ML18086B042

Revises Operator Requalification Training Program Incorporating Experience Review of 800611 St Lucie Event. Lesson Plan,Instructor Lesson Plan & Facility Study Assignment Sheets Encl

ML18086B042
Person / Time
Site:	Salem, Saint Lucie
Issue date:	01/20/1981
From:	Public Service Enterprise Group
To:
Shared Package
ML18086B041	List: ML18086B040 ML18086B042
References
NUDOCS 8111200661
Download: ML18086B042 (50)

Text

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SALEM OPERATOR REQUALIFICATIOX TRA!~ING S~3JECT: St. Lucie Event (6/11/80) r.:=:ss~:m ~a.

'I'Q? IC: Operating Experience Review

~!ISION ~O. DA7:S: _ l /20 /81

<'=""

'.C:RZNCES:

I. INPO 2/NSAC 16 Report II. PSE&G Salem Units #1 and 2 Precautions/Limitations/and Setpoints for Nuclear Stearn Supply Systems; Revision -7, August 1979 III. Westinghouse Instruction Book - Controlled Leakage Seal Reactor Coolant Pump Model W-11001-Bl e IV. Emergency Instruct.ions:

A. Safety Injection Initiation (I-4.0)

B. Blackout (I-4.9)

C. Loss of Component Cooling (I-4.15)

D. Partial Loss of Reactor Coolant (I-4.17)

E. Failure of a.RCP (I-4.20)

\/. J~)cr.ating Instruction!::\:

A. II-1. 3. 1 VI. Station Plant ~!anua.L A. Book 6 - Alarms 8111200661 811116 PDR ADOCK 05000272 V------- - - - -- ------- - -- PDR Page*l of 30

Analysis and Evaluation of St Lucie Unit 1 Natural Circulation Cooldown t

Page 2 of 30 I

-LESSON NO. NOTES L OVERVIEW A. Objectives

1. The student will be capable of discussing orally and/or in writing the three key elements significant to this event:

a. Unexpected formation of a steam bubble in the reactor vessel head.

b. Undetected loss of reactor coolant inventory.

c. I~strumentation unavailable, due to design or time constraints, but necessary for observable natural circu lation cooldown control.

2. Given a figure representative of plant parameters relative to time, the student will be able to recognize and identify the following evolutions:

a. Reactor trip;

b. RCP operation (restart);

c. Natural circulation;

d. Reactor vessel voiding;

3. The student will be able to list orally and/or in ~.;riting the requirements which authorize or require a trip of the reactor coolant pump(s).

• 4. The student will be capable of discussing orally and/or in writing the requirements necessary to start a RCP after tripping.

5. The student will be able to discuss orally apd/or in writing the following relationships concernin~ natural circulation:

a. Plant operational actions which enhance natural cir-culation;

b. Plant pararneter(s) that identify natural circulation in progress;

c. Plant parameter(s) that identify a loss of natural circulation;

d. Method of controlling cooldown rate while operating in natural circulation; PAGE 3 of 30

-LESSON NO. NOTES

e. Starting a RCP(s) while operating in natural circu-lation.

6. The student will be able to discuss orally and/or in writing the fol'iowing relationships concerning voiding within the reactor vessel during natural circulation:

a. Void formation in the head and not in the reactor coolant loops;

b. Methods to stop voiding once .recognition of the event has occurred;

c. Method(s) of void removal;

d. Plant response to voiding; hydraulic, mechanical, control systems, etc.

7. The student will be capable of explaining orally and/or in writing the following operational problems associated with a natural circulation cooldown:

a. Plant boration;

b. RCS chemistry analysis;

c. RCS inventory control;

d. Safety syst.em operations;

e. Condensate inventory requirements.

8. The student will be able to discuss orally and/or in writing the relationship of the St. Lucie event relative to Salem 1 and/or 2; including the initiating event, procedural guidance, and method (s°) of control.

B. Event

1. June 11, 1980, an electrical failure caused a component cooling water isolation valve to close. This valve closure forced a manual reactor trip in order to secure the opera-tion of the reactor coolant pumps .

. The loss of component cooling water to all four reactor coolant pump seals could have damaged the reactor coolant pumps if action had not been taken to secure their opera-tion in a timely manner.

However, without reactor coolant pumps, a forced convection flow cooldown could not be accomplished until plant conditions (temperature - pressure) allowed operation of the low-pressure shutdown cooling system.

PAGE 4 of 30

-LESSON NO. NOTES Approximately four hours into the natural circulation cool-down towa*rd plant conditions acceptable for the low pres-sure shutdown cooling system some unexpected behavior oc-curred; pressurizer level increased inconsistent with pri-mary system makeup evolutions. As the pressurizer cooled down, steam formed in the reactor vessel head when saturation conditions were reached, displacing water from the reactor coolant system and into the pressurizer.

Despite a steam bubble in the pressurizer and the reactor vessel head, natural circulation cooldown continued without any evidence of perturbation until forced circulation cool-ing was established with the shutdown cooling system (approximately six hours).

The initiating electrical failure was caused by an ac-cumulation of moisture in an electrical junction box con-taining a terminal board for the component cooling water valve solenoid.

II. Sequence of Events (on following pages)

Plant: St. Lucie 1 Utility: Florida Power and Light Company Location: 12 miles SE of Ft. Pierce Type: PWR - Combustion Engineering Capacity: 810 ME electric PAGE 5 of 30

e.

Sequence of Events St. Lucie June* 11, 1900 Plant Status Prtor to the Event:

st. Lucte was operatin~ at 99.6% reactor power with all the control element assemblies (CEA's) fully withdrawn. Four reactor coolant pumps (RCPs) 1-1ere operating~ The letdown and charging systems were in service. Reclctor Coolant System pressure was approximately 2250 psi and average temperattire was 563.4°F. A steam leak in the penetration room just outside of the containment building caused a fault in a 1:solenoid on a pilot actuated pneumatically operated valve. At 2:26 am, the plant experienced the failure (failed closed) of a single Component Cooling Water (CCW) return header e

isolation valve. The failure caused a loss of cooling capability to all RCPs, so a reactor power level reduction was initiated and the following sequence of events ensued.

"'CJ Cll (JQ ro Time Event Remarks and References CJ\

0 Hi 2:33 am Reactor trip The operator tripped the reactor. System responses w Turbine/Generator trip were normal.

0 Auxilary Feedwater pumps A and B on Steam Generator IA and 1B 1ow 1eve l

- al arms

• Feedwater pump 18 off 2:34 am Feedwater pump IA off Actions were taken to control steam generator water Feedwater pump lfi on levels. The stopping of the RCPs indicated that actions Reactor coolant pump lBl off were underway lo pl ace the reactor cool ant system (RCS) in natural circulation. Natural circulation is used to remove decay heat from the core \'1hen ncrs are 11nava i 1ab1 e 2:35 am RCPs IAI, 1B2 and 1A2 off These actions along wfth bleeding steam from the steam generators removed decay heat from the RCS using natural circulation 2:38 am RCP llH on The continued rate of increase in the RCS hot leg temperature prompted act ion by the operat.or to force

Time Event Remarks and References coolant water through the reactor core to remove decay heat 2:39 am RCP llH off Returned to cooling of the RCS usjng natural clrculat1on 2:45 am Feedwater pump 1B off Steam generator water levels were controlled from this time on by using Auxiliary Feedwater pumps lA and 1B 2:50 am Indicated pressurizer level returned *changes in loop temperatures, RCS pressure, and (approx.) to the control setpoint level 1:pressur1zer level following the initial transient had steadied 3:00 am Natural c1rculation cooldown A decreasing trend in loop temperatures with apnroximately (approx.) cornnenced 20 to 25°F AT indicates that a cooldown of the RCS us1ng natural circulation was taking place 3:12 am l\ux1l 1ary feed pump lC on w

o_

3:50 am The CCW flowpath was restored to all RCPs.

3:54 am Auxiliary feed pump IC off 4:00 am Conm~nced boratlon of RCS 4:35 am Pressurizer low pressure safety RCS pressure had decreased to approximately 1700 injection actuation pre trip signal psi. The cold leg temperature was indicating received. approximately 400°F. Cooldown of the RCS using natural circulation appeared to be normal 4:45 am Completed boratlon of RCS Shutdown boron concentrations were established in the RCS 4:45 am Conu1enced lowering steam generator lB This action was taken by the operators to make room level from 63% to 20% in the steam generators for the addition of a cold slug of

Event Remarks and References water during the later stages of the cool down process. The additfon of cold watet at that pbint ts used to speed the cooldown process when the amount of .steam produced by the heat transfer from the RCS to the steam generators has been reduced to low values. These low values slow the cooldown process when the temperatures and pressures in the RCS are slightly above the oper*at1ng temperature and pressure requited for shifting cooling from the steam generators to r: shutdown coo 1i ng using the LPS I system e Plant Status at Approximately 4:55 am The reactor was shutdown. Reactor coolant pumps in the loops were stopped. The indicated cold leg temperature (*380°F) and in cofe temperatures (412-415°F) were indicating a cooling trend. RCS pressure was approximately 1450 psi. Both steam generator water levels were being maintained at levels to support a natural circulation cooldown. Pressure in both steam generators indicated approximately 200 psi. CCW had been restored to the reactor coolant pumps.

w 0 5:00 am Coim1enced 1ower1 ng steam generator lA Reference 4:45 am Entry.

(approx.) Level from h65 to 20%

5:40 am Steam Generator 10-Level reached 20% With the MSIVs closed and vacuum broken (pressure at (approx.) Main steam isolation valves (MSIVs) atmospheric values in the condenser), heat transfer from closed - the steam generators to the main condenser was stopped.

Main condenser vacuum breaker open Pressure* in both generators indicated approximately 100 psi. Steam discharge from the steam generators was directed to the atmosphere through the atmospheric duinp valves.

6:00 am RCS pressure f ndicJtes 1140 psi lndfcatfons at this time show that cooldown using natural lncore temperatures indicate in the circulation was proceeding as anticipated range of 349-352°F Loop T11 indicated approximately 349°F Loop le indicated approximately 325°F 6:13 - Pressurizer level increasf ng Auxiliary spray into the pressurfzer was used lo decrease 6: 15 ain Loop T11 f ndfcatcs .. J46°F the RCS pressure ( 1140 to 690 ps f) from 06 :00 through (approx.) Loop Tc indicates u320°F 06:30. This cictton caused the format ion of a steam bubble

Time Event Remarks and References Incore temperatures indicate in the under the reactor vessel head, and resulted in an range of 342°F to 344°F unexpected pressurizer level increase 6:20 am Ste~m Generator lA level reached 20% Refer to the 05:00 and 05:40 entries (approx.)

6:35 am LPSI pumps lA and lB on LPSI pumps were placed in recirculation to warm up the Shutdown Cooling System piping in preparation for shutdown cooling using the LPSI system 6:40 am Pressurizer level increased significantly Jr The level change (46i to 963) and the rate of level increase changed siqnificantly from approximately 06:40 through 7:05 am 7:00 am Loop Tu indicates "325°F Plant parameters indicate that cooldown using natural Loop Tc indicates u305°F circulation was continuing, uninterrupted by the Pressurizer pressure indicates 500 psi steam bubble 1n the reactor vessel head Incore temperatures indicate in the range of 325°F to 327°F 8:45 am Steam Generator lA and lB levels Indicated levels in the IA and 18 steam generator~ were (approx.) increasing approximately 27% of the operating range. Indicated pressure fn steam generator IA and lB was <75 pst. F1111ng the generators with cold feedwater aids fn cooling down the primary system to a temperature compatible with initiating shutdown cooling using the low Pressure Safety Injection System. Refer to the 5:00 am and 5:40 run entries.

9 :40 am Steam Generator lA and IB (approx.) levels reach ~GSI 10 :00 am Letdown isolated Anomalies in pressuri~cr level response were continuing.

(approx.) It was reported that lpvel rose rapidly when auxiliary spray from the charging pumps was directed to the pressurizer followed by a r~p1d decrease 1n level when charging was directed to the loops. The letdown system was isolated by the operators to prevent draining of reactor

Time Event Remarks and References coolant from the reactor coolant system. Although this was the first logged isolation, Lhe operators reported that letdown was shutoff between 8:15 am and 8:55 am.

10:06 am LPSI pumps IA and IB off Wann up of the LPSI system was complete. Valve alignment was started to align the LPSI system for shutdown cooling 10:10 am Auxiliary Feedwater pump IA off ll One of the two runn1 ng auxi 11 ary feedwater pumps (IA & lB) was stopped. Capacity to meet steam generator level e

requirements was maintained with one pump

'"d PJ OQ 10:28 am LPSI system lined-up to the RCS LPSI syst~n pressure indicated -235 psi whf le RCS ro (approx.) pressure indicated 235 psi. Loop Tu indicated .. 290°F t--' while Loop Tc indicated ~275°F.

• 0 0

Hl 10:33 am LPSI pump In on LPSI pump discharge pressure spiked to a value 1n excess of w 400 psi. It was reported that the discharge relief valve 0 lifted resulting 1n the subsequent stopping of the LPSI pump lB .

10:34 am LPSI pump 10 off Reference 10:33 entry 10:37 am Pressurizer pressure - 230 psi llot Leg Temperature - 289°F Cold Leg Temperature - 273°F The parameters listed .\o1ere manually recorded fn the control room log. e 10:51 am LPS I pump lB on To establish shutdown cooling of the reactor coolant system using the LPSI system 11 :00 am Flw indfcated fn the LPSI system (approx.)

11:07 am Shutdown cooling established using (approx.) the LPSI system

Time Event Remarks and References 1.1: 27 am Auxiliary Feed\'1ater pump lB stopped As heat removal of the RCS had been established using the LPSr system, heat removal through the steam generators \'1as no longr.r "required 11 :35 am- Indicated LPSI system pressure LPSI system pressure was steadily decreasing from the 12: 23 pm decreasing pressure indicated at start-up (~255 psi) to an (approx.) indicated low value of *150 psi at 12:26 pm. Anomalous pressurizer level response was continuing in response to shifting the charging water entry point into the RCS.

(Refer to 10:00 am entry). Subsequent analysis of a

!f pressurizer level re~ponses during this period from 10:28 ~

am indicate a loss of RCS water inventory. Temperatures through this period remained relatively constant. Incore

'"Cl Pl temperatures indicated *281°F, Loop TH indicated -276°F OQ ro with the Loop Tc indicating ~275°F 0 12:23 pm LPSI pump lA on A marked increase in LPSI system flow and pressure t-h indicated the addition of inventory to the RCS from the w Refueling Water Tank 0

12:26.pm LPSI pump lA off 12:27 pm LPSI pump lA on Pressurizer level response indicated that the RCS system LPSI pump 1ll off inventory was being replaced 12:30 pm Suhcoo led conditions \'/ere reached 1n the pressurizer*

12:34 pm LPSI pump 113 on Shutdown cooling using the LPSI system was re-established. Pressure in the LPSI system was increased to

• 260 psi with the two LPSI pumps running. IA LPSI punv had supp 1i ed the t nventory and pressure needed in the ncs while the 113 LPSI pump and heat exchangers provided cooling in the RCS.

Time

• Event . Remarks and References 12:34 pm Recorded pressurizer level (approx.) indication went off scale on the high end.

1:30 pm Saturated conditons v1ere restored ;n The water in the pressurizer had been heated to saturated the pressurizer conditions through the use of pressurizer heaters.

1:57 pm LPSI pump lA off *The following plant conditions were recorded:

1: Pressure in the LPSI system was indicating uJ05 psi at a

• flow .rate of .. 3300 gpm. Loop Tu was indicating ~220°F, Loop Tc was *205°f, with incore temperature of .. 225°f.

Pressurizer pressure was recorded at 260 psi at a temperature of 410°f whith represent saturated condil ions in the pressurizer 2:50 pm Recorded Pressurizer Level indication Letdown from the RCS had been established at a flow rate 0

Hl (approx.) _returned to the indicating range -greater than the flo~1 rate of the charging pumps. This displaying a downward trend action wa~ taken to get pressurizer level in its normal w

0 recorded range 3:45 pm The downward trend indicated:on the

• Charging into. the ncs loops using bm charging pumps \'las (approx.) recorded pressurizer revel instrumentation. initiated *. Pressurizer level indication responded \'lilh an was stopped at an 1mticated level of 50% Upward trend *. This trend is anttcipatcd WhC!ll the HCS is full of water with a hubble in the upper portion of Lhe pressurizer. Prior to this time, charging inlo th1? loops brou9ht a decreasing pressurizer level indication signaling that steam existed at SOlllC other location in the nCS in addition to the steam in the pressurizer.

4:20 pm RCS degas if 1catf on procedure was commenced Plant status:

The plant had been shutdown and was being cooled usfn9 the LPSI system in the shutdown cooling mode of operation.

e LESSON NO.

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NOTES III. Natural Circulation/Plant Cooldown A. Reactor Trip The reactor was tripped manually at 2: 33 ~l. All the reactor coolant pumps were tripped by 2:35 A~!. One pump, the lBl reactor coolant pump was restarted at 2:38 A.'!, operated for approximately one minute and then stopped. All pumps (RCP's) remained secured to eliminate any risk of pump seal failure. Pump operation was not resumed after component cool-ing was restored; approximately one and one-half hours into the event.

B. Reactor Coolant Pumps The reactor coolant pumps were manufactured by Byron Jacks.on and have a three stage mechanical seal plus a vapor seal.

Sealing water is supplied from the reactor coolant system after being cooled by component cooling water. This heat exchanger is an integral part of the reactor coolant pump package. St.

Lucie reported that the seals did not experience excessive leakage during the plant cooldown but intermittent reactor coolant pump seal alarms occurred throughout the cooldown.

Here at Salem Nuclear Station, a total loss of component cool-ing flow would re:quire the following actions relative to the reactor and reactor coolant pumps:

1) Manually trip the reactor;

2) Within five.minutes trip all reactor coolant pumps, or if any motor bearing temperature exceeds 175° F.

C. Natural Circulation Natural circulation of reactor coolant and continued heat transfer from the reactor to the steam generators occurs as a consequence of the higher elevation of the steam generator re-lative to the reactor. A temperature differential becween* the hot and cold legs is required to sustain natural circulation.

The coolant flow can be estimated by comparing the measured temperature differential to the full power temperature differ-ential, knowing the decay heat at any given time after the trip The natural circulation cooling worked well throughout the cooldown using the steam generators. Temperature differentials between Th and T ld ranged from 20 - 40° F, compared with at co 44° F ~T during power operation. With core power at 1 - 2%

from decay heat, the temperature differential corresponded to a reactor coolant flow rate of 2 - 3% full flow.

The higher coolant flow rate under natural circulation occurred PAGE 13 of 30

e LESSON NO. NOTES soon after reactor trip, corresponding to the higher decay heat production. At about 7:00 AN, during the time that evidence of the first drawing of the steam bubble under the reactor head appeared, the temperature difference between That and Tcold was about 20° F. With decay heat just under 1%, the c6olant flow rate is estimated to have been near 2%

full flow. Throughout the time that the steam generators were used to cool the unit with a steam bubble under the head, coolant temperatures indicated that about 2% of rated flow was being maintained.

After shutdown cooling was initiated at about 11:00 AH, the steam generators continued to be functional for a few hours, assisting in the removal of heat from the reactor coolant system.

From the time of reactor trip until initiation of shutdown cooling (2:33 AM - 11:00 Aft), the coolant subcooling ranged from 67° F at the moment of trip to as high as 225° F at 5:30 AM, based on the highest core exit thermocouple reading or That instrumentation located in the hot legs. The That 0

instrumentation reached the low end of its range (515 F) at 3:20 AM and was not useful for about 2-1/2 hours, until it had been re-ranged by an I & C technician. Below 515° F, the.

core exit thermocouples were used to indicate the amount of coolant subcoolirfg. From 5: 30 AM, coolant subcooling fell slowly, reaching a low of 55° Fat 12:23 PM, just prior to raising coolant inventory and pressure through operation of Low Pressure Safety Injection (LPSI) pump lA in the safety injection mode.

At no time did the That instrumenation indicate a transfer of hot water from the reactor vessel head region into the hot legs as the steam bubble formed. However, flow from the upper

• vessel and head region into the hot legs each time the steam bubble formed was always a small fraction of coolant flow through the core and loops. Between 11:05 AM and ll:~S A~ hot water flowing from the pressurizer during a relacively low water

• level dip was indicated by the Th ot instrumentation which by that time was being recorded .

D. Steam Formation During Cooldown Cooldown on natural circulation, by feeding the steam gener-ators and dumping steam to the condenser, began at about 3:00 AM. Natural circulation of reactor coolant had been well established by the time the cooldown started. The cooldown progressed at an average 60° F per hour until about 6:00 AM.

Shortly after that, an attempt was made to cool the pressurizer and reduce pressure through the use of auxiliary spray from the charging pumps. Between 6:15 AM and 7:15 AN the water level in the pressurizer rose unexpectedly, much more than PAGE 14 of 30

e LESSON NO.

---

NOTES could:be explained by the volume of water being pumped into the reactor coolant system. The pressure at 6: 15 A~!. when the steam bubble apparently first started to form under the head of the reactor vessel, was somewhere between 1140 and 690 psig, the pressure log entries at 6:00 AN and 6:30 A..'! respectively.

Saturation temperature at 6:15 AM corresponding 0 -

to the mean pressure between 6:00 AM and 6:30 AM was 535 F, likely very nearly the temperature of the reactor vessel head and its contents at the beginning of bubble formation. The cold leg temperature was 320° F at this time, so a temperature differ-ential of about 200° F existed in the reactor vessel between the top of the flange and the coolant nozzles. From the ob-served level rise in the pressurizer, the coolant shrinkage from cooling, and estimated charging and letdown, the size of the first steam bubble expansion is estimated to have been somewhat larger than the hea~ volume down to the closure flange, but well short of the hot leg nozzle.

It is not possible to estimate reliably the inventory of reactor coolant during the cooldown since the charging flow indicator was out of service, starting and stopping of the charging pumps may not have been noted always, (the operator's written log had no entries on charging pump operation) and the letdown flow was variable and not recorded. However, from the hourly log entries, it appears the letdown flow, in response to the rapidly: ri:sing pressurizer level, may have been slight-ly higher than charging flow (auxiliary spray) during the drawing of the initial steam bubble. A slow loss of inventory apparently continued in response to higher than desired pres-surizer level until 8:15 AM, when letdown was reported to have stopped for 40 minutes. As many as 3 charging pumps may have been operated for several minutes during this interval, afford-ing an increase in coolant inventory. At 9:00 AH, charging was evidently back to the flow of one pump (about 44 gpm), and letdown flow was 30 gpm. The charging flow appeared to have continued .at 44 gpm past 3:00 PM. Letdown flow was stopped*

shortly after 10:00 AM, remaining off until sometime bet\*een 2:00 PM and 1:00 PM, when it was re-initiated at 3baut SO ~p~.

Between 9: 00 A.:.'1 and 12: 30 PM the charging flow was alternated several times between the reactor loops and pressurizer auxiliary spray, causing rapid pressurizer level transients, as the steam bubbles in the pressurizer and the reactor h~ad alternately swelled and shrank in opposite directions.

E. Inventory Loss At 10:28 AM the isolation valves between the reactor coolant system and the shutdown cooling loop lB (LPSI-lB) were opened, in order to start shutdown cooling. Pressurizer pressure was indicating 235 psi. It. was reported that the suction relief valve on the shutdown cooling loop lifted at this time, and the discharge relief valve lifted at 10:33 AN upon starting PAGE 15 of 30

ttLESSON NO.~~~- NOTES the low pressure safety injection (LPSI) pump lB. The LPSI )

pump was then stopped, *and initiation of shutdown cooling delayed until system pressure decayed another 20 psi.

Upon opening the valves to the shutdown cooling loop at 10:28 AM, the pressurizer level transient, which was then trending down, exhibited a distinctly different response than the pre-vious downward movements. Whereas the previous two downward legs of the pressurizer level transient indicated an exponen-tial decay of the steam bubble under the head, once the shut-down cooling loop isolation valves were opened, steeply fall-ing levels were experienced.each time the charging flow *was switche~ from pressurizer spray to reactor loop injection.

Since letdown had been shut off shortly after 10:00 AM, and reactor coolant average temperature was nearly constant f~om 9:30 AMuntil noon, a steadily rising reactor inventory result ing from the charging flow would have been expected. However, a generally downward pressurizer water level trend was experi-enced.

• Analysis of the pressurizer level and system inventory indicate that a reactor coolant leak was in progress. Further evidence of.this leak appeared on the refueling water tank level strip chart recording. From the RWT strip chart, the leak appears to have been about 5000 gal up to 12:23 P:-L Averaged over the time period from 10:28 AM to 12:23 PM, the leak.rate would have been about 45 gpm. However calculations of coolant invent;_ory over that time period based on pressurize level response, assuming letdown shutoff, charging in-flow fro.

one p'ump * (44gpm) and reactor coolant pump seal out-flow of. 4 gpm, gives an estimated leak flow of about 60 gpm. The net loss from the reactor coolant system was somewhere in the rang of 5 ~ 20 gpm. Evidently the transfer of reactor coolant to the refueling water tank occurred through a partially open valve in the recirculation line from LPSI pump lB. The orific in this line is rated at 50 gpm at pump shutoff head (200 psi).

Nqting that pressurizer pressure had decreased to 115 psi and pressurizer water level had repeatly dropped below the refer-ence value, low pressure safety injection pump lA, which h3d been aligned for safety injection, was started at i2:2J P~ in order to increase pressure. LPSI pump lB was turned off for s.everal minutes while pump lA was injected, lowering the pressure in the common LPSI pump discharge header, thus per-mitting LPSI pump lA to raise reactor pressure to the pum~

shutoff head, about 200 psi. About 5000 gallons of refueling water was pumped into the reactor system between 12: 23 Pi-I and 12:34 PM; at which time LPSI pump lB was re-started, raising the discharge header pressure well above the shutoff head of the lA pump.

At 12:23 PM, just prior to starting LPSI pump lA to inject water into the reactor, the steam bubble under the head had been allowed to decay to a small, but unknown size, while water flowed from the pressurizer to replace the decaying PAGE 16 of 30

e LESSON NO.

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NOTES bubble. From rough coolant inventory calculations, the steam bubble is estimated to have been no larger than a few hundred cubic feet at that time. Operation of LPSI pump lA, injecting water and raising pressure would have reduced the bubble size by half. Complete collapse of the steam bubble in the reactor vessel head is estimated to have occ-urred within minutes of the repressurization.

Once the pressurizer water level had been raised above 100%

(indicated) on the hot calibrated level recorded at 12:34 PM (above 67% corrected for temperature), the pressurizer steam space was compressed, and the water in the pressurizer was then subcooled. The pressurizer heaters were energized to raise the water temperature, so that the pressurizer could be used t_o maintain pressure above 200 psi. The heaters had raised the temperature of the water to saturation at 175 psi by 1:15 PM and continued to raise pressure, reaching 260 psi by 2:00 PM.

Pressure was then maintained near 260 psi for several hours, preventing any reappearance of a steam bubble under the reactor vessel head.

After the reactor inventory and pressure had been increased through operation of LPSI pump lA at 12:34 PM, and shutdown cooling had resumed using LPSI pump lB, the lA pump continued to operate in th~. injection mode. The lA pump was merely recirculating water back to the refueling water tank through its miniflow recirculation line, as the pressure in the dis-charge header was too high to permit further injection to the reactor coolant system. Leakage from the shutdown cooling system to the refueling water tank continued until 1:57 PM at which time the lA pump was stopped and the recirculation lines from both pumps were isolated from the refueling water tank.

From 12:34 PM until 1:57 PM, it is estimated that approximately 5000 gallons of reactor coolant leaked to the refueling water tank. During this time period the letdown was secured, and most likely one charging pump was operated continuously. The net rate of loss from the reactor system is estimated to h~ve been 20 gpm.

By 2:00 PM the pressurizer had been heated sufficiently to re-store the reactor coolant pressure to 260 psi, and the slow coolant leak had been stopped. The coolant inventory was now slowly increasing from continuous charging flow. It was re-ported that at 2:32 PM letdown flow was re-initiated at .:ibout twice the charging flow rate. The pressurizer steam space ex-panded slowly, bringing the water level indication back on scale on the recorder (hot calibrated channel) by 2:50 PM. At about 3:45 PM charging was increased to a higher flow rate ~han letdown, and the pressurizer water level exhibited a normal slow rise in response over a 45 minute period. Conditions had been restored to normal.

PAGE 17 of 30

-LESSON NO.  :.JOTES IV. 1977 ~atural Circulation Cooldown During 1977, the St. Lucie plant conducted a cooldown under natural circulation conditions. The possibility that a steam void had been formed under the reactor vessel head during that event had not been recognized. The pressurizer water level transient during that cool down has now been examined, and anomalous behavior similar to the recent transient has been confirmed. The cooldown prior to the time of the first rise in pressurizer level was about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> longer in 1977 than in 1980. Additional training and understanding which came as a result of the TMI accident, *enabled operations personnel to recognize the significance of the pressurizer level behavior in 1980.

The 1977 transient pressurizer water level response did not exhibit any evidence of a loss of coolant inventory during the event. Shut down cooling was not initiated until approximately 7-1/2 hours af te

._the first sharp rise in pressurizer level, compared with 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to initiation of shutdown cooling in 1980. Delaying the reduction of

.reactor system pressure to perinit initiation of shutdown cooling gave the reactor vessel head and its contents more time to cool,

. lessening the tendency to form a large steam bubble.

V. Radiological Effects At ' .:, 10:28 AM on 6/_11L80, after previously heating up the water in the shutdown coolin*g Tines, shutdown cooling loop lB isolation

• . va1ves to the reactor coolant system were opened in preparation for initiating shutdown cooling using the lB low pressure safety in-

.* jection )LPSI) pump. A gradual increase in the refueling water

• tank (RWT) level began at this time. The most probable pathway that reactor coolant took to reach the refueling water tank was through the lB LPSI pump mini-flow recirculation valve (V-3205) which may have been at least partially open while the system was operating in the shutdown cooling mode. This allowed the transfer of approximately 5000 gallons of reactor coolant to the refueling water tank between 10:28 AM and 12:23 PM at which time the lA LPSI pump was turned on to fill the reactor coolant system and repressur-ize ~ack to 200 psi. About 5000 gallons of refueling w:iter \v:::i.s transferred back to the reactor coolant system at this time. The lB LPSI pump was turned off at 12:27 PM and restarted at 12:34 PM and left on thereafter in the shutdown cooling mode. The lB LPSI pump mini-flow recirculation valve (V-3205) remained open during this time allowing the continued transfer of an additional 5000 gallons of reactor coolant to the refu~ling water tank until 1:57 PM when all mini-flow was isolated using the motor operated valves on the LPSI pumps mini-flow header outlet line (V-3659 and V-3660).

Although the lA LPSI pump remained in operation until 1: 57 P:-l, its head was sufficient to pump water from the refueling water storage tank to the reactor coolant system only when the pressure in the pump discharge line was below 200 psi. Once reactor pressure had been restored to 200 psi (~ 12:30 PM) and the lB LPSI pump restarted (12:34 PM), pressure in the common discharge line was too high to PAGE 18 of 30

. . -LESSON NO. ~OTES permit further return of coolant from the refueling water tank to the reactor.

Samples of the reactor ~oolant taken in this period (1:45 PM on 6/11/80) contained specific activity levels of approximately 2

10 x 10- microcuries/ml. The transfer of a total of lO,ooq gallons of reactor coolant containing specific activity of 10 x 10--

microcuries/ml to the refueling water tank represents the inad-vertent transfer of approximately 3.8 curies of activity from the reactor coolant system to the refueling water tank.

Plant vent gaseous, particulate and iodine activity levels remained steady before and throughout the transient with plant vent gaseous activity maintaining approximately 200-300 counts per minute.

Subsequent to 6/11/80, Florida Power and Light conducted a con-tainment purge and three releases from the gas decay tanks also without any abnormal releases of activity.

There may have been a small off-site* release of gaseous radioactive material from the refueling.water tank external vent due to the transfer of reactor coolant to the refueling water tank. A reactor coolant sample taken at 1 PM on 6/11/80 contained noble gaseous activity of l.35 x 10- microcuries/ml. I~otopic analysis of this 2

sample showed this gaseous activity to be composed of greater than 90% Xe-133, 8% Xe-135 ::with a remainder of (less than 2%) Kr-85. If as much as 10,000 gallons of reactor coolant had been transferred to the refueling water tank this would have resulted in the transfer of 0.51 curies -0f noble gas activity. The allowable environmental limit for unplanned or uncontrolled release of noble gas is ap-proximately ten times this amount.

A small fraction of the 0.51 curies of noble gas actµally transfer-red to the refueling water tank would be expected to have outgassed from the relatively cool refueling water tank prior to the decay of the xenon fraction in the tank to nonvolatile cesium.

VI. Conclusions A. Natural circulation decay heat removal and subsequent cooldown of the reactor coolant system were adequate.

B. Anomalous pressurizer level behavior occurred after several hours into the cooldmm when a steam bubble was formed in the reactor vessel head. This resulted from cooling the pressurizer below the temperature of the head area.

C. There is no temperature instrumentation in the reactor vessel head area. Such instrumentation could alert the operator of approaching saturation conditions in the reactor vessel head area.

PAGE 19 of 30

-LESSON 'NO. ~OTES D. Z..:onnal instrumentation for reading RCS hot feg temperature was not usable once the system was cooled below 515° F.

E. The subcooling monitor instrument is also not usable below 515° F since it receives its hot leg temperature input from a narrow range instrument.

F. Although wide range indicatfon was available to the operators, no automatically recorded pressure trending was available once system pressure was below 1500 psi. The operators nanually recorded reactor coolant pressures at 30 minute intervals durin the cooldown.

G. After the shutdown cooling system was valved into the RCS ap-proximately 10,000 gallons of reactor coolant was drained in-advertently to the refueling water tank at a rate of about 50 gpm.

H. Letdown system operations continued for a period of about three hours after the pressurizer level anomalies started.

I. At the time the steam bubble first started to form under. the reactor vessel head, there existed in the reactor vessel a temperature difference of about 200° F between the top of the vessel flange and_the coolant nozzles. This condition is not normally analyz.ed*-:as part of the ASHE Section III analysis.

J. A single malfunction in the common reactor coolant pump seal water cooling system led to the shutdown of all four reactor coolant pumps.

PAGE 20 of 30

S/G /G p*

R z

R Ir SIS (typical 1 of 2)

--tx:l--+ SDCS (outlet)

SIS (1 of 4)

SOCS (inlet)

e e e ~*

Air Air Air

~supply .--tti<l- supply supply 81 t-.....

.i i:rJ Seal cooler 1

Seal cooler Seal cooler Soal cooler

-~~

CJ 0

Ul 0 o-3 r-<

"d * ~ H

t> 1A1 1A2 1B2 z C)

C) r-< '"rj trl C:: H

(") G") "d NH C::

N trl :;d ~

t)j trj a trl /tJr r

<: tv tTl

• Hi trl supply ~

wz H' o~

z H

CCW supply t-:1 H

.~

HCV14*7 HCV14*1 ~

• . H

! z

-~

I Roactor (::'.

Building '* <:

[Tl

! ~

I Air 1-:1

"f supply ccw return I HCV14-6 I HCV14*2 L---J

DOUBLE BUBBLE (ARTIST CONCEPTION) 0

-C,!)

en '--~~~~~....,t-~~~~~~~~__.

FIGURE 3 ST. LUCIE EVENT PAGE 23 of 30

/ CEDM nozzle

/

• Instrumentation

~nozzle Outlet Upper guide Inlet nozzle structure l)OZZie

~-*

Active .

core Fu~I I assembly length Estimated Size of Initial Steam Bubble FIGURE.4 ST. LUCIE EVENT PAGE 24 of 30

' . \

600 T

.l r Tripr at 0233 I

Ttlot and TCOid*

T natural circulation

-< St. Lucie,6-11-80 580 I

560 f Ir Reactor coolant pump operated

....__ ... ~ ,,...... Ii~~rr~ ~~

I , L

~ rcr ~

I Thot

~ff scale instrumen/

~

>\

-~ I rTcold I

I

~Trip at 0233 "1 I 1/~ .. T

~ ,.

""t'o-500 *-* -~

" ~

480 I\

460 0 5 10 15 20 25 30 35 40 ..45 so 55

\ 60 nme After }rip (minutes) 2:33 2:43 2:53 3:03 3:13 3:23 3:33 Time of Day (a.m.)

Reactor Coolant Temperatures First Hour After Trip FIGURE 5 ST. LUCIE EVENT PAGE 25 of 30

2300 rr~,-i--r-1""T'"9--r1-r-~,:--"T"""'"-r1-r--~1~-r--or--l-,---.l--.....--..--l...,.......__,l-......--.---.--...---.

• la,t 1 I I I 1 1 , 1 ~  :

22CO H-~,,f!~7t--t\-~,-+-~-+--+--~-'--t--~--11--~.-~--l.--+--+,-~--+-+---!--.L.-~.-1 2ico ~.--r.n--:---r-~~~~,--:-~~+--~1--+-~':..._+-~'--+-_:_1-4~.1-'-+-.......!.1--1__~1--l.-~'---J

.., I ~i I '\(\I I* ,....,-..,erating

. . . . . . range I I 1 I I I I I

' 000 Ii ', ~ l I pressure recorder, T 1

I I,  ; I 1 1 I 1 1900 _.__._____ reference 62 -i---+-_.._--+--+--+-.._-+-.i.-.'-+--.i.-.'-1-.......11_~-J.'-~

1 I I I I I I I I I I I I I 1800 __

~,~,r--+-~~r--.,_~,~+-~,-+~+,--1---~1+-~-+---,--~_.__.....___.,~-----+---~

, ~r., I I l I

~

~= :1==:1==:===i==~:~~::1==:==:,*==:==~,*==:==i~====::==:==~;==:==~:==:==:;==:==~*l==~1==i~=~

I I I ,1 I i-;--+-,--+--1 --+--1~..+-_1-* Lower limit of pressure recorder I I ' ' I I I  ! I I 1

1500 1 1 1 1 I

. .9: 1400 I I .... I ' I ' ' I I I

~ '- r I

~ 1300 1+--:-1-i--~'--ir---~'-~~~'--~~l--+-~'--+-~'---+--l=---+--"--'-+-.....;..l__.___,~~~' -

UJ I I I I\ I I I I I I I I I

£ ,200 I I I \! I I I I f I I I c 1100 I I I ~ I I 1 I I I I I

-a I I I ~ I I 1 I I I I I 8 1OOO I I 1* 1 I\ 1 I 10 Reference 1, shutdown log, - - -1.--.

- 900 ----*---*-.....__,.._......___ ._..__..._._..__...._.......__.._wide range pressure indicator_-+---+-~

0 I I ' I \ I I I I

• I I I

• I

• I I g

£ 800 II1 1 *1 '

I 1 **

I ~

1 I

1 I ~

' Between 1 2:00 noon and 12:30. reactor_

pressure estimated from shutdown 1

700 '""--.,--+---.1--+--1 600 1 I I I

-+--- 1

<...'-\-+-

1 I

- i - - -.---+--"-I -~ooli~g 1

I I syste~ press~re reco~der I I I 500 I I I I .:;*: I '~  ;  : ""  : I I I I I I 1'...._ I I I~ 1 I I I 400 l-"----+--+--+~-+---+-+--+--!---+--.::t:).,,.--+-~-...;.--+----+-'-*....--4--.+--4--+--+---1-----I 300 1-+--+:--+-_.:~+--+:-+--.j.-:-!--.j.-!~~~.....__c?-1----=-+-........z~l~:::--J.:__-+-_~~*---.....~_.:_,__~;-l-__ :l.--l 200 1 1 1 1 t 1 I "t: --o- "'(' ~ j . . "'-... 1 I ,.,__ er -<

H--~1~+-~,~f--~,--+~lr--+-~,~f--~,--+---:1'--+-~1__::~~1-~~--~~~--......-}-<'>-;'~.-o;:-;=....~,*~

I I I I I I I ~I I 100 14--+---+--+,--+--+,-+--!-,-!---.j.-r-f---i1!---+---.,~-+--+-_._-+-,-+--+,..:=....~-+--+-~,~---I 0 u....----'--'--'.._~1.__.i...-~'--'~~'--'~._1..._~1~..__~1~.._~*~.__~1~~.i.-'~-'L--.0 0 1 2 3 4 5 6 7 a 9 10 11 12 Time After Trip (hours)

I 2:33 3:33 4:33 5:33 6:33 7:33 8:33 9:33 10:33 11:33 I 12:33 1:33 2:33 Time of Day a.m. I p.m.

Reactor Pressure FIGURE 6 ST. LUCIE EVENT PAGE 26 of 30

Natural circulolion cooldown ot St. Lucie 6* 11 *80 Jr 400 ~~~-n-~~--+~~~-1-~~-t-~~-1-~~~l--~~-t-~~-t-~~~f--~--i CL' L _,,,- Thol (lrom coro exit thermocouples)

~

E 1--~~-t-~~~+~Uc:-.,;_.1--~~-t-~~-1-~~~f--~~4-~~-+-~~--t~~---1 e '350 8-E Cl>

~ 300 --~~....-~~--~~-r~~-ci.c._-~---',-.~:---t-~~-t~-=-~t-~~-t-~~-t 250 l--~~1-~~+-~~+-~~t--~~1--~~1~~--t~-LoopB Loop A../

200 1 2 3 4 5 6 7 8 9 10 11 limo Aller Trip (hours) 3:33 4:33 5:33 6:33 7:33 6:33 9:33 10:33 11:33 12:33 1:33 Time ol Day a.m. p.m.

Rooclor Coolanl Temperalures During Cooldown

J I

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Oua:uad) (8/\91 JazµnSSoJd C\I FIGURE 8 ST. LUCIE EVENT PAGE 28 of 30

. Most probable reactor

~coolant leak path V*3464 --..--:>~ Loop Alr 1A2 supply

.-------=111o Loop 1A1 "B" LPSI pump

.--..---_.Loop 161 V*3400

...__ _ _ ___,.Loop 182

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0 1 2 3 4 5 6 7 0 9 10 11 12 13 Time Allor Trip (hours)

Pressurizer Level. 1977 SI. Lucie Natural Circulnllon Cooldown

SUBJECT:

St. Lucie Unit 1 Event (6/11/81) LESSCN ~'O.

'IDPIC: Operating Experience Review/Licensed Operator LXJRATICN: 4 hrs.

Requalif ication Training ------

IV. Emergency Instructions:

DiSTRGCTOR REE fil0iCES: A. I-4.0 Safety Injection Initiation I. INPO 2/NSAC 16 Report B. I-4. 9 Blackout II. PSE&G Salem Units #1 & 2 C. I-4.15 Loss of Component Cooling Precautions/Limitations/ & Setpoints D. I-4.17 Partial Loss of Reactor Coo ant for Nuclear Steam Supply Systems; E. I-4.20 Failure of a RCP Rev. 7/August 1979 V. Operating Instructions:

III.* W Instruction Book - Controlled A. II-1.3.1 Leakage Seal RCP Model W-11001-Bl SEE BELOW FOR VT VI'~ Station Plant Manual A. Book 6 - Alarms I. Lesson Plan II. INPO 2/NSAC 16 Report III. Transparencies/Overhead Projector IV. Chalkboard/Chalk I. Student handout package II. References A. Emergency Instructions (applicable)

B. Operating Instructions (applicable)

C. Precautions, Limitations and Setpoints (applicable)

D. Alarms (applicable)

ILP.SSCN 7.;._;..N P.E\iISIC'N NO *_ __;(c.:....)- - - - 1//0/81 TI'l'LE: SR. *Nuclear Cu....--ricul1=.1. Supervisor Date 01/11/80 T - ;

Bev. 1

LESSON PL.\:.'l LESSON ~~O.

OUT!..I~IE .:\IDS INTRODUCTIO~

I. Objectives A. The student will be capable of discussing orally and/or in writing the three key elements significant to this event:

1.

• Unexpected formation of a steam bubble in the reactor vessel head.

2. Undetected loss of reactor coolant inventory.

3. Instrumentation unavailable, due to design or time constraints, but necessary for observable natural circulation cooldown control.

B. Given a figure representative of plant parameters re-lative to time, the student will be able to recognize and identify the following evolutions:

1. Reactor~trip;

2. RCP operation (restart);

3. Natural circulation;

4. Reactor vessel voiding.

C. The student will be able to list orally and/or in writing the requirements which authorize or require a trip of the reactor coolant pump(s).

D. The student will be capable of discussing orally and/or in writing the requirements necessary to start a RCP after tripping.

E. The student will be able to discuss orally and/or in writing the following relationships concerning natural circulation:

1. Plant operational actions which enhance natural circu-lation.

2. Plant parameter(s) that identify natural circulation in progress.

De. t:e 1/20/81 Page 1 of 16 rteV. 0

LESSON ?L..:l.i.~ LESSON NO.

OUTLI~IE .:l..I!JS

3. Plant parameter(s) that identify a loss of natural circulation.

4. Method of controlling cooldown rate while operating in natural circulation.

s. Starting a RCP(s) while operating in natural circula-tion.

F. The student will be able to discuss orally and/or in writing the following. relationships concerning voiding within the reactor vessel during natural circulation:

1. Void formation in the head and not in the reactor coolant loops;

2. Method(s) to stop voiding once recognition of the event has occurred;

3. Method(s) of void removal;

4. Plant. r~~sponse to voiding; hydraulic, mechanical, control systems*, *etc.

G. The student will be capable of explaining orally and/or in writing the followingoperational problems associated with a natural circulation cooldm..m:

1. Plant boration;

2. RCS chemistry analysis;

3. RCS inventory control;

4. Safety system operations;

5. Condensate inventory requirements.

H. The student will be able to discuss orally and/or in writing the relationship of the St. Lucie event relative to Salem Unit 1 and/or 2; including the initiating event, procedural guidance and method(s) of control.

Da ~e 1/20/81 Page 2 of 16 ~ev*. 0

LESSON NO.

.:;,.: D S II. Reason(s) for Study A. Review and discussion of Nuclear Industry Operating Experiences can be utilized for the development of licensed operators; especially, those experiences which involve fundamental principles and/or plant design problems applicable to Salem Unit #1 and/or #2.

LESSON PRESENTATION I. Overview A. Objectives B. Event

1. Basic Combustion Engineering plant T/P-1

2. Initiating problem T/P-2

3. Basic plant response

4. Natural circulation cooldown

5. Vessel head voiding T/P-3 & 4 II. Sequence of Events (SOE)

A. Plant status (pre-event) T/P-5 B. SOE from 2: 33 AM to 4: 45 AM T/P-6 C. Plant status (4:55 AM) T/P-7 D. SOE from 4:45 tu'1 to 4:20 PM T/P-8 TTT

.!. .L .1.. Natural Circulaticw./Plant Cooldc~..,"TI A. Rx Trip (manual initiation)

B. RCP's Tripped (manually)

1. Guidelines for RCP trip at Salem Nuclear Station

a. EI-4.20 Da'::e 1/20/81 Pase 3 of 16  ;:{ev. 0

LESSON PL;....:..'1 LESSON NO.

OUTLJ:~1E .:\.!!) s

b. OI-II-1. 3.1

c. Precautions, Limitations and Setpoints

2. Basis for RCP tripping

a. Problem - Byron Jackson T/P-2

b. Purpose C. Enchancement to Natural Circulation T/P-1

1. Steam generator level

2. Pressurizer level

3. Sub-cooling margin

4. Loop fJ.T D. Natural Circulation Recognition T/P-1

1. Parameters

2. Expected parametric values

a. 'liT.

b. Flow E. Steam Formation During Cooldown T/P-3 & 4

1. Steam bubble/space formation

a. Rx vessel head

b. Temperatur*e

c. Pressur*e

d. Pressurizer level F. Inventory Loss T/P-9

1. Indication Date 1 /20/81 Pa<;e 4 of 16 Rev. . o

l I

~~I I..~~ I LESSON PLAN OUTL:::~iE L:C:SSON NO.

AIDS

2. ~!easurement

3. Flow path 4~ Treatment IV. ,1977 Natural Circulation Cooldmm T/P-10 A. Plant Status B. Differences

1. Pressurizer level behavior recognition

2. Inventory loss (lack of)

3. Vessel head voiding V. Radiological Effects I

A. Radio-chemistry B. Radioactive Effluents VI. Conclusions A. Natural circulation decay heat removal and subsequent cooldown of the reactor coolant systems. were adequate.

B. Anomalous pressurizer level behavior occurred after severar hours into the cooldown when a steam bubble was formed in the reactor vessel head. This resulted from cooling the I pressurizer below the temperature of the head area. .

i c.

I There is no temperature instrumentation in th~ reactor vessel head area. Such instrumentation could alert the I operator of approaching saturation conditions in the reactor vessel head area.

i I

I II D. Normal instrumentation for reading RCS hot leg temperature:

0 I I

i was not usable once the system was cooled below 515 F.  !

j I E. The subcQoling monitor instrument is also not usable belowl 515° F since it receives its hot leg temperature input I from a narrow range instrument.

~I Da ':.e 1/20/81 Page 5 of 16 0

'LESSON PLAN LESSON NO.__J I

OUTLINE -~IDS F. Although wide range pressure indication was available to the operators, no automatically recorded pressure trend-ing was available once system pressure was below 1500 psi. The operators manually recorded reactor coolant pressures at 30 minute intervals _during the cooldown.

G. After the shutdown cooling system was valved into the RCS approximately 10,000 gallons of reactor coolant was drained inadvertently to the refueling water tank at a rate of about 50 gpm.

H. Letdown system operations continued for a period of a-bout three hours after the pressurizer level anomalies started. ' .

I. At the time the steam bubble first started to form under the reactor vessel head, there existed in the reactor

. 0 vessel a temperature difference of about 200 F between the top of the vessel flange and the coolant nozzles.

This condition is not normally analyzed as part of the ASME Secti9-n I I I analysis.

J. A single malfunction in the common reactor coolant pump seal water cooling system led to the shutdown of all four reactor coolant pumps.

Date 1/20/81 Page 6 of 16 Rev. ___o__

.e S/G * '/G p

R * --0 z

1<

SIS (typical 1 of* 2)

'--D<'I-+ SDCS (out let)

SIS (1 of 4)

Sl>CS (inlet)

Senl Seal Seal Soni cooler cooler cooler cooler 1A1 1A2 f.jr supply CCW supply HCV14*7 HCV14*1 Roaclor '.'

Ouildino '*

AJr aupply ccw rolurn I HCVl 4--0 I HCVl 4*2 L---J

~BLE BUBBLE (ARTIST CO~CEPTIO~

0 C l) 0 FIGURE 3 ST. LUCIE E\'ENT PAGE 9 of 16

/CEDMnoule

/

• lnstrume~!atio:i

~ no:.z.le Outlet Upper guide Inlet nozzle structure no::zle

.I' I 1 Active Fuel I.I core assembly length I' I

Estimated Size of Initial Steam Bubb!e FIGURE .4 ST. LUCIE EVE~T PAGE 10 of 16

I. .I 600.-.---,r---.--r--.,----,--.--~-..----.---...--......---,

I I I Thot and T cold*

r Trip at 0233 natural circu!ation

~*-o St. Lucie. &11-50 580 t---:r----+~-+--+---+~4--!---+-~-l-~'~-l-~+'----J 560 ~\--+-+--+--+--~-+----1----l

~' \ ~ /~Z:f]' ou~o ottOO

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~

480 i---+----+--~+---+-+-~-+--+---j\~

0 5 10 15 20 25 30 35 40 45 50 I\

55 6()

Time After Trip (mim.ies) 2:33 .2:43 2:53 3:03 3:, 3 3:23 3:33 TimB of Day (a.m.)

Reactor Coolant Temoeratures First Hour A.'"!er Trip FIGURE 5 ST. LUCIE EVE~T PAGE 11 of 16

2300 rr--;-,--'-r--r--1.--i1--r--1,--,---,--1.---,.1--,_...;-,l--...--.,.-I--,-..........,.1--..--..-,-.---...,---l---.1--~

~ ~ "' .A I I I I I , I , I I I 2200

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1900 I 1 reference 62 1 '-+--+-'-t--_..._-i--_.1._-+-___.'-+---'--~_._1

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1 1 1700 1600r.11~'~-r--:--r-t-:-'--1r-~'-+~~'-+--"'""--+-+'--;'r-~'--1~~'*-+---'~+--~i~~l--!..1~

I I \I I .I I I I I I I I I I I

-._.._ __.__~T'-* Lower limit of pressure recorder-.,r-----+--...-+---+--1---+--i---+---I "2' 1500 i.-;....._...

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0 0 , 2 3 4 5 6 7 8 9 10 11 12 Time After Trip (hours)

I 2:33 3:33 4:33 5:33 6:33 7:33 8:33 9:33 10:33 11:33 I 12:33 1:3.3 2:33 Tir.ie al Day a .."71. I p.m.

Reactor P:'essure.

FIGURE 6 ST. LUCIE EVENT PAGE 12 of 16

Natural circulation cooldown DI St. Lucio 6*11*80 450 400 1--~--a---~-+------+-----+------1~----+---~-+-~---1------+-----i G:'

l.- ,,,,-- TtK>t (lrom core exit lhormocouplos)

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350

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Cll r 300 r-----;------t-------ir~~""=-CX::---~-1""~~-t-----~1--~~-t------i-~---i 200 2 3 4 5 6 7 8 9 10 11 Time Aller Trip (hours) 3:33 4:33 6:33 6:33 7:33 0:33 0:33 10:33 11:33 12:33 1:33 Tima ol D<iy a.m. p.m.

Reactor Coolant Tompcrolures Ourlno Cooldown

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FIGURE 8 ST. LUCIE EVENT PAGE 14 of 16

ill Loop

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-~~ o 1 I I I I I I I I I I I I I I I I I I I I I I I I I I I I 0 2 3 4 5 0 7 0 9 10 11 12 13 limo After Trip (hours)

Prossurlzor Levo I. 1077 St. Lucio Natural Clrculnllon Cooldown

ST. LUCIE STUDY ASSIG).:!ENT SHEET STUDENT:

DATE:

TIME to co~~LETE:

1. Determine and list* the requirements which authorize or require a trip of the reactor coolant pump(s),

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... . '.~'Ill

, *v '

,,, ~ .* ,I~~ -

2. Determine and list the requirements necessary to ::estart a reactor coolant pump(s).

• 3. D~velope a scenario, involving a single component or valve failure, which could create a similiar initiating event as occurred at. St. Lucie on June 11, 1980, at Saleci Nuclear Station.

4.

Based on review of the St. Lucie Event, outline the method you would recommend, as a licensed operator, for a natur3l circula-tion cooldown. Include in your outline the. c.ooldown rate you would utilize, how this rate would be maintained, your method of control to minimize the opportunity of ~oiding within the reactor c6olant system, and the parametric values when forced convectiqn cooling could be initiated (do not restart the re~

actor coolant pump(s) ) .