ML20059D167

Intervenor Exhibit I-MFP-155,consisting of LER 1-91-002-01, Re Docket 50-275,dtd 910517

ML20059D167
Person / Time
Site:	Diablo Canyon
Issue date:	08/21/1993
From:	Shiffer J PACIFIC GAS & ELECTRIC CO.
To:
References
OLA-2-I-MFP-155, NUDOCS 9401070042
Download: ML20059D167 (9)

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May 17, 1991 eq PG&E Letter No. DCL-91-134 it ' O %

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Re: Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Licensee Event Report 1-91-002-01 Reactor Trip on Steam Generator Low Level with Steam flow /Feedwater Flow Mismatch due to Personnel Error Gentlemen:

Pursuant to 10 CFR 50.73(a)(2)(iv), PG&E is submitting the enclosed Licensee Event Report (LER) concerning a reactor trip on steam generator low level coincident with steam flow /feedwater flow mismatch due to personnel error. This revision is being submitted to present the results of investigations and corrective actions regarding four nonsafety-related equipment failures following the trip.

This event has in no way affected the health and safety of the public.

Sincerely,

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cc: Ann P. Hodgdon a .,, prinnto V John B. Martin '" '

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Phillip J. Morrill Paul P. Narbut

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LICENSEE EVENT REPORT (LER) 170G20

.ao u.. n) t m r,.o ... ,,, ,a,., ,, '. l 0l5l0l0l0l2l7l5 DIABLO CANYON UNIT 1 1l' 8 l viru REACTOR TRIP ON STEAM GENERATOR LOW LEVEL WITH STEAM FLOW /FEEDWATER FLOW MISMATCH DUE TO PERSONNEL ERROR sorri onri (s) un musen ts) atsomt ontt (7) ovata racitrists =vcLeto (si arme car en va sa er ou som one ya coctri su sta (s) gat 0 5 0 0 0 02 01 91 91 -

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MARTIN T. HUG, SENIOR REGULATORY COMPLIANCE ENGINEER ^" ' '

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EXPECTED SUSMISSION i I YES (if yes, complete EXPECTED SUBMI5510N CATE) lX l Wo ATE (W ogstaa;f (16)

On February 1, 1991, at 0909 PST, with Unit I operating at approximately 90 percent power, a Unit I reactor trip occurred due to Steam Generator (SG) 1-4 low level coincident with a steam flow /feedwater flow mismatch. A four-hour, non-emergency report was made in accordance with 10 CFR 50.72 (b)(2)(ii) on February 1, 1991, at 1149 PST.

The root cause of the event was determined to be personnel error (cognitive).

Personnel erecting scaffolding in the area of feedwater regulating valve FCV-530 inadvertently closed valve AIR-1-1-1041, which isolated instrument air supply to feedwater regulating and bypass valves FW-1-FCV-530, 540, 1530, and 1540. The regulating and bypass valves failed closed on loss of instrument air, isolating feedwater flow to two SGs. SG levels decreased and the unit tripped.

The corrective actions include: (1) issuance of a procedure to secure sensitive air valves against accidental closure; (2) issuance of a maintenance policy requiring that all work to be performed prior to the outage must be on the approved daily schedule; (3) issuance of a Work Planning policy on marking work packages; and (4) revision of Administrative Procedure (AP) C-59, " Elevated Work Structures," to clarify responsibilities and restrictions for erection and removal of scaffolding.

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1. Plant Conditions Unit I was in Mode 1 (Power Operation) at approximately 90 percent power.

II. Description of Event A. Event:

On December 5, 1990, Maintenance Plar.ning requested that scaffoldin, be provided to support the inspection and repair of Main Feedwater Regulating Valve (SJ)(FCV) FW-1-FCV-530, which was scheduled for inspection during the upcoming Unit I refueling outage.

In December 1990 and January 1991, the work package for erection of the scaffolding was prepared. The work package contained the work order, elevated work structure request form (scaffolding request),

component information, and an area map showing thE location of equipment. However, the scaffolding request was not fully completed.

The dates by which the scaffolding was required were not filled in and the request was not marked as " outage," " pre-outage," or "non-outage" work.

On January 28, 1991, Operations signed off on their review of the scaffolding request. In accordance with Administrative Procedure (AP)

C-59, " Elevated Work Structure:," Operations reviews scaffolding requests for their potential effect on equipment. Although AP C-59 requires a sketch on the scaffolding request form, it does not specify the level of detail required. The sketch on this scaffolding request form did not provide sufficient detail to show the exact location where the scaffolding was to be erected. Operations did not recognize that the area designated for scaffolding erection included FW-1-FCV-530.

Outage management had scheduled the scaffolding for erection on February 4, during the outage and after plant shutdown. Since erection was scheduled to be performed after plant shutdown, the work package did not appear on the daily (pre-outage) schedule. Work packages on the daily schedule receive additional plant management review and approval before being issued for work.

On January 30, 1991, prior to the scheduled time and during plant operation, erection of scaffolding began near valve AIR-1-1-1041, which is the instrument air isolation valve (LD)(ISV) for valves FW-1-FCV-530 and 540 and Bypass Feedwater Regulating Valves FW-1-FCV-1530 and 1540.

On February 1, 1991, three contractor personnel were setting planks on the scaffolding structure around FW-1-FCV-530 when they heard a hissing noise followed by a loud bang. At approximately 0908 PST, the 4

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I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION

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DIABLO CANYON UNIT 1 0l5l0l0l0l2l7l5 91 -

0l0l2 -t o io 3 l'l 8 control room received steam generator (AB)(SG) level deviation alarms for Steam Generators (SGs) 1-3 and 1-4. Operators responded by attempting to manually open valves FW-1-FCV-530 and 540 from the control room.

On February 1, 1991, at 0909 PST, a Unit I reactor trip occurred due.

to SG l-4 low level coincident with a steam flow /feedwater. flow mismatch. A four-hour, non-emergency report'was made in accordance with 10 CFR 50.72 (b)(2)(ii) - February 1, 1991, at 1149 PST.

l B. Inoperable Structures, Components, or Systems that Contributed to the  !

Event: '

None.

C. Dates and Approximate Times for Major Occurrences: l

1. January 30, 1991: Erection of scaffolding began.

l

2. February 1, 1991, at 0909 PST: Event / Discovery date.

Reactor trip due to SG l-4 low level with. steam l flow / feed flow mismatch' l while contractor personnel were setting planks on scaffolding. structure.  !

3. February 1, 1991, at 1149 PST: .The 4-hour non-emergency report to the NRC required by 10 CFR 50.72 (b)(2)(ii) was made.

D. Other Systems or Secondary Functions Affected:

1. Circulating Water Pump (KE)(P) 1-1 Circulating Water Pump (CWP) 1-1 failed to automatically restart following the 12kV bus power transfer to startup power. All CWP l-1 automatic transfer logic circuits were tested, but no defective components were identified.

2. 25kV Motor-0perated Disconnect (EL)(MOD)

The 25kV Motor-Operated Disconnect-(MOD) switch did not open fully. In preparation for outage work on the MOD, plastic sheeting had been securely taped to the MOD manual drive shaft and switch actuating shaft. When the MOD was. signaled to open, the turning drive shaft began to wrap the attached plastic 53665/0085K

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION 170G20

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Ygal (37) around the shaft. The shaft continued to turn until enough plastic had accumulated to contact the nearby switch actuating shaft. Continued wrapping forced the switch actuating shaft outward (horizontally), simulating "stop logic," which de-energized the MOD power and control circuit.

3. Control Rod Drive Mechanism Cooling Fan (CD)(FAN) E-13 Control Rod Drive Mechanism (CRDM) Cooling Fan E-13 failed to automatically restart following the power transfer. Electrical Maintenance believes that a combination of reduced component part clearances due to age-related shrinkage and dimensional changes resulting from elevated operating temperatures caused the magnetic starter to fail.

4. Main Turbine Stop Valve (SB)(ISV) FCV-145 Main Turbine Stop Valve FCV-145 did not fully close. Stroke time was determined to be very slow (in excess of two minutes).

A brass bushing on the actuator spring can assembly was found to be very dry and scored. After replacement of the bushing, stroke times were acceptable.

E. Method of Discovery:

The event was immediately apparent to plant operators due to alarms and indications received in the control room.

F. Operator Actions:

Operators noted that valves FW-1-FCV-530 and 540 were closed and attempted to manually open them from the control room. No response was observed. Operators were about to manually attempt to open bypass ,

valves FW-1-FCV-1530 and 1540 from the control room when the reactor i tripped. (However, since closure of AIR-I-1-1041 isolated instrument l air to the bypass valves as well as the regulating valves, the bypass valves would not have opened either.) l Following the trip, operators stabilized the plant in Mode 3 (Hot ,

Standby) using Emergency Operating Procedures (EPs) E-0, " Reactor Trip l or Safety Injection," and E-0.1, " Reactor Trip Response." Operators  !

restarted CWP 1-1, manually opened the MOD switch, restarted CRDM fan E-13, and locally verified the position of FCV-145. Immediately downstream of FCV-145, Turbine Governor Control Valve (SB)(FCV)

MS-1-FCV-141 had closed on the turbine trip, so no other imediate operator action was necessary regarding FCV-145.

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G. Safety System Responses: l

1. The reactor trip breakers (JC)(BKR) opened.

2. The main turbine (TA)(TRB) tripped.

3. The motor-driven and turbine-driven auxiliary feedwater pumps (BA)(P) started.

III. Cause of the Event A. Immediate Cause:

l The immediate cause of the unit trip was valves FW-1-FCV-530 and 540 failing closed upon loss of instrument air. '

3 B. Root Cause.

l The root cause of the unit trip was determined to be personnel error (cognitive). Following the trip, an immediate investigation determined closed position. that instrument air isolation. valve AIR-I-1-1041 was in a The investigation concluded that the valve had been inadvertently closed by the personnel who were erecting scaffolding.

C. Contributory Causes:

1. Prior to the commencement of outage activities, Maintenance management decided that outage-related work should not be performed before the outage unless it was on the daily schedule.

Management expectations regarding this decision were not met, I since the decision was implemented only within the outage scheduling group, and not emphasized to all other applicable plant personnel. Pre-outage work is screened by plant management, and work that may affect safety-related components ^

is not allowed to be put on the daily schedule. However because the personnel building the scaffolding were unawa,re of the management decision, they started the work early even though it was not on the daily schedule.

2. AP C-59 did not provide adequate guidance to preclude the i

erection of scaffolding around sensitive plant equipment during operation. Specifically, AP C-59 did not define sensitive equipment that would require additional review, did not adequately identify components to be worked, did not specify responsibilities for completing the scaffolding request form, and did not identify the allowed start date or plant modes for the work.

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3. The AIR-I-1-1041 valve handle requires a 90 degree rotation to move to the closed position and rotates easily. Therefore, though the handle is near a wall and not easily accessible, it can be closed accidentally.

IV. Analysis of the Event A. Safety Analysis:

Inadvertent loss of all main feedwater flow is a Condition II event, which has been analyzed in Final Safety Analysis Report Update Section-15.2.8, " Loss of Normal Feedwater." The reactor trip and automatic start of the auxiliary feedwater. pumps ensured that an adequate supply of water was provided to the SGs to-provide for the cooldown of the reactor per design. All safety-related equipment functioned.as designed.

Failure of a CWP to automatically restart.upon transfer to offsite power could render the condenser steam dump valves unavailable to remove excess heat following a turbine' trip. A previous Westinghouse analysis, in support of raising the setpoint of the reactor trip on turbine trip permissive from 10 percent to 50 percent thermal power, took credit for condenser steam dump valve operation. However, unavailability of the condenser steam dump system during a turbine trip has been analyzed in Final Safety Analysis Report Update Section 15.2.7, " Loss of External Electrical Load and/or Turbine Trip." The-reactor would trip on the turbine trip if above.50 percent power, or on reactor coolant system conditions if below 50 percent.. Pressurizer spray and pressurizer pu,ier-operated relief valves -are available, though the pressurizer safety valves and steam generator safety valves alone are adequate to protect the reactor coolant system and steam ge..erators against overpressure.

Based on the above analysis, the conditions of-this event were' bounded by previous safety analyses. Consequently, the inadvertent isolation of main feedwater and subsequent reactor trip did not adversely affect ,

the health and safety of the public.

V. Corrective Actions A. Imediate Corrective Actions:

1. The plant was stabilized in Mode 3 using EPs E-0 and E-0.1. <

2. CWP l-1 and CRDH fan E-13 were restarted and the MOD was manually opened.  !

1 1

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION 1700dU (

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B. Corrective Actions to Prevent Recurrence:

1. Reactor Trip 1

a. Operating Procedure (OP) K-1:lV, " Compressed Air System -

Preventing Inadvertent Operation of Instrument Air Valves," was issued to require that the handles of sensitive air valves be tie-wrapped to secure the valves against accidental closure.

b. A maintenance policy was issued requiring that all authorized pre-outage work must be shown as a line item on the daily schedule to ensure that proper reviews are performed.

c. Fork Planning Department Policy No. 11 was issued requiring work package folde.s to be labeled as follows: 1 (1) Work package folders for work to be performed on the daily schedule shall have no schedule designated on the cover; and (2) Work package falders for work to be performed during a planned refueling outage shall have the outage designated on the cover (e.g. IR4, 2R4).

d. AP C-59 will be revised to clarify responsibilities and restrictions for erection and removal of scaffolding.

Training on the revision will be conducted in accordance with AP E-4 Supplement 6, " Procedure Review and Revision Control."

2. CWP l-1 With no CWP available, the 40 percent condenser steam dump ,

valves are not available to remove heat following a turbine i trip. A previous Westinghouse analysis, in support of raising the setpoint of the reactor trip on turbine trip permissive from 10 percent to 50 percent thermal power, took credit for CWP and i steam dump valve operation. The analysis showed that there was no increase in the probability of challenges to the pressurizer power-operated relief valves (PORVs) during a turbine trip at 50 percent power if the steam dump valves are available. l Challenges to the PORVs increase the probability of a small break LOCA resulting from a stuck open PORV.

However, potential unavailability of the CWPs and condenser steam dump valves does not affect the ability of the PORVs and other safety-related equipment to perform their safety functions, if needed.

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i Unit 1 CWPs have failed to automatically restart following three a recent Unit I reactor trips. As a~ compensatory measure to l prevent challenges to the PORVs following a turbine trip, the i trip permissive setpoint has been set to nominal 15 percent of.

rated thermal power. This compensatory measure is documented in Justification for Continued Operation (JCO) 91-02, which will l 4

remain in effect until the reliability of a Unit I circulating

, water pump to restart after a transfer from auxiliary power to j start-up power is.shown. Action has been initiated to 4

investigate and resolve this problem.

3. 25kV MOD Signs were installed near the MOD drive shafts and chain

linkages to warn against placing equipment near the moving parts.

4. CRDM Fan E-13 Starters of the type in CRDM Fan E-13 have been replaced.

5. Main Turbine Stop Valve FCV-145

A preventive maintenance activity will be initiated to periodically lubricate the brass bushing in hydraulic actuators of the type used in FCV-145.

VI. Additional Information

A. Failed Components

. None. 1 1

J B. Previous LERs:

None.

C. Additional Information:

Historically, the most comon reactor trip initiators have been associated with SG level control, turbine trips, main feedwater pumps, maintenance, and procedure or design deficiencies. PG&E has an i ongoing Trip Reduction Program to evaluate further plant improvements i to reduce reactor trips.

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