ML20138A120

Trip Rept of 850617-20 Site Visits Re Reduction of Wrong Unit/Wrong Train Events at Facilities

ML20138A120
Person / Time
Site:	Surry, North Anna, 05000000
Issue date:	09/23/1985
From:	Persinko D Office of Nuclear Reactor Regulation
To:	Black K, Booher H, Lyons J, Regan W NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD), Office of Nuclear Reactor Regulation
Shared Package
ML18030B105	List: ML20138A120
References
NUDOCS 8510080463
Download: ML20138A120 (29)

Text

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- SEP 2 31985 .

l MEMORANDUM FOR: Harold R. Booher, Chief -

Licensee Qualifications Branch Division of Human Factors Safety, NRR 4

William H. Regan, Jr., Acting Chief Human Factors Engineering Branch Division of Human Factors Safety. NRR .

- Kathleen M. Black, Chief

' Nonreactor Assessment Staff 4

Office for Analysis and Evaluation .

of Operational Data

-

• James E. Lyons, Acting Chief -.

Technical and Operations Support Branch Planning and Program Analysis Staff, NRR . . .

THRU: Gregory C. Cwalina, Section Leader

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1. b[MaintenanceandSurveillanceSection Licensee Qualifications Branch C- Division of Human Factors Safety, NRR FROM:

Drew Persinko,-Maintenance and Surveillance Engineer Maintenance and Surveillance Section .

Licensee Qualifications Branch f

] - Division of Human Factors Safety, NRR -

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SUBJECT:

i ,

TRIP REPORT FOR SURRY AND NORTH ANNA SITE VISIT REGARDING '

WRONG UNIT / WRONG TRAIN.

).. . .

This memorandum documents the'activitiet and findings of an NRC staff visit  !

to the Surry and North Anna sites on June 17-20, 1985. Members of the NRC team for this visit included A. Ramey-Smith (DHFS) E. Trager (AE00) and  :

' D.Persinko(DHFS). The site visit was conducted as part of the short-term l effort to determine whether simple, low cost improvements can be identified and implemented to reduce the frequency of wrong unit / wrong train events occurring at nuclear power reactor facilities. Upon completion of all site

visits, a compilation of factors contributing to the events will be performed and a report issued which discusses causes and recommendations. Long term i resolution will be accomplished as part of the Maintenance and Surveillance Program Plan being conducted by the Maintenance and Surveillaisce

.Section of LQB/DHFS.

1 I l General Infonnation

l The Surry site is located 17 miles northwest of Newport News, Virginia, on the James River. There are two reactors. Surry I and Surry 2, located at the - i site. Surry I has a maximum dependable capacity (net) of 781 MWe and was

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placed into comercial operation on December 22, 1972. Surry 2 has a maximum ~

dependable capacity (net) of 775 MWe and was placed into commercial operation on May 1, 1973. Both units are Westinghouse PWRs and the architect / engineer for both units was Stone and Webster.

• The North Anna site is located 40 miles northwest of Richmond, Virginia, on which are two reactors, North Anna 1 and North Anna 2. Both units are

Westinghouse PWRs and the architect / engineer for both units was Stone and

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Webster. North Anna 1 has a maximum dependable capacity (net) of 890 MWe and was placed into comercial operation on June 6,1978. North Anna 2 has a maximum dependable capacity (net) of 893 MWe and went into comercial operation on, December 14, 1980.

The licensee for both sites is Virginia Electric Power Company (VEPCo).

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Site Visit Agenda i

The discussions and in-plant observations were centered around four wrong unit / wrong train events that occurred at Surry between 1981 and 1983, and two that occurred at North Anna between 1982 and 1985. The LER numbers for these events at Surry are 81-001,82-072, 83-033, and 83-051 and at North Anna are 82-022 and 85-006. During both site visits, the NRC team inspected the locations of the reported wrong unit / wrong train events to the extent possible, and discussed the events with the Human Perfomance Evaluation G< System (HPES) coordinators at each site as well as many of the individuals directly involved with the event. At Surry, the licensee's corporate HPES coordinator also participated in the discussions while at North Anna, a

_ nuclear specialist participated. At both sites, we brtefly spoke wit.h the assistant station manager. Enclosures 1 and 2 provide a sequence of -

i events for each of the LERs a sumary of the licensee's conclusions regarding the event, NRC staff observations and a sumary of the discussions

. concerning each event.

GeneralObservations s., .

1 A. Human Perfonnance Evaluation System (HPES) l Both Surry and North Anna are participating in the HPES initiated by INPO. A large part of the discussions at Surry and North Anna focused

on the HPES implementation, which has two main thrusts

1) to

investigate incorrect human actions, and 2) to prevent further i incorrect human actions from occurring. Upon completion of an investigation, recomended corrective actions are presented, to management for review and approval. To the extent possible, the system is non-punitive and relies on confidentiality. The HPES it a l

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forward-looking program and only a few retroactive investigations have been conducted; no HPES investigations have been conducted on the LERs reviewed during this site visit. All completed reports are provided to

' INP0. The system has been in place at VEPCo for approximately one year but has been functional only since October 1984. Prior to the HPES, no formal investigation of inappropriate human actions was conducted outside of normal LER investigations.

Although Surry and North Anna are both owned and operated by VEPCo each site implements the HPES differently. At Surry, one person is assigned as HPES coordinator and devotes up to 75% of his time to the HPES. This person conducts the investigations, writes the reports, and promotes the HPES to plant personnel. He is highly visible as the focal point for '

the HPES. At North Anna, one person is also assigned as HPES coordinator; however, five STAS in addition to the coordinator conduct

- investigations and write reports. ,As a result of the differing .

approaches to implementation, it appears that the HPES at North Anna may be more ingrained into the usual course of business; however, near misses may not be as readily reported due to a perceived decrease in . .

confidentiality because of the additional people involved in an investigation. Feedback sessions are held to infonn plant personnel about information learned from the HPES investigations. The. feedback is supplied mainly to operators and instrument technicians, and in some

,O., instances, to mechanical, electrical and HP technicians.

i l As a result of the HPES, both sites have installed distinctive, identically designed signs which are color coded by unit. As shown in 1 Figure 1, the signs are approximately 1 foot x 1 foot and warn plant -

personnel that equipment in a given area is associated with a specific i

. unit.

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B. Observations at Surry -

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Surry utilizes a color-coding scheme of green for Unit 1 and yellow for t

Unit 2. This color-coding scheme is evident on the unit ilesignation

! signs described above and on other plexiglas signs located throughout both units which state the building, the unit number, the elevation, and the type of safety protection required (e.g., earplugs, safety glasses, etc.). The signs are approximately 1xli feet and are shown in Figure 2.

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. Station operating procedures are also color coded green for Unit 1 yellow for Unit 2 and blue if the procedure applies to both units.

. . W/WT events have occurr'ed at'Surry which were non-reportable and were investigated by the HPES. The HPES coordinator at Surry agreed to provide the staff an edited description of these events.

C. Observations at North Anna Although the unit identification signs described in Section A are in place,are Surry no other utilized. unit designation signs such as the plexiglas signs at

- The unit designation signs that are utilized are color cdUen blue for Unit 1 and yellow for Unit 2.

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The operating procedures are color coded blue for Unit 1, yellow for -

Unit 2 and pink if the procedure applies to both units.

In addition to the events discussed in the enclosure, approximately 5 additional wrong unit / wrong train or wrong component events have occurred at North Anna which were non-reportable. At the site visit, the HPES coordinator agreed to provide an edited sumary.of these events for staff use; however, after further review subsequent to the site l

' visit, the HPES coordinator believed that these reports would not be very useful to the staff after editing out plant-specific infonnation.

D. Observations Comon to Surry and North Anna

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• Brass component identification tags on valves are currently being

'- replaced with aluminum embossed tags for better visibility. The tags contain a description of the component and the component identification as shown in Figure 3. .

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Emergency buses, switchgear, and cable trays are identically color coded in both Units 1 and 2. Train H is orange Train J is purple and any comon components (e.g. the line feeding a safety injection pump before splittingupintoHandJTrains)aregreen(TrainsHandJdesignations are equivalent to redundant Trains A and B). To distinguish between units on large electrical panels containing individual motor control switches, one must look for either toe larger unit. designation signs described above or the motor control switch tags which will contain unit designation (e.g.,101A for Unit 1, 201A for Unit 2). Addit 3cnally, larger component labels which designate unit number are located on electrical panels as shown in Figure 4. ~~

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A five shift rotation scheme is used with shifts rotating between Unit 1 control board, Unit 2 control board, liquid waste operations, auxiliary

_O. building, service building (switch-gear, die:e1 generators), Unit I turbine building, Unit 2 turbine building, and outside (switchyard, intake structure). There are usually 10-11 people per shift who rotate every 6-7 days. Assignments within a shift crew rptate at the discretion of the supervisor. -

~~. - Within the last few years, training programs have been revised to .

upgrade them and new ones implemented. Prior to these revisions, all generalemployeetrainingwasobtainedonthajob(0JT). Currently. -

l af,ter receiving 4 days of required radiological training for all .

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employees, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of general training is offered on a voluntary basis at the discretion of an employee's super v~ isor to any interested employees (e.g., mechanics, corporate, etc.) to provide the " big picture" regarding plant operations. ,

j Beyond the training already mentioned, operators and engineers are i required to take an 80 hour9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> general employee training course and a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> course covering watch-standing principles. To become a Licensed 1 Reactor Operator, 58 weeks of training both at the plant and classroom j is required.

i l VEPCo utilizes a step program for operators, electricians, i

mechanics, engineers, and other plant personnel. The step trogram is similar to an apprentice / journeyman program used by unionized labor ,

whereby one increases in steps with increased experience and training. -

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k Since February 1985, the licensee has instituted corporate ide training -

on watch-standing techniques on independent verification. The training emphasizes the need, methodology, and circumstances regarding independent verification. Independent verification is rNufred on tagging of safety systems and on non-safety systems at the discretion of the shift supervi'sor. It is used mainly on electrical components and on valve line-ups.

Exit Meetings At the exit meetings at both sites, the NRC team expressed its appreciation to the Surry and North Anna staffs, in particular to the site HPES coordinators.at each site and the corporate HPES coordinator who assisted

.. with the discussions at Surry and tha nuclear specialist who assisted with '

m the discussions at North Anna. Their cooperation in planning the visit, coordinating the tour and discussions, and providing information made the site visits informative and productive. .

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Drew Persinko, Maintenance and Surveillance Engineer O

Licensee Qualifications Branch Division of Human Factors Safety, NRR

Enclosure:

As Stated ,

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Enclosure 1 WU/WTEVENTSATSURRY(DOCKETNO.50-280 UNIT 1) '

(DOCKET NO. 50-281 UNIT 2)

I. LER 281-81-001 -

Boric Acid Improper Valve Line-Up (Wrong Unit)

A. The following event information was provided by the licensee in the LER:

"With #

Unit No. 2 at 100% power valve 2-CH-226 was . inadvertently clo, sed. This made one of the two boric acid flow paths to the .

i core inoperable. The inoperable flow path was discovered soon thereafter when operators attempted to use the boric acid blender to replenish the volume control tank."' '

"The cause of the event was personnel error. Unit No. I's valve.

1-CH-266. was to be tagged closed. However, an operator T' inadvertently closed Unit No. 2's valve. 2-CH-226."

s "The seriousness of the event was stressed to all personnel l __

involved and the individual was appropriately? disciplined."

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8. NRC Discussions with Plant Staff.

I During the site visit, the NRC team spoke with the HPES coordinator

[* and the individual who discovered the event. From these i

discussions it was learned that the valves are located in the same

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. space below the boric acid tank and are not far' apart. There was general agreement that the individual who perfoiined the event was relatively inexperienced as he had been in operations only a few l days before tagging the valve. Additionally, the labe] was crusted over with boric acid residue, the individual .was usifig,,a respirator due to high radiation -end no independent verificatio1,was required.

Although not positively stated, the NRC team was told that possibly ,

( the supervisor gave incorrect instructions and that the lighting may -

have been poor.

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.The HPES coordinator described steps which have been taken -

subsequent to this event which would cause such an event to be less likely to' recur. These steps are:

1. The area where the event occurred has since been cleaned up.

2. Lighting in the area is adequate.

3. Since the brass labels in use then became difficult to read '

1

., over a period of time all valve labels are being replaced with aluminum embossed labels.

4. Plant procedures are now stamped with a section which is to 1

be checked if the label needs to be replaced. - -

j 5. At the time of the event, only DJT was in ,effect. Now, improved j -L/ employee training is in effect, as described earlier.

i C. NRC Observations a

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The NRC team was unable to view the area where the incident

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, occurred because of radiological considerations.

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1. The. aluminum embo~ssed tags being installed on all valves are difficult to read, except in good lighting, due to a low contrast between the lettering and the background.

j 2. Although all employees are required to receive 4 days of i

radiological training, the 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> course.regarding plant operations is voluntary at the discretion of an_eniployee's l1 supervisor. ,

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II. LER 280-82-072 "A" S. J. Accumulator Level Below T. S.' bait with ,

"B"S.I.AccumulatorValveClosed.(WrongTrain)

A..Thefollowingeventinformationwasprovidedbytheiicenseeinthe LER:

1

! "With Unit I at 100% power, during the performance of P. T. 18.5 l (flushing of sensitizee stainless steel piping), "A" Safety l Injection Accumulator was inadvertently drained to a level below the i

, Tech. Spec. minimum. Also, at this time, "B" 5. I. Accumulator *

'. discharge valve (MOV-18058) was under adninistrative control, in the

, closed position, to facilitate performance of P. T. 18.5." .

. "The cause of the event was due to an operator opening the wrong .

test valve during the perfonnance of the P. T. This action may have i been enhanced by the arrangement of the accumulator test valve

,_ switches on the control board."

4 "The arrangement of the accumulator test valve' switches will be

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incorporated into the NUREG-0700 review proce'ss." (i.e., detailed-j- . control room design review) . i t

. "The operator involved was disciplined and reinstructed on the i l... importance of follo' wing ' procedures.."

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8. NRC Discussions with Plant Staff. .

, The NRC team spoke with the HPES coordinator and the individual who 4

turned the incorrect switch in this event. The individual had just

become a licensed reactor operator a few weeks' prior to this event, l was working on the swing shift, and indicated that thiis particular t, ~*

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P. T. was never performed on the simulator. The individual had the -

procedure but still turned the wrong switch. Both units contain the switch configuration shown above. ,

p C. NRC Observations 4

1. The NRC team viewed the switches in the control room where the error occurred. The switches are rotating knobs such that the

, associated valve is closed if the switch is pointing to the left _

., and open if it is pointing to the right. The switches are associated with three accumulator trains A 8 and C. The ,

configuration of the switches and the associated train is as i -

follows:

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A B C -

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A B C ACCUM DRN ISOL VLYS

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ACCUM FILL ISOL VLVS (P .

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A B C ACCUM DISCH ISO VLVS

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2. The arrangement of these valve switches is a clea violation of -

human factors engineering principles. The NRC team cannot presently state hew this human engineering discrepancy is being '

addressed as part of the detailed control room design (D' CRDR) review because the licensee's DCRDR sumary report is not l

expected to be received by the staff until early 1986.

j III. LER 280-83-033 - Torque Switch Removed from MOV-12898 (Wrong Unit) t "i A. The following event infomation was provided by the Licensee in the LER: 5

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"With Unit 1 at 100% power, an electrician perfoming maintenance removed a torque switch from MOV-12898 (normal charging isolation -

valve). With the torque switch removed, the MOV would not close on a Safety Injection Signal."

"At the. time of the event, Unit 2 was at cold shutdown and a i

_ maintenance request had been initiated to adjus't limits on

' MOV-22898(Unit 2chargingisolationvalve). However, the

• electricianassignedtothetaskwenttotheoperatingunit(Unit 1) i

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to perform the maintenance. As a result, the torque switch was

} removedfromMOV-1(898.", -

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" Subsequent Corrective Action: None" 4

"The electrician involved received disciplinary action."

{ B. NRC Discussions with Plant Staff. .

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During the site visit._the NRC team spoke with the HPES coordinator andtheelectricianresponsibleforremovingthetohueswitch.The

( electrician had started working at this site during an outage and

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had been there for only 4-5 months. He had worked previously at fossilplantsandsaidthathedidn'trealizethattwounitswogd share the same basement because at fossil plants, separate. rooms are used for each unit. He said that he was not familiar with the plant '

and did not realize that the first number on the component identification tag designated the unit. He had to clean off the tag l

be'ere he could read it and only checked the last four numbers be,cause he thought he was in the correct area. The electrician

., siated that the labels are clearer now than they were before. .

'] Subsequent to this event, the licensee instituted the general training program which covers unit designation, tags, procedures and dual systems which the licensee believes will reduce the likelihood

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of this type of event recurring.

j C. NRC Observations The NRC-team was unable to view the area where,the event

__ occurred due to radiological considerations.?

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1. Although plant personnel now receive general training, j -

contractors do not.

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$~ 2. The auxiliary basement does not contain unit identification signs, either at the time of the event or now.

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< IV. LER 280-83-051 CV-P-18 Suction Line Blank Flanged (Wrong Train)

A. The following event infomation was provided by the licensee in the LER:

l "With Unit 1 at 100% and with 1-CV-P-1A ("A" Containment Vacuum Pump) tagged out for maintenance, the suction line for 1-CV-P-1B

("B" Cor,tainment Vacuum Pump) was inadvertently blank flanged. As a

. result, both containment vacuum pumps were inoperable. -

"A blank flange was to be placed on the suction line for 1-CV-P-1A

. to support maintenance activities on that pump. However, an -

operator incorrectly identified the proper suction line for ~

l 1-CV-P-1A and the flange was installed on the suction line for -

1-CV-P-1B."

4

- O. " Subsequent Corrective Action: None" f

"The operator involved was disciplined. The re"quirement for

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procedures and independent verification will be re-emphasized for the control of blank flanges." -

B'. NRCDiscussionwith, Plan,tStaff. '

l .- . . . .

The NRC team spoke with the operator who incorrectly identified the suction line to be blank flanged and the HPES coordinator regarding this event. -

The reactor operator told the NRC team that he was supposed to show amechanicthelinetobeblankflangedinsideconttisent. He had received verbal instruqtions from his supervisor in thIr ,change room that the blank flange was to be installed after check valve 4 or 5;

! however, the supervisor never got back with the operator to confirm

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the valve number. The operator was told to get the job going and -

thus was in a hurry. The reactor operator found the check valve with a 4 or 5 on it and identified this line to the shchanic as the line to be blank flanged. Because the operator thought that only one line ended in this area, he did not look any further and l believed he had identified the correct line. The reactor operator did not have a print with him, was using prctective clothing and a respirator, and said that the area was confined. The reactor-operator knew the pump number on which maintenance was to be ,

performed; however, the pumps are located on the opposite side of

' - - the containment from where the blank flange was to be installed and

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. thus, he was not able to trace the line in the field. Since the  ;

operator knew the pump number, he would have been able to trac,e,the l line on the drawing and identify the correct check valve (4 or 5) -

i associated with the pump requiring maintenance if he had had the drawing with him. The operator said that the system had been

- O_. changed before he arrived at the phnt but that the drawings had not been updated to reflect this change. The old drawing had been used during training. The NRC team was told hat. reactor operators are not required to carry a print with them but that doing so is good practice.

The licensee believges that the following changes which have been - I

.* implemented subsequent to this event would make the recurrence of this type of event less likely: 1

1. Trainingnowreviewsdesignchangepackages-(DCP)fortraining l needs. Operators are now trained on design changes which have occurred. .

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2. Since this is a safety-related system, independent verification is now required.

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3. Engineering and construction currently require that safety .

related changes be reflected on drawings within 15 days and that other changes be reflected on the drawings within 60fdays.

4. More formalized mechanisms now exist to change drawings.

C. NRC Observations The NRC team was unable to view the area where the event occurred because it was inside containment.

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3, Enclo'sure 2 WU/WT EVENTS AT NORTH ANNA (DOCKET NO. 50-338 UNIT 1- i (DOCKETNO.50-339 UNIT 2 .

I. LER 339-82-022 Inoperability of Quench Spray Subsystem and the Recirculation Spray System (Wrong Train) - r i

A. The following event information was provided by the licensee in the i LER

] "On May 28, 1982, with Unit 2 in Mode 4 at approximately 345"F, both

, , trains of the Quench Spray Subsystem and the Recirculation Spray .

l  : System were inoperable for 17 minutes. The Chemical Addition System q was isolated for 43 minutes."

j . -

i' l . "This event occurred during the performance of the Emergency Bus Blackout and SI Functional Tests, 2-PT-83.5, when steps were being

] completed in preparation for the 23 Bus Blackout and SI. In j accordance with the procedure Train 8 pumps 2-RS-P-18, 2-RS-P-28,

, and 2-QS-P-18 were placed in " Pull-to-Lock". Pump 2-RS-P-38 was running on recirculation. The Train A pumps, which had been previously tested, were in the " Auto" positidn with the exception of ,

i 2-RS-P-3A which was in " Normal." An Engineer and an Electrician l

• were instructed to go to Solid State Protection Train B Output Cabinet and install jumpers which insure that CDA loads which are shed from the emergency bus are reloaded onto the bus following

• l- restoration of bus voltage. At 2243, they installed the jumpers in Train A of Solid State Protection, instead of Train 8, which began 1

starting Train A Quench Spray and Recirculation' Spray pumps which

were in the " Auto" or " Normal" modes of operation. Operations I j personnel began placing Train A pumps in " Pull-to-Lock" as their j energization was annunciated in the Control Room. Casing Cooling Pump 2-RS-P-3A had to be held in the "Stop" position tince the control switch does not have a " Pull-to-Lhck" functios. The breaker

, for the Chemical Addition Tank Train A discharge valve MOV-QS-202A ,

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was deenergized and the associated manual isolation valves,were l closed. At-this time. Train 8 Quench Spray and Recirculation Spray pumps were in the " Pull-to-Lock" or held in "Off" to teminate an <

i; inadvertent actuation of the Containment Spray Systems."

" Scheduled Corrective Action

The test procedure is being reviewed 1 ..

., for modifications which will minimize the possibility of personnel i -: error while it is being performed. Color coding of the Solid State

' -" l Protection Cabinets is being considered as a prevention against 1 .

personnel entering the wrong train." *

"The incident was discussed with the engineers responsible for '

i testing in order to minimize this type of error and to emphasize the proper cautions to be taken."

i 8. NRC Discussions with plant Staff o '

During discussions with the HPES coordinator and the individual who' 1 identified the incorrect cabinet, this individual said he had worked

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.. 13 hour1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> backshifts for 4 consecutive days and, for approximately one

• l week prior ,to the event, he had worked on the A (or H) Train -

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preparing to run the A bus procedure. When he arrived at work on the day of the test he was informed that the previous shift had completedtestingTrainAandthathewouldbetestingTrain8(or J). He believed that he incorrectly identified the Train A cabinet instead of the Train 8 cabinet because of a mindset developed as a r.sult of working on Train A the previous week. The individual had the correct procedure with him at the time of the nest; however.

theprocedurehaddiagramsforbothTrainsonthesadpage. The j

procedure required approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of set-up time for the

( first test and 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the second. I 1

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1 Actions taken by the licensee subsequent to this event which the licensee believes reduce the likelihood of this event recurring are:

1. The one long procedure has been broken down into 3 separate

, procedures, each requiring about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to set up and run.

This avoids any shift overlap.

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2. Each train has a separate procedure with the connon part l remaining as one test.

3. Both trains are no longer diagranned on the same page. ,

l C. NRC Observations

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1.

The NRC team viewed the Solid State Protection Cabinets where

the jumpers were incorrectly installed and where they should have been installed. The room in which the cabinets were located consisted of a main aisle down the center with rows of

• l cabinets on either side of the aisle. The rows of cabinets were i

• perpendicular to the main aisle and created aisles between the

! cabinets. To enter.a cabinet, one must walk down the main aisle and ' turn right or left in'to the correct aisle between cabinets.

l and then select the correct cabinet in that row. In this case,

! the electrician and engineer entered the second aisle to the

{ right of the main aisle where the Train A cabinets were located f when they should have entered the third aisle where the Train 8 i

cabinets were located. .

2.

4 Although these par.ticular procedures have been , separated by train, this was not done for all procedures which contain

( references to both trains.

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) - - -- - . - - - - .- .- - - - - .. - -. .. - . - . . - . - _ ,

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3. There is no identification at the head of an aissle of cabinets to aid in locating the correct cabinet. All aisles coming off of the main aisle look extremely similar from the main aisle in the room.

4. A row of cabinets is approximately 6-7 feet high by 2 feet wide by 15 feet long. There are approximately 5 cabinets in a row.

The cabinets are gray in color and are identified by a gray

) label (approximately 6" x 6") with black lettering stating the function of cabinet contents and the associated train and a.

separate component numbering label. This latter label is approximately 1" x 3" and contains white letters and number's on -

a black background. Both labels are.shown in Figure 5.

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i, II.. LER 339-85-006 De-Energization (J 120V AC Vital Bus 2-1 A. The following event information was provided by the licensee in the LER

j "At 0915 on April 26, 1985. Unit 2 tripped from 100% power when the p 120V AC Vital Bus 2-1 (EIIS System Identifier EF, EIIS Component . ,

l Identifier BU) was inadvertently de-energized. The 120V AC Vital

' l Sus 2-I was de-energized when an unlicensed Control Room-Operator S (CRO)openedthepowersupplybreakertotheinverterwhichfeeds ~

{ , the 120V AC 2-I Vital Sus. The 120V AC Vital Bus 2-I supplies power s to the relay which senses the breaker position of Reactor Coolant '

j Pump 'A'. When the 120V AC 2-I Vital Bus was de-energized, this.

! relay was de-energized which caused the reactor protection system to sense,that the 'A' Reactor Coolant Pump breaker was open. A reactor j trip signal was generated as a result of the reactor protection l __ system sensing the 'A' Reactor Coolant Pump breaker open coincident

{ '

with reactor power greater than 30%. Reactor Coolant Pump 'A' did'

not actually trip during this event. The 120V AC Vital Bus 2-I was l .

re-energized seconds after the trip when the unlicensed CRO, who i

opened the , supply b.reaker to ,the inverter, realized his mistake and

{' ' ' ,- closed the inverter supply breaker.-

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All equipment powered from the 120V AC Vital Bus 2-I responded as s expected when the bus was de-energized. All eight circulating water i

waterbox vacuum breakers opened when the vital bus was de-energized, .

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! which caused all circulating water pumps to the main condenser to '

trip. As a result, the condenser was not available-to remov.e L-secondary side heat via the condenser steam dumps. ,,Ipstead, steam ,

j was released through the steam generator PORV's to the atmosphere.

il Inaddition,tworupturediscs(EIISComponentIdentifierRPD)were

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{ blown and four rupture discs were damaged on low pressure turbines l because circulating water was not available to cool the condenser.

These rupture discs were replaced.

4 l All Auxiliary Feedwater pumps started as a result of low Steam l

Generator level. TV-BD-2000, 'B' Steam Generator blowdown inside j

containment isolation valve, indicated mid-position following a low J -

Steam Generator level isolation signal. The redundant blowdown trip 1 -

valve. TV-BD-200C, closed during this event. The position 4 -

indication problem with TV-BD-200C was corrected on April 26, 1985.

i l

The 2-I 120V AC Vital Bus was de-energized when an unlicensed CR0 -

j opened a breaker in the wrong power supply cabinet. The Unit 2 l Assistant Shift Supervisor (a SRO) had instructed the unlicensed CR0 j- to open a breaker in a specific power supply cabinet as part of j preparation for a maintenance activity. The SR0 had obtained the i

breaker number and power supply cabinet identification from name tags on the Main Control Board. The power supply cabinet

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j information on the Main Control Board was misleading which caused -

.. the SR0 to associate the power supply cabinet information on the Main Control Board with,another power supply cabinet. Corrective -

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actions to prevent recurrence are being evaluated.

The Unit was returned to criticality on April 27, 1985 and reached 3 100% power on April 30, 1985." -

B. NRC Discussions with Plant Staff .

The licensee believed.that differences in component identification

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schemes contributed to de-energiration of the incorrect breaker.

The labelling used in the control room on the control boards  :

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sometimes uses cosenon names w',thout unique scrk numbers (individual  ;

componentnumbers). Individual ' component or mark numbers can take i various forms, all of which are equivalent. For example, each of l the following mark numbers can designate the sa:ne piece of equipment:  :

TV

CC-104A

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i. . TV-CC-104A -

l 1-CC-TV-104A Component number

.[ Component -

System Designation (e.g.

A componentcooling)

Unit Number In this particular instance, the SRO gave an,u'nlicensed CR0 a slip

_. of paper containing the valves to be closed, the panel number in "

l 1

which the circuit breaker is located and the circuit breaker number. l

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. The information was taken from the main control board valve switch 1abel. The SR0 tu[ned to another CR0 to confirm the location of the

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panel and t'old the unlicensed CR0 that it was located in a level below the control room. The unlicensed CR0 went to that location and opened the circuit corresponding to the circuit number on the piece of paper given to him by the SRO. Both panels were known as 2-I; however, the panel where the unlicensed CR0 should have gone was Main Control Boasd DC 50V Panel 2-1 whereas he went downstairs to DC Distribution Panel 2-1 as directed by ,the SRO., _'The licensee said that there are se,yeral panels commonly referred to as 2-I.

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l Procedures reference unique mark numbers but were n'ot used in this I -

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. 1 g., we-,-.ew.-ay, --.-w..n,-w.,w mve, - ew sw P T'*T v vev-P'-vf'-*i7'e 9 W y -=r-

7 case. The following depicts the labels involved in this error.

Name tag of walve switch to be closed as shown on main control i board:

LOOP B FILL HDR ISO VV HCV - 2556B

' FC DC PNL 2-1 CKT 13 Information on the slip of paper given to the unlicensed CR0 by the SRO: -

HCV-2556 A,B,C - -

DC PNL 2-1 i C' CKT 13

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Name tag on the power supply panet that the unlicensed CR0 should have gone to behind the vertical boards in the' control room:

MAIN CONTROL BOARD o

. . DC SOY PANEL 2-1 2-EP-CB-26A -

Name tag on DC Distribution Panel 2-1 located below the control room that the unlicensed CR0 mistakenly went to as directed by the SRO:

2-EP-CB-12A .

Name tag next to breahr that incorrectly was opened in DC Distribution Panel 2-I located below the control room:

(  :

2-VB-I-01 '

2-VBBANK001 f

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- Although the SR0 had not given the explicit mark number of,the cabinet and had given the incorrect location of panel, the unlicensed CR0 could have noticed the error by checking the valve  ;

numbers on the slip of paper given to him by the SRO against the tag <

. next to the breaker in the panel. These tags identify the valve

. numbers associated with the particular breaker adjacent to the -tag.

ThItagshouldhaveread2-50V-2556A,B,Cinsteadof2-V8-I-01.,

The panels usually also have a list of the circuit breakers and j their associated valves, known as the load list, located on the l

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inside of the panel door. The unlicensed CR0 believed that the' load list may have been useful in stopping the error because it would have provided him with a list showing what valves are associated with each breaker; however, the load list was not in place at the j 7 time the error occurred. Because of different amperage ratings, the breakers in the correct panel are smaller than these in the i incorrect panel which could have provided the unlicensed CR0 with a

_ clue that something was in error. The unlicensed CRO, however, said that he saw 2-EP-C8-12A on the panel and knew it was DC. Because o'f

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this and because it was the location told to him by the SR0, he ,

thought he was in the correct panel.

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He did not check the number next to the breaker but rather just went to breaker 13 as written on the slip of paper and de-energized 'the circuit. He said that perhaps the missing load list may have thrown j him off.

i Subsequent to this event, the licensee has taken the following steps

to reduce the likelihood of this event recurring

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1. The event has been discussed with other operators pointing out ,

( that one should stop and check if there are any discrepancies -

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..ww...,.w,-.,%,,,. . , . ,e,m, ,.w_,.,,,w,,,3.,,_-9_,,,,., , . , , .-

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t l between unique mark numbers. During operator quaTification and requalification, it will now be stressed that operations personnel use unique mark numbers.

2. Control room labels are being revised to supplement common names with mark numbers.

C. NRC Observations I .

. 1. The power supply cabinet that was incorrectly entered is a dark gray cabinet approximately 6 feet high by 2 feet wide located at ,

a level which is below the control room. There were two similar

cabinets (2-Iand2-II)inthesamegeneralvicinity. The 2-II cabinet in that vicinity in which no error had occurred had a

- large 2-II written in magic marker on the outside as well as a 1" x 3" black label with white lettering designating the component number glued to the front. The ' cabinet in which the

,,, error occurred had a large 2-I written in magic marker on the "

outside. The location for the black component label was clearly i = . visible;however, the label was not in place. On the inside,of.

the cabinet dopr is,a load list giving the breaker number and -

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its associated component number. The component number is also i l located next to the individual breaker. The NRC team also viewed the panel and breaker that was supposed to be de-energized which was located in the control room behind the vertical boards. The interior of the cabinet also contained the load list and breaker identifications described above.

2. Procedures at Nort;b Anna do not contain a stamp ( s at Surry) to ,

i check off if labels need replacement. Thus, it is less likely l[

to correct labelling deficiencies such as that noted by the NRC

^

I team regarding the label denoting mark number 2-EP-CB-12A which had come off and had not been replaced.

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III. Non-Reportable Event: MaintenanceonIncorrectLubeOliPump(Wrong ,

Train)

A. Discussion Due to a high differential pressure across an oil strainer,

! maintenance hid been scheduled on a lube oil pump in Unit 2. The

' pumps for each unit are the same and are located in separate pits, that are accessible by ladder. The Unit 1 pumps have mark numbers at the top of the ladder whereas Unit 2 does not. However, when inside the pit, identifying mark numbers exist on the pump. The '

maintenance workers entered the wrong pit and began to perfonn, work on the pump associated with the wrong unit. The breaker to the pump -

that was requiring maintenance had been de-energized; however,

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the red and green lights indicating whether the pump was running j

were both off because of a problem with either the switch or the bulb. Knowing that the breaker had been de-energized, the workers assumed the pump they werz about to work on wa's off. The noise i _

level in the area is high with the main pump running; however, the*

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mechanics were wearing ear protection and only the auxiliary oil

, pump was running; the main pump was off. The mechanics proceeded to remove a bolt from,,the pump and realized the pump was running when-oil came shooting out of the bolt hole.

To reduce the likelihood of this event recurring Unit I and Unit 2, signs with associated mark numbers have been placed on the wall in the cubicle.

4 B. NRC Discussions With Plant Staff . __' _

The NRC team did not speak with the workers who peho~rmed the

( incorrect action.

. - - , - - - . - - - - - - _ . - - ..,..-----.-,-r .-,--......_,-e,--..-. --r- - ,-w---m .-. ne-mw +- , 1.m.m.-,,,-. m41-- ,-r. ,r,- - - - - , <

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C. NRC Observations The NRC team was unable to view this area due to radiological considerations.

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