 Standpipe level normal ,

Pump 2-2:

4:20 High vibration 7:04 Pump tripped 36:07 Lost seal injection for about one minute .

36:22 Seal return valve shut for about 40 seconds 1

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.CJ!fA.. JUT OF ISACAOu cuouder ruvat

• PURPOSE:

of abnormsl Asscos whe.ther maintenance is requirca of RC pumps as a resulf transient of 9/2 8/77. 1 Operational checks vill be required to demonstrate that no significant demage has occurred to the pump bearingc, shaft and -

seals. First series ofLater tests will be performed in Mole 5 due to operational on operationc! checks vill be performed in Mode restrictions by URC. duration not to execca

3. Each pump will be operated individually for et ten (10) minutes, providing all defined parameters renain within limits .

t establishe'd in this procedure.

Operational sequence will be as follows:

Torque values

1. Lift pumps vill be started and pump shafts rotated by hand.

A stethoscope vill be provided to detect are not to exceed 200 ft-lbs.

' ~

any unusual mechanical noises in seal housing area. (Tnis has been satis-factorily completed on 10/3/77).

! 2. Mode 5 testing 225 psig. .

j 2 .1. Instrumentation Required a see attached (.1A).

2.2 . Computer Data -

Printout NSS'special summary trend for running RCP every 15 seconds.

2.3 Folloring limits shall not be exceeded:

A. Shaft vibration - 15 inills peak to peak.

3. Total standpipe leakage (upper seal les.kage) plus seal return should 1

not exceedO.6 spm. If, during the test this limit f a exceeded, the possibility exists of an open seal. InIfno case vill total seal this'linit is exceeded, leakage be allowed to exceed 1.5 gym.

maintenance vill be required before further pump operation. .

C. All other normal plant limits and precautions prevail.

~. 2.k Sequence of Operation:

A. Secure standpipe flush.

3. Establish seal injection in accordance with plant operating procedure.

O. Neasure and record standpipe Icakage and return flow, confirm that

• total leakage limits are not exceede.l. .

D. Assure communication between contral ' room and personnel stationed at RCP standpipe leakece drain lino.

E. Countdown from 10 to 0 Start strip chart recorders at hir,h n.' Vpeed;

• ordanec with plant op. procedure.

b Start Reactor CoMant Busp 0-3 In After opprox. 11 ccc., reduco *r . hart apeed,.

r-, - - . - , - - ,.p _ _ - .,--,__,v...----.. .-- .- --- ,,-,-m , . - . , . - - - , - - - - - , - - - - - . ---_ . _ _ - - - - -

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'F. Hun purep for two (2) minuten tuu.cco any aoove .umit.c are excueued.

G. Data taken vill be accessed by Bf41 and P,-J representative:c.

II . Following ancec ment of data, pump may be run for an additional five (9) minutes to allow for venting proc:: dure requiremento.

I. Follow above r cquence on 2-1,1-2 and 1-1.

I J. Assessment of this data vill determine whether any maintenance in required before higher pressure operation is allowed.

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{ 3. Aho<a test vill be repeated with system pressure at creater then 1300 psic before final deter:aination on condition of the pumps is completed. .

CCE:nif 10/5/77 -

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. ST/dn' UP:

~ 107 0F POIl!TS TO. DE Il!OTitU:7.':TED Fo!! RC Ptf?

1. Upper and lover cavity pressures - all four pumps. .

2. Both horizontal D/N Vibration Probes - all four pumps.

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3. WR System Pressure or suction pressure, .

i Is . Vertical probe on 2-2 pump. .

h. Standpipe leakage vill be collected and measured during the test.

I ROTE: All of above should be recorded on an 8 channel brush. recorder located '

I in the control room.

RES:n1f 10/5/77

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'. STATUS OF CHECKOUT OF REACTOR COOLANT PUP 2S: /*/7.[7D g,7 h . f.

All four Reactor Coolant Pumps were run on 10/5/77, per the attached procedure, with the following results:

]

RCP 2-2 10/5/77 Run (2" min.):

i System pressure 225 psis 3rd Seal leakage plus p'~_ ~

2nd Seal cavity pressure 165 psig seal return flov 3rd Sesi cavity pressure 123.9 psis Horizontal vibration 5 - 7.5 mins Vertical vibration .25 mins ,

After the two minute run, the pump vas run for ten minutes for system venting.

About 30 seconds before the pump was shutdown, there was a step increase in eN vibration to 2.5 mills. The pump was run again on 10/6/77 for 10 minutes to checkout this phenemenon. 'Ihe vertical vibration was again .25 mills until about 5 seconds before shutdown where it increased to 2.5 mins. To anov a longer run tir.e, 2-1 and 2-2 pumps were run together for 10 minutes , then 2-2 was run.alone for 10 minutes. The vertical vibration stayed at .25 mins for the entire run. This will continue to be monitored during pump runs for plant heat up.

RCP 2-1 .

System pressure 225 psig 3rd Seal leskase plus

• • g 8E" 2nd Seal cavity pressure 132 psig -

return flow 3rd Seal cavity pressure 70 psig Horizontal vibration .5 - 7.5 mnis RCP 1-2 System pressure 225 psis 3rd seal leskage plus **g EE" 2nd Seal cavity pressure 40.29 psig return flow 3rd Seal cavity pressure 81.3 psig Horizontal vibration 5 - 7.5 mins RCP.1-1 System pressure 225 psig 3rd Seal leskage plus ,,g 2nd Seal cavity pressure 77.98 psig return flow 3rd Seal cavity pressure 89.27 psic Horizontal vibration 5 - 7.5 mins -

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• l The apparent discrepancy on seal cavity pressures on 1-1 and 1-2 was checked )

on 10/6/77 by installing pressure gauges at the pressure transmitters. The

- g- S read as follows: ,

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f 184 2nd cavity lil-3'rd cavity i

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1-2: '

184-2nd cavity ll2-3rd cavity .

The readings indicate the seals are staging properly.. _

Based on the above performance, BW sees no concern which would justify mainten-ance at this time.-

Further Testing to be Done:

1. During heatup, contact BW (E ". S ';'m m C'. C. L le ever TICo plans to start a RCP, so additional data can be taken at BW's discretion.

2. At system pressure > 1300 psis, 3 pumps running, data vill be taken on all four pumps. .

CCE:nif

• 10/7/77 e .

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, 10/13/77 t

' STATUS OF CECK0tTf 0F EACTOR COOLANT PU195:

t than 1300 psi.

ld pu=p starts.

All four EC Pu:nps have been run at system pressure Below is a typical line of data from each pump.

RCP 2-1 .

- 1650 psis System Pressure 2nd Seal Cavity .500 psis Pressure - 103h psig 3rd Seal Cavity Pressure - t Horizontal Vibration - 3 mils .

ECP 2-2

- 1650 psig -

System Pressure 2nd Seal Cavity Pressure - 1075 588 psispsig 3rd Seal Cavity Pressure -

Korizontal Vibration - 3.5 mils

'RCP 1-1 1650 psig - -.

System pressure -

2nd Seal Cavity Pressure - 1056 540 psigpsig" 3rd Seal Cavity Pressure -

4 mils ,

Horizontal Vibration ,

RCP 1-2_.

System Pressure - 1650 psig 2nd Seal Cavity Pressure - 920 psig 3rd Seal Cavity Pressure - 520 psig Horizontal Vibration - 3 mils fee t' hat all four pumps are in iodiogood operating monitoring.

Based on the above data, condition and require nothing more at this tim s

Inif . .

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DB 1 CORE -

AtlALYSIS OF SEPTE!!BER 24 DEPRESSURIZATI0fl EVEtlT A more detailed analysis was done to assess core thermal ~ conditions during the September i 24 depressurization event at Davis-Besse 1. Core conditions were analyzed to (1) i determine if steam was produced in the core, (2) determine the maximum internal fuel rod pressure during the transient, and (3) detennine if maximum lift force exceeded

the limit. -

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CORE C00LAtlT C0flDITI0flS .

Attachment D.2 shows transient thermal conditions as monitored by the reactimeter.

The system pressure is measured at the pressure tap, which is approximately 65 feet above the top of the core. The RC pressure at the top of the core is approximately 50 psi higher than the measured pressure because of unrecoverable and elevation pressure losses. As shown in Attachment D.3, the predicted core coolant temperature is slightly higher than the minimum saturation temperature (based upon measured

pressure), however, there is some uncertainty in both the measurement and the prediction, therefore, it is possible that some vapor bubble formation (steam bubbles in water) could have occurred within the core. An examination of.the reactimeter data (attachment D.4) indicates the' the RCS pressure level was near the saturation pressure for les's than one hour and that during this time period the pressure oscillated with a variation of + 50 psi. Therefore, the maximum time period during which the core could have been subjected to bubbly flow was less than one hour. Appred="cly E -

fjftee" '"h"tes after-reactor trip the coolant temperature dropped below the minimsdia estimated satutation +aieraterr, therefore,- the bubbly-flew r f~it =i:ted :t all,C J:cmed-for, c are th= t;n -" utes. If bubbles were formed during this period, the formation would be in the liquid as well as on the surface, as opposed to formation from a hot surface. With the temperatures, time duration, and type of formation, no significant effect on the components would be predicted.

i FUEL R00 PRESSURE

• prior to the depressurization event the reactor had been operating at 15% power for approximately one week. Inunediately prior to reactor trip the power level was 9% of j

rated power. The core burnup was 1 EFPD, therefore no signficant fission gas -

5 nimum inithl b:::'.;f ui pressure-production had occurred and nong was released.

Af-th4 #"al m: iSEp: M st 70 F. During the 60 minute time period in which the indicad RCS pressure was estinated to vary from 900 tg 1000 psia at the top of the core the average coolant temperature was less than 540 F and no significant heat generation occurred in the fuel. An initial evaluation (Waranc87 had predicted tensile stresses in the cladding based upon a maximum pressure differential across the cladding of 200 to I 300 psi. This evaluation had been based upon a BOL TAFY analysis with an arbitrary l safety factor added to ensure that actual conditions would be bounded by the prediction.

I A more recent analysis, again using TAFY, has resulted in a predicted maximum internal l fuel rod pressure of 1000 psia. This analysis considered as-built fuel gproperties and hot, near zero qWer conditions at a coolant average temperature of 540 F. On the i

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r e basis of this analysis it is concluded that the fuel rod cladding was not subjected to any significant level of tensile stress during the subject depress 0rization event.

! Since the cladding was not subjected to a large, long term tensile stress, no signi ficant long term effects on the cladding resulted. The tensile stresses which could have

_ occurred would have little effect on the cladding due to the small stress level and the

'short duration of the tensile stress.

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CORE LIFT . ,

6 Assuming a coolant teaperature of .537 F and 150 X 10 lb/ min system flow '(per Attachments D.5 and D.6) the net lift force will be less than 375 lb. ; The maximum allowable lif t force is 472 lb., therefore fuel assembly lift-off 4: . ;,t p;;dictes

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