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Provides Writeups to Substantiate Conclusions BWT-1578 & 1579 in 771005 & 07 Ltrs,Re 770924 Depressurization Event. Categories in Write Ups Listed
	ML20024C580
Person / Time
Site:	Crane, Davis Besse
Issue date:	11/01/1977
From:	Lauer J, Lazar A; BABCOCK & WILCOX CO.
To:	Domeck C; TOLEDO EDISON CO.
References
	TASK-*, TASK-03, TASK-06, TASK-3, TASK-6, TASK-GB BWT-1589, GPU-2012, NUDOCS 8307120833
	Download: ML20024C580 (32)
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P.o. Bos 126o.Lynctm g.v'a.2< Sos Te m (so w s111 November 1, 1977 RfCEIVED BWT-1589 NOVO a 1577 "I'*

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POWER ENG.

J. D. Lenardson w/a cc:

J. C. Lewis D. J. DeLacroix Mr. C. R. Domeck P. P. Anas/4c w/a lluclear Project Engineer E. C. Novak/lc' w/a Toledo Edison Company Power Engineering & Construction 300 Madison Avenue Toledo, Ohio 43652

Subject:

Toledo Edis'd Company REPORT OH OI. PRESSURIZATION EVENT Davis-Besse Unit 1 B&W Reference NSS-14

Dear Hr. Domeck:

By telecon of October 10, you have requested B&W input for a report to NRC regarding the depressurization event of September 24. The HRC exit interview notes dated October 7 summarize <i the necessary content of the report. B&W is providing write-ups in the following arus in order to substantiate.the conclusions of BUT-1578 and BUT-1579 dated October 5 and 7:

A.

Description of the event B.

Evaluati'on of the reactor coolant componer.ts C.

Evaluation of RC pumps D.

Evaluation of the fuel In order to expedite submittal of your report, we are sending Sections A, C and D at

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this tine, as agreed in our telecon of October 24. We e.xpect to forward Section B by November 7, and we will try to improve on this date.

Section A describes the sequence of events as reconstructed from computer alarm print-out, reactime:er plots, and. control room recorden (Attachment A.1). We have attar.hed A4) and reactimeter plots of R6 yin RC pressure, pressurizer level (Attachments A2, A3 pertinent recorder charts of T let temperature, RCS flow in each loop, RC pressure, pressurizer level, and water level and outlet pressure of each steam generator (Attach-ments A5 through A1.3).

Section B will include evaluations of stresses in the pressure boundary, the depressuriza-tion transient, boiling the SG dry, jet impingement or the SG, and effect upon fatigue 1

life.

l TEC c28s The Babcock & Welcos Cornpeny / Establisheiile67 ~ ~

8307120833 771101 PDR ADOCK 05000289

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Babcock;Wilcox Noven6er 1,1977 BWT-1589 Section C explains the evaluation which was perfomed to verify that there was no significant damage to RC pump bearings, seals, or impe11ers (attachment C1). The transient as it affected the pumps is surr.arized in Attachment C2. Attachment C3 defines the instrumentation and operational checks applied to the pumps. The results of the operational checks are tabulated in Attachment C4.

Section D evaluates the effect upon the core to detemine (1) whether steam was produced in the core (2) the maximum internal fuel rod pressure, and (3) whether maximum lift force exceeded the limit (Attachment D.1). Reactimeter plots are attached for reference Attachments D.2 through D.6.

Very truly yours, A. H. Lazar Senior Proje, Manager W

JAL/hj A. Lauer reject Manager Attachments TEC cass

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Sequence of Events The event started at time 21:34:20 on September 24, 1977. The plant was in Mode 1 with Power 06.T) = 263. The turbine had been shutdown earlier in the evening 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 Feedvater Rupture Control System (SFRCS) was initiated by an unknown cause. This trip shut the startup feedvater valve to #2 steam generator and stopped all feedvater to this generator (because of the low power level the main feedvater block valve was already shut, isolating the =ain feedvater control valve).

The lov level alam was reached in #2 steam generator at 21:34:44. Before the operator could identify and correct the problem, the lov level in #2 steam generator produced a full trip of the SFRCS. This trip shut the main steam isolation valves and feedvater isolation valves in both steam generators (time 21:35:18). STRCS also started both auxiliary feedvater pumps. The number one pump perfor=ed as in-tended, however, number two amr414mry feedvater pump only came up to 2600 RPM, in-sufficient to feed its steam generator (#2).

The loss of feedvater, first to one and then both steam generators, caused an increase in primary water temperature, which resulted in an increase in pressurizer level and thus reactor coolant system pressure. At 2255 PSIG the pressurizar electro-metic relief valve received an open signal. During the next 40 seconds, it received nine different open and close signals. After one of those signals the valve stuck-open. This provided a continuous 2h" vent path from the pressurizer to the quench tank. When pressurizar level got to 290", the operator manually tripped the reactor (time 21:36:07). Energy escaping from the electromatic relief valve and three main steam relief valves caused a rapid cooldown and depressurization of the reactor coolant system. Reactor coolant system pressure dropped to 1600 PSIG (time 21:37:17) initiating the Safety Teatures Actuation System (STAS). This started high pressure injection and closed numerous containment isolation valves, including the quench tank cooling lines.

With the electromatic relief valve still 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 tore off approximately a 10' high x 20' circumferential section of insulation from #2 steam generator. The paint from the then exposed area of the steam generator was blasted away. The steam in the containment also resulted in two fire alarms (one near RCP 2-2 and one near the pressurizer) and a single channel RPS trip on high reactor building pressure (4 PSIG).

When the main steam relief valves reseated the decrease in reactor coolant system temperature stopped and the high pressure injection pumps started to raise pressurizer level. At time 21:40:34 the operator stopped the high pressure injection pumps.

(The operators had been heavily involved before this time in regaining seal injection flow to the reactor coolant pumps. This flow had been stopped by the STAS 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 pressu'izer and pressurizar level r

vent off scale high at 320 inches. During this level increase the operator, seeing average reactor coolant system temperature and pressurizer level increasing, stopped one reactor coolant pump in each loop (time 21:43:11).

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Due to decreasing pressure in f 2 steam generator, the STRCS system gave a low pressure block permit signal at time 21:48:33. This alerted the opeator to the lov level and feed condition of f2 steam generator. He blocked the low pressure trip (time 21:49:38), took manual control of the speed of f2 auxiliary feedwater pump and fed #2 generator (time 21:50).- The operator saw the rapid addition of cold feedwater dropping the reactor coolant system temperature and stopped the feedvater addition to this generator.

j At approximately 21:55 the operator shut the block valve for the electromatic relief valve on the pressurizar and stopped the venting of the reactor coolant system to the quench tank. At 22:05 pressurizar level came back on scale. At 22:15 the cperator started a second makeup pu=p to try and st'op the pressurizer level decrease.

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 pressurizer heaters. At 22:23 the operator started a high pressure injection pump to try and stop the decreasing pressurizar level.

The level and 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 pressurizer level well on its way to recovering,the operator stopped the high pressure injection purp (time 'e2:27:44). 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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c ATTAutMENT C.1 RC PtkiPS 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 pt.9ps, a series of cperational checks were performed as outlined in Attachment C.3.

The results of the operational 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 to the pump cogonents. B&W finds the pugs to be serviceable for sustained full operational conditions with no inmediate 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 cil peak-to-peak amplitude.

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,r ATTACIUtENT C.2 RC PUP 95 SEPTEMBER 24 TRANSIENT 9 DB-1 All four RC pumps were subjected to the following:.

0:00 Reactor trip 1:10 SFAS trip 1:12 Seal return valves shut for 1:16 1:13 Seal injection valves shut for 1:52 all four pumps operated for 1:15 with no seal injection and no seal return flow during an RCS de-pressurization 2:28 Seal return valves open 3:05 Seal injection valves open

6:00 Steam fonnation pressure oscillating near P for $30 to 45 minutes 36:07 Total seal injection flow low aladAT Pug 1-1:

7:04 Pump tripped 7:45 Shaft stopped 36:07 About one minute of low seal injection flow (near 2 spm) flow imbalance starved seal injection 36:30 Seal return valve shut 1:12:55 Standpipe level high 1:17:07 Standpipe level normal Punp 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 9

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... m'?,r,tph' g5 C..'$() I / CII1W0tTf OF REACTon C00LtJrf ptscs. ~3 t PURP0 2: Assess whether maintenance is required of RC pumps as a result of abnormal Operational checks vill be required'to demonstrate transient of 9/2Is/77. that no significant damage has occurred to the pu p bearings, shaft and First series of tests vill to performed in Mode 5 due to operational se als. Later on operationc1 checks vill be perfomed in Mode restrictions. by !!RC. Each pu:np vill be operated individually for a duration not to exceed 3ten (10) =inutes, providing all defined parameters remain within limits established in this procedure. Operational sequence vill be as folicvs: d Torque values

1. Lift punps vill be started and pump shafts rotated by han.

are not to exceed 200 ft-lbs. A stethoscope vill be provided to detect (This has been satis-any unusual mechanical noises in seal housing area. factorily co=pleted on 10/3/77).

2. Mode 5 testing 225 psis.

Instrumentation R,: quired.; see attached (1A). 2.1 2.2. Co=puter Data - Printout NSS special su= nary trend for running RCP every 15 seconds. 2.3 Following li=its shall not be exceeded: A. Shaft vibration - 15 inills peak to peak. Total standpipe Icakage (upper seal leakage) plus seal return should 3. not exceedo.6 syn. If, during the test this limit is exceeded, the possibility exists of an open seal. In no esse vill total seal leakage be allowed to exceed 1.5 spm. If this met is exceeded, maintenance vill be required before further pump operation. All other nor=al plant limits and precautions prevail. C. Sequence of Operation: 2.16 A. Secure standpipe flush. Establish seal injection in accordance with plant operating procedure. 3. l Measure and record standpipe leaksge and return flow, confirm that. C. total leakage limits'are not exceeded. i l Assure coe=unicatien betvcen contral room and personnci stationed D. at RCP st, andpipe lesknce drain linc. E. Coudtdownfrem10to0 TECcada I Start strip chart recorders at hi;?t speed; Start pcactor Coolteit int =p O-2 Jt: n.-.'rdance with pinnt op. procedure. i After approx. 11 cce., reduce n 4 r. hart speed,. o .~...-

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O e F. Run pump fcr two (2) minutes unless any abovs. Limits are c7.ceeded. G. Data td.en vill bc assessed by DIAT and B-J representative. H. Following assessment of data, pump may be run for an additional five (5) minutes to allow for venting procsdure requirements. I. Follow above sequence on 2-1,1-2' and 1-1. J. Assessment of this data vill determine whether any maintenance is required before higher pressure operation is allowed. .:F/ Mil.48 3. Abo e test vill be repeated with system pressure at Srcater than 1300 psic before final deter:sination on condition of the pu=ps is completed. e CCE:nif 10/5/77 e 4 O e e 8 e I e e e e

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i 1. Upper and lower cavity pressures - all four pumps. 2. Both horizontal B/N Vibration Probes - al.1. four pu=ps. 3. WR System Pressure or suction pressure. Vertical probe on 2-2 pump. 5 Standpipe lenhage vill be collected and measured during the test. All of above should be recorded on an 8 channel brush. recorder located NOTE: in the' control room.

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.=.. - - s =- ~ - -. ->.w-- ,s. g s- /7[7*7 f STATUS OF CEECK01/f 0F REACTOR COOLANT PUMPS: /* yg 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 hu (2 min. ): System pressure 225 psig 3rd Seal leaka6e plus

- gEP" 2nd Seal cavity pressure 165 psig seal return flow 3rd Seal cavity pressure 123.9 psig Borizontal vibration 5 - 7.5 mills Vertical vibratica.25 milia After the two minute run, the pump vas run for ten minutes for system venting.

About 30 seconds before the p.=p was shutdown, there was a step increase in de'"S#u vibration to 2.5 mills. The pump was ran again on 10/6/77 for 10 minutes to checkout this phenemenon. Me vertical vitration was again.25 mills until about 5 seconds before shutdown where it increased to 2.5 mills. To allow a longer run time, 2-1 and 2-2 pumps vere run together for 10 minutes, then 2-2 vas run.alone for 10 minutes. he vertical vibration stayed at.25 mills for i the entire run. his vill continue to be monitored during pump runs for plant heat up. RCP 2-1 _ System pressure 225 psig 3rd Seal leakags plus ,,g, 2nd Sesi cavity pressure 132 psig return flow 3rd Seal cavity pressure 70 psig

Ecrizontal vibration 5'- 7 5 mills RCP 1-2 System pressure 225 psig 3rd Seal leakage plus

,*g EP" 2nd Seal cavity pressure k0.29 psig return flow 3rd Seal cavity pressure 81,3 psig Borizontal vibration 5 - 7.5 mills RCP'l-1 System pressure 225 psig 3rd Seal' leakage plus ,*g EE" 2nd Seal cavity pressure 77.98 psig return flow 3rd Seal cavity pressure 89.27 psig Horizontal vibration 5 - 7.5 mills 1 G TEC caos essee o e g ,y e .ep--- -- e'w.eA*h----A te 4-* e--= -*-9------- ^

~ .w... a,~ ... w.,... -. u.~. @HavT'(,kp4 ..,.'*'t.- ~ Se 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. he gauges read as follows: 1-1:* 184 2nd cavity 111-3rd cavity 1-2: 184-2nd cavity 112 3rd cavity he readings indicate the sesis are staging properly. Based on the above performance, BW sees no concern which vould justify mainten-ance at this time. Further Testing to be Done: 1. During heatup, contact BW whenever TEC0 plans to start a RCP, so additional data can be taken at B&W's discretion. 2. At system pressure > 1300 psig, 3 pumps run=ing, data vill be taken on all four pu=ps., Q e. CCE:nif 10/7/77 l 4 g S# e TEC cac7 o.._-. e 8 e 9 l

. L. ~ - [ %.. aemevrC4,p4, 4 .-===-.4 c STATtB 0F CEECK0tTf 0F REACTOR COOLANT PdMPS 10/13/T7 All four RC Pu=ps have been nn at system pressure greater than 1300 psi. RC Pumps 2-1 and 2-2 have continued to run from the initial cold pu=p starts. Below is a typical line of data from each pump. RCP 2-1 1650 Psig System Pressure 2nd Seal Cavity Pressure - 103h psig 3rd Sad Cavity Pressure -.500 psig Horizontal Vibration - 3 mils RCP t-2 1650 psig System Pressure 2nd Seal Cavity Pressure - 1075 psig 3rd Seal Cavity Pressure - 588 paig Korizontal Vibration - 3.5 mils RCP 1-1 System pressure - 1650 psig 2nd Seal Cavity Pressure - 1056 psig 3rd Seal Cavity Pressure - 5ho psis - Iorizontal Vibration k siis RCP 1-2 System Pressure - 1650 psig 2nd Seal Cavity Pressure - 920 psig 3rd Seal Cavity Pressure - 520 psis Horizontal Vibration - 3 mils Based on the above data, B&W feels that all four pu=ps are in good operating condition and require nothing more at this time than periodic monitoring. O R75: nit 10/13/T7 g. e e TEC eaos ~ 2 z_,

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. w.~ q- .? DB 1 CORE i .#(ALYSIS OF SEPTEleER 24 DEPRESSURIZATION EVENT + A more detailed analysis was done to assess core thermal conditions during the September i 24 depressurizstion event at Davis-Besse 1. Core conditions were analyzed to (1) determine if steam was produced in the core, (2) determine the maximum internal fuel rod pressure during the transient, and (3) determine if maximum lift force exceeded the limit. CORE COOLANT CONDITIONS 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 lesses. 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 (sten bubbles in water) could have occurred within the core. An examinaticn of the reactimeter data (attachment D.4) indicates that the RCS pressure level was near the saturation pressure for less 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. Approximately fifteen minutes after reactor trip the coolant temperature dropped below the minimum estimated saturation temperature, therefcre, the bubbly flow, if it existed at all, occurred for no more than ten minutes. If bubbles were famed 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. 4 FUEL ROD PRESSURE . Prior to the depressurization event the reactor had been operating at 15% power for approximately one week. Immediately prior to reactor. trip the power level was 9% of i rated power. The core burnup was 1 EFPD, therefore no significant fission gas production had occurred and none was released. During the 60 minute time period in which the indicated RCS pressure was estimated to vary from g00 to 1000 psia at the top of the core the average coolant temperature was less than 540 F and no significant heat l generation occurred in the fuel. An initial evaluation had predicted tensile stresses in the cladding based upon a maximum pressure differential across the cladding of 200 to 300 psi. This evaluation had been based upon a BOL TAFY analysis with an arbitrary safety factor added to ensure that actual conditions would be bounded by the prediction. A more recent analysis, again using TAFY, has resulted in a predicted maximum internal fuel rod pressure of 1000 psia. This analysis considered'as-built fuel properties and hot, near zero power conditions at a coolant average temperature of 540 F. On the 0 l l TEC caos . - --, : ~ *.,. : ~. :... ~ ~ ~ ~ '~ ~ ~~ ' "* *

.. -. e,..- 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 depressurization event. Since the cladding was not subjected to a large, long term tensile stress, no significant long tem 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. ~ CORE LIFT 6 Assuming a coolant temperature of 537 F and 150 X 10 lb/ min system ficw (per Attachments D.5 and 0.6) the net lift force will be less than 375 lb. The maximum allowable lift force is 472 lb., therefore fuel assembly lift-off is not predicted. S e G e e e e e e O Tga mio -M- .r. ~ -.. p,

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