ML13317A424

To Operator Requalification Program Re General & Specific Operating Characteristics

ML13317A424
Person / Time
Site:	San Onofre
Issue date:	07/01/1982
From:	Kirby M, Kuhner P SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML13317A422	List: ML13317A420 ML13317A423 ML13317A424
References
4424A, OT-1062, OT-1062-R, OT-1062-R00, NUDOCS 8211240169
Download: ML13317A424 (9)

Text

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0 DA 07/01/82 SOUTHERN CALIFORNIA EDISON COMPANY SAN ONOFRE NUCLEAR GENERATING STATION NLCLEAi TRAINING DIV]SION COURSE UNIT 1 CPERATOR REQUALIFICATION TOIC GENERAL AND SPECIFIC QPERATING CHARACTERISTICS LENGTH OF LESSON 8 HOURS 01-1062 INSTRUCTIONAL ATERIAL Revised By:

M.

. Kirby Revieweo by:

P. R. Kuhrier Reviewed and Approved by:

M. J.

rby 8211240169 821119 PDR ADOCK 05000206 V

PDR

REFERE2CES

1. USNIC PIS Audit - June 3-4 Preliminary Results.

2. Baskin to USNFC - May 26, 1982.

3.

INPO, SOER, Pressurized Thermal Shock.

4.

Westinghouse Lesson Plan B-4, tlssion Product Poisoning Effects.

5. SCE Engineering Notes: Net Reactivity Eltect of Sm14 9 and pu239 Buildup on Load Change.

6.

SONGS 1 Operating Instruction S01-3-6, Plant Operation with Natural Circulation.

7.

Westinghouse, Mitigating Core Lamage.

B. Westinghouse Handout on Natural Circulation Operation.

38JECTIVES

1. To familiarize the operator with the history of PIS events.

2. To review results of USNFC PTS Auait.

3. To reinforce the effects of fission product poisoning on Normal Reactor Operations.

1. lo review the behavior of the plant during Natural Circulation conditions both normally and during cooldown.

INTROUCTION INSTRUCIOR/TRAINEE ACTIVITY A. Establish contact.

Introduce selt; write name on boara.

Write topic on boara; hand out attencance sheet, as applicable. Have student till out., Explain cottee mess.

B. Create interest.

Create interest.

D. Overview

1.

Objectives Ask questions as to student understanding ot objectives.

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INSTRUCTOWAIIN[E PRESENTATION ACTIVITY I.

STATUS OF PRESSURIZED THEI2fM SHLCK A. Review June 3 & 4 Audit Preliminary Results

1. Operators intervieweos 2 SRO"s, 2 ho's, and 1 STA.

2. Operators had a good understanding of PTS and the instrumentation used to mitigate a PIS event.

They had a good priority system, i.e., core cooling vs.

PIS.

3. The operators were not quizzed on pro cedures due to major procedure revisions that they were being trained on concur rently with NRR's audit.

4. The operators exhibited three nreas of weakness.

a.

They were weak on past PIS events both in the industry and at SONGS 1.

b.

They were not aware of problems in volved with cold leg injection vs.

mixing and cold water on contact with the vessel wall.

c.

They all felt that repressurization was necessary to challenge vessel integrity.

5.

Procedures:

a.

New procedures were very good and very thorough. NRR was very pleased that they had been based on plant specific analysis (best estimate).

Procedures providea adequate direction to prevent PIS.

b.

NRR felt that some direction should be provided to terminating charging to stagnate loops.

They also felt more human factors review of proce dures was needed.

6.

Training:

a.

The auaitors felt that the training outline and lesson plan were compre hensive and thorough.

The evalua tion of the operators through oral and written exams was adequate, the upcoming control room walkthroughs

INSTRUCTOR/TRAINEE PRESENTATION ACTIVITY with the new EOI'a was good and that the simulator training was adequate to insure understanding ot PTS events.

b.

Training needs to be upgraded In the areas of weakiess demonstrated by the operators in their evalua tions (see above).

7. Overall:

a. The overall assessment was good but it was emphasized that it was pre liminary and further review, parti culary of procedures, was ieeded.

B. History of PTS Events

1. 3 classes of PTS events.

a.

SBLOCA.

b.

Overfeeding Transient.

c.

Steam Line Break.

2.

Worst case examples in the industry for these events:

a.

Rancho Seco -

(excess feedwater transient).

On March 20, 1978, the Rancho Seco plant RCS was cooled from 582 degrees F to about 285 de grees F. in slightly more than one hour (approximately 300 F./hr.),

while FCS pressure was about 2000 psig. The transient was initiated by an inadvertent short in a DC power supply causing a loss of power to the plant's non-nuclear instru mentation (NNI).

Loss of NNI power caused the loss of most control room instrumentation and the generation of erroneous signals to the plant's Integrated Ontrol System (ICS).

The ICS reduced main feeawater, causing the reactor to trip on high pressure. The cooldown was initia tea when feedwater was readmitted to one steam generator by the ICS (auxiliary teedwater was automati cally initated) ana by the operators (ain teedwater was restored).

ihe cooldown caused system pressure to drop to the setpoint (1600 psig)

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INSTRUCTOF&ERANEE PERSENTATION ACTIVITY for the safety teatures actuation system which started the high pres sure injection pumps ano auxiliary teedwater to both steam generaLors.

High pressure injection flow re stored pressure to 2000 psig. With control room instrumentation either unavailable, or suspect for one hour and ten minutes (until NNJ power was restored), operators continued auxi liary teedwater and main feedwater to the steam generators whtle main taining ICS pressure with Ihe high pressure injection pumps.

b. Crystal River 3 -

(small break LOCA transient). On February 26,

1980, the Crystal River 3 plant experienced a small break LCCA transient when a power operated reliet valve (PDV) was inadvertently opened.

The re sulting transient caused a decrease in NCS temperature of about 90 de grees F. in 30 minutes (approximate ly 200 F./br.) with a system pres sure of about 2400 psig.

TIhe tran sient was initiated when an electri cal short in a DC power supply for the plant's NNI caused a pressurizer POFV to open, a loss of most control room instrumentation, and the genera tion of erroneous signals to the plant's ICS.

The ICS caused a reduc tion in feedwater flow and a with drawal of control rods. ICS pressure initially increased, tripping the reactor on high pressure, and then decreased as coolant dischargea through the open POW.

The high pressure injection pumps started at 1500 psig and repressurized the RCS to about 2400 psig.

The POWV block valve was closed, but flow out of the ICS continued through the pressurizer safety valves. After approximately 30 minutes, the high pressure injection pumps were throt tled back, but ICS pressure was maintained at about 2300 psig for the next one and a half hours. The FCS temperature decreased by about 90 degrees F. in the first 30 minutes and was thereafter brought to cold shutdown conditons by normal opera ting procedures since NNI power had been restored.

INSTRUCIOEWINEE PRESENTATION ACTIVITY

c.

Borssele -

(steam line break tran sient).

On March 2, 1981, an Inad vertent opening of a main uteam safety valve at the Borssel.e plant (located in the Netherlandu) caused one steam generator to boil dry and resulted in a primary system temp erature decrease troin 446 degrees F. to 284 degrees F. in abuout 20 minutes with primary system pressure above 2000 psig.

Opening o'f the steam safety valve was cauused by a maintenance error when the cable connectors of two solenoid pilot valves controlling the steam safety valve were interchanged.

When power to the solenoids was switched on during plant startup, the safety valve opened.

Operators discon nected the power supply to the pilot valves, but the safety valve re mained open because the piston of one of the pilot valves stuck in its open position.

For about 25 minutes, steam was released through a main steam safety valve into the atmosphere causing a blowdown of one steam generator until it boiled dry.

The primary system teperature de creased at a rate of about 550 F./

hr. for 20 minutes.

The primary system pressure dropped from 2234 psig to 2060 psig.

3. There have been 3 potential PTS at SONGS 1. They are summarized below:

a. April 30, 1972. On April 30, 1972, with the unit at 55 MWe during startup fram an outage, a tailure of the "C" feedwater controller re sulted in a reactor trip on high "C" steam generator level.

Overfil ling this generator caused average FCS temperature to fall from 5500F. to about 4600F. in 18 minutes and pressure fell trom 2035 to 1550 psig. The event was termi nated when the safety injection set point of 1685 psig was reached ano safety injection initiated. No actual flow from the safety injec tion system was added to the reactor coolant system. Nine minutes after actuation the safety injection sys tem was secured.

INSTRUCTO/TRAINLE PRESENTATION ACTIVITY

b. October 21, 1973. On October 21, 1973, unit load was being gradual ly reduced from 450 MWe to perform plant maintenance when a turbine trip ana resulting reactor trip occurred.

AT that time, the teedwatei regula ting system was programmed to open the regulating valves to 80% open on any trip. This resulted in a rapid filling of the steam generators and cooldown of the iCS trom 5480F. to 4700F. in about eight (8) minutes.

Initiation of Safety Injection at 1685 psig terminated the event. As a result of this event, the teed water regulating system was repro grammed to provide 5% flow on a reactor trip, thereby preventing a recurrence of this event.

c. September 3, 1981.

On September 3, 1981, with the unit operating at 390 MVe, a failure in the #1 Regulated Power Supply caused several alarms and the loss of several plant para meter indications.

As a result, the operator manually tripped the plant, but feedwater flow continued, re sulting in an overfilling of the steam generator. This resulted in ICS temperature falling from 5500F. to 480 0F. ana pressure falling from 2077 to 1700 psig in about tive (5) minutes. Safety Injection which terminated the event was automatically initiated at the.

new setpoint of 1735 psig.

As a re sult of this event, Westinghouse was consulted about the cooldown of the vessel.

They confirmea that vessel shock was not a concern in this event.

C.

PTS Concerns Over Mixing Phenomena

1.

Worst case can be when no mixing takes place between cold injection water and water in the loop. This places coldest water in contact with vessel wall and hence most severe thermal transient.

This can occur when loop stagnates which Westinghouse says has minimum possibil ity of occurance at San Onofre due to no MSIV's.

INSTRMCOWTRAINEE PRESE MATION ACTIVITY

2.

Also there are no RID's in downcomer region of vessel and SI injects down stream ot loop Tc Rib's so thai actual water temp. in downcumer is not known.

Therefore the operator needs to be cognizant of this actual temperature difference.

D.

"New" PTS Definition

1. Original PMS concerns required repres-Overhead of Westinghouse surization to present challenge to P S Curve.

vessel integrity. Further analysis has shown that temperature alone can be sufficient to challenge vessel integ rity. Westinghouse has come out with a graph that depicts areas of temp vs.

pressure that vessel integrity is chal lenged.

This is preliminary and will be finalized later this year.

E. Current PTS Status with NIC

1. NFC to come out with interim require ments for PTS. Now expected for Oct.

1982.

2. Interim criteria based on RTNDT value being argued between NYC and owners groups.

3. More to come!