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Insp Repts 50-498/89-41 & 50-499/89-41 on 891001-31.No Violations Noted.Major Areas Inspected:Plant Status,Onsite Followup of Plant Events,Licensee Action on Previous Insp Findings,Monthly Surveillance Observations & ESF Walkdown
	ML19332C348
Person / Time
Site:	South Texas
Issue date:	11/21/1989
From:	Holler E; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML19332C347	List: IR 05000498/1989041; ML19332C346;
References
	50-498-89-41, 50-499-89-41, NUDOCS 8911280002
	Download: ML19332C348 (33)
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APPENDIX U.S. NUCLEAR. REGULATORY COMMISSION.

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. REGION IV l

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_NRCInspectionRep6rt:

50-498/89-41 Operating License: NPF-76-l

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50-499/89-41'

NPF-80-

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Dockets: 50-498 50-499 j

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Licensee: Houston Lighting.& Power Company (HL&P)

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P.O. Box 1700

Houston.1 Texas 77001 l

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' Facility Name: South Texas Project (STP), Units 1 and 2 l

Inspection At:, STP, Matagorda County, Texas

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Inspection Conducted: October I-31, 1989

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Inspectors:

J. E. Bess, Senior Resident Inspector, Project Section D

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Division of Reactor Projects J. I. Tapia, Senior Resident Inspector -Project Section D Division of Reactor Projects i

R. J. Evans, Resident Inspector, Project Section D Division of Reactor Projects.

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D. M. Hunnicutt, Senior-Project' Engineer, Project Section D--

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Division of Reactor Projects Approved:.

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E. Jy Holler, Chief, Project Section D Date'

Division of Reactor Projects l

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8911280002 891121 DR ADOCK 0500

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b Inspection Summary Inspection Conducted October 1-31, 1969 (Report 50-498/89-41; 50-499/89-41)

Areas Inspected: Routine, unannounced inspection of plant status, onsite followup of plant events, licensee action on previous inspection findings,

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onsite followup of written-reports of nonroutine events, monthly maintenance observations, monthly surveillance observations, operations safety

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verification,' installation and testing of modifications, engineered safety

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feature system walkdown. nonroutine reporting program, and plant startup from

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refueling.

l Resultsi Within the areas ' inspected, no violations were identified. The inspectors noted a number of discrepancies regarding the vibration and loose parts nonitoring system (paragraph 8).

Procedure discrepancies associated t

with the issential cooling water and other systems were identified to the licensee for inclusion in the licensee's ongoing procedure upgrade program (paragraphs 10 and 12). The licensee's actions regarding a Part 21

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notification concerning Limitorque supplied melainine torque switches appeared i

appropriate (paragraph 11).

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DETAILS l

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Persons Contacted

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. J.-R. Lovell, Technical Services Manager

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W. H. Kinsey, Plant Manager

A. C. McIntyre. Manager, Support General

T. J. Jordan, Manager, Plant Engineer '

J. W. Loesch, Manager, Plant Operations t

C, G. Brown, Outage Manager

W. S. - Blair, Maintenance Support Manager

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L. G. Weldon, Operations-Training Manager

D. A. Leazar, Reactor Support Division Manager

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A. K.= Khosla, Senior Licensing Engineer

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S. L. Rosen, Vice President Nuclear Engineering and Construction

R. W. Chewning, Vice President, Nuclear Operations

A. W. Harrison, Supervising Licensing Engineer

J. E. Geiger, Genera 1' Manager, Nuclear Assurance

C. A. Ayala, Supervising. Licensing Engineer j

In addition +,o the above, the inspectors also held discussions with various licensee, architect engineer (AE), maintenance, and other contractor personnel during this inspection.

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Denotes those individuals attending the exit interview conducted on

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. November 1,1989.

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Plant Status

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Unit 1 began this inspection period in Mode 5-(cold shutdown) with reactor reassembly essentially complete following the first refueling outage. 0n October 14 1989, zero power physics testing was conraenced and the reactor was taken critical on October 15, 1989.. Unit 1 entered Mode 1 operation

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on October 17, 1989.- Reactor thermal power was increased in stages and, at the close of this inspection period, Unit I was operating at 96 percent reactor thermal power.

l Unit 2 began this inspection period at 100 percent reactor thermal power.

On October 13, 1989, the unit tripped due to negative flux rate from a

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dropped control rod. Unit 2 entered Mode 1 on October 15 -1989, and reached 100 percent reactor thermal power on October 17, 1989, and l.

remained at that level through the close of tL!s inspection period.

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Onsite Followup of Plant Events (93702)

On October 13.-1989, an engineer reviewing completed safety injection check valve surveillance tests prior to taking Unit I critical after the first refueling outage, noted that all the test results indicated zero leakage. This raised his suspicions and he and a colleague questioned the operators and shift supervisors involved regarding how the tects were

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perfonned.. The engineer's suspicions that the test equipment was

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malfunctioning were strengthened after the equipment was demonstrated for

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him and he raised his concerns through his. management to the plant

- manager. The startup was delayed to investigate the concerns which t

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ultimately proved to be well founded, in that the test equipment was, in i

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. fact, malfunctioning. The test equipment was repaired, the test rerun, and i

the startup, subsequently completed on October 15, 1989.

On October 13,1989, at 5:45 p.m., 'Jnit 2 tripped f rom 100 percent reactor l

power. The trip occurred when two of four power range neutron monitoring-channels detected high neutron, flux negative rate.- The turbine tripped on i

the reactor trip and feedwater isolation valves closed on low reactor

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coolant system everage temperature. Auxiliary feedwater was initiated

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approximately 10 seconds after the reactor trip from icw steam generator t

water level as expected. Approximately 4 minutes following the reactor

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trip, the operators closed the main. steam icolation valves to prevent

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- excessive cooldown. No safety injection actuation occurred and the plant c

was stabilized in Mode 3.

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The licensee examined the sequence of events (SOE) report and other

computer alarm logs to determine the cause of the negative rate trip. Due

to computer scan frequency limitations, no conclusive evidence was found

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to confirm whether one or more control rods had inadvertently fallen into

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the reactor core. Operations personnel inspected the rod control system power supplies (two motor-generator sets). Both were found running normally and supplying the required voltage with no indicated faults. The rod control system power cabinets (five) were then inspected and no blown fuses nor abnormal conditions were detected. The reactor trip breakers were reclosed and all seven rod banks were sequentially withdrawn to six steps (approximately 4 inches) and reinserted in an effort to isolate any dropped control rod (s) because of rod control system failure. All rods

(57) responded on the digital rod position indication (DRPI) system.

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Because an intermittent failure affecting one or more control. rods was i

indicated, the licensee inspected the rod control system power cabinets r

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for loose connections. When no loose connections were found, the resistance of all stationary gripper coils was measured from the power The power cabinet direct

cabinets,noabnormalreading{swerefound.fourpercabinet)weretested. One of the current (DC)powersupplies

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power supplies (backup) was found inoperative and replaced. Later, the

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licensee determined that this was unrelated to the reactor trip because the primary) power supply was operative and did not lose alternatingThe license current (AC supply.

rod control system (Control Rod Drive Mechanism Timing Test) to obtain DC current profiles of the stationary and moveable grippers and lift coils for each mechanism. No abnormal conditions were detected from the resulting data, i

The licensee installed a recorder to monitor stationary gripper circuits and isolate the intermittent failure upon recurrence. At approximately 6:09 a.m., on October 15, 1989, while withdrawing Control Bank A rods as

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part of a reactor startup, Rod F-8 dropped from 21 steps (approximately 13

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inches). All withdrawn rods were reinserted. Troubleshooting revealed an open diode in the stationary gripper circuit for Rod F-8.

The diode was replaced and a reactor startup was begun at 6:53 p.m.

The reactor neutron j

flux was mapped at approximately 10 percent power to confirm the DRPI system indication. The licensee intends to inspect all the stationary

gripper circuit diodes in the rod control system during the next scheduled

. outage on each unit.

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No violations or deviations were identified in this area of the

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Licensee Action on Previous Inspection Findings (92701)~

z (Closed)OpenItem 498/8923-01:

" Obtaining Plant Operations Manager

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Approval to Perfonn Tests" - Unit I

_j In NRC Inspection Report 50-498/89-23; 50-499/89-23, a concern was identified in the area of administrative approval for performing work.

The shift supervisor gave technicians approval for perfonning a

surveillance test, but the procedure designated +.be plant operations

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manager to provide the approval. The licensee sv.ated a mvision to the procedure was to be implemented to delete the recairement for the plant

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operations manager signature.

Procedure 1 PSP 05-RC-0458, " Pressurizer Pressure Set 4 Calibration (P-0458),"

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Revision 1, was recently mvised by Field Change Request (FCR) 89 2088.

FCR 89-2088 deleted the requirement that the plant operations manager's signature (and associated concurrence) was needed prior to performing the procedure. FCRs were written to revise other procedures applicable to the

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same concern. This item is closed.

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(Closed) Violation (498/8775-01): -" Chemical Detection System Inoperability" - Unit 1

The improper status of the toxic gas monitors and resulting irop(erability

of the contml room heating, ventilation, and air conditioning HVAC)

system to automatically isolate the control room in case of ar, accident was an apparent violation of NRC requirements. This event was reported in

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Unit 1 Licensee.EventReport(LER)07-22.

The inspector verified that the licensee's documentation of corrective action was appropriate..The licensee had conducted training of key personnel on the proper method.for toxic gas system surveillance channel checking and applicable system indications. The training was documented

on attendance sheets. The licensee had completed corrective fictions

including training personnel, revising procedures, and convening a toxic gas monitoring system task group. Work had been completed as required by MaintenanceWorkRequest(MWR)HE-87035823.

Plant Procedure OPGP03-ZO-0004, " Plant Conduct of Operations," had been revised (Revision 7). Unit 1 LER 87-22 was closed in MRC Inspection L

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t Report 50'-498/88-70; 50-499/88-70.

The inspector detemined that the

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licensee's corrective actions were adequate..This item is closed..

(Closed) Violation (498/8847-01):

Failure to Meet Technical I

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Specifications Surveillance Requirements" - Unit 1 l

The licensee discovered that the surveillance test for the undervoltage

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and shunt trip devices of the reactor trip breakers did not meet the

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staggered test basis requirement and that the 31-day check of fuel oil contamination for the No.11 Diesel Generator Fuel Oil Storage Tank had

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been missed.- This event was reported in Unit 1 LER 87-40.

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The inspector verified that the licensee's documentation of corrective.

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action;was appropriate. The licensee issued a directive to emphasize the requirement that surveillances be performed on their due dates and that

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the use of the grace period be limited. The licensee had completed corrective actions. Procedure OPGP03-ZE-0004, " Plant Surveillance Program," had been revised (Revision 8). Unit 1 LER 87-40 was closed in NRC Inspection Report 50-498/88-70; 50-499/88-70. The inspector j

determined that the licensee's corrective actions were adequate. This

item is closed.

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(Closed) Violation (498/8727-02):

" Failure to Provide Adequate Design Change Control" - Unit 1

A' ciesign change notice issued.in 1980, but never implemented, was erroneously incorporated. As a result of this error, both the Bechtel drawing and calculation indicated a 1/?-inch weld on the column-to-base i

plate connection of certain msidual heat removal (RHR) pump supports while the actual as-built condition was a 5/16-inch weld.

The licensee performed a reinspection of the affected RHR pump supports to verify that the weld size was 5/16-inch. The licensee determined that an

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increase of the weld from 5/16-inch to 1/2-inch was not required. A

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design calculation (Bechtel Calculation No. SC045-4B, Revision 7), was revised to specifically document the conclusion that the smaller weld size

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.was adequate. Drawin9 3001-9-S-1600, Revision 8, was revised to show the

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appropriate size weld. The current revision' for this drawing is Revision 9..The licensee's investigation of this matter indicated that this was an isolated case. The inspector determined that the licensee's corrective actions were adequate. This item is closed.

(Closed) Violation (498/8811-01):

" Inoperable Feedwater Flow Transmitters Due to a Procedural Deficiency" - Unit 1

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A licensed shift technical advisor discovered during performance of precritical feedflow/steamflow transmitter calibrations that 7 of 12 feedflow transmitters had been isolated. This event was reported in Unit 1 LER 87-16.

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The inspector verified that the licensee's dobumentation of corrective

- action was appropriate.; The licensee vai.':ed that TS required

, icstrumentation was in service and TS: instrumentation checklists were-

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Lincorporated into Plant Procedure IPOP03-JG-0001 " Plant Heat-up,"

'A Pevisior: -7.

The licensee also reviewed the status of plant systems and

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-J reperfonned valve. lineups. ~ Unit 1 LER 88-16 was closed in' NRC Inspection

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. Report 50-498/88-70; 50-499/88-70. Th's violation was ' addressed as a part

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cof Eni'orcement Action- (EA)~ No.88-112, dated September 21'; 1988. The inspector detennined that the licensee's corrective actions were adequate.

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This item is closed.

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'(Closed)LViolation(499/8907-01):

" Failure to Follow Procedures /

m Inadequate Procedure"'- Unit 2-l

l Activities affecting quality were not prescribed by-procedures appropriate

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to the circumstances, in that three examples r:re noted where approved

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' procedures were not followed and one example were an approved procedure was not appropriate to the circumstances.

Th'e.. licensee.identif0ed the root causes of each of the four examples in

this ' violation. ~ The corrective actions included:

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The' licensee corrected seven discrepancies identified during the a

review of 53F,rocedures. -The scope of this review-included the

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current revis on of quality-related procedures that had FCRs against the revision ist C just prior to the current revision.

A flaw in the procedure program which permitted an FCR 40 be issued

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against a procedure and not evaluated for incorporation into the next

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revision prior to its issuance was_ corrected by Revision 17 to Procedure OPGP03-ZA-0002, " Plant Procedures."

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' Preparation of Procedure OPGD03-ZM-0028, " Erection and Use of

Temporary Scaffolding," Revision 0, and by revision to Procedure OPGP03-ZI-0002 (Revision 2), " Erection and Use o.f Temporary

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L Scaffolding."

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The licensee completed appropriate corrective actions to preclude recurrence and corrective actions were responsive to the four identified examples in the NOV. This item is closed.

(Closed) Violation (498;499/8911-01):

" Failure to Analyze Possible n'l Effects of Energized Space Heaters Used Within The Motor Operator" -

Units 1 ind 2

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The equipment qualification file for eight auxiliary feedwater valves with

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space heaters installed and energized did not support qualification. This was due to a failure to analyze all possible effects of energized space heaters used within the motor operators. No analyses were contained in the licensce's files to fully establish qualification for air temperature rises of cables, wires, splices, and components resulting from their a

proximity to the heators.

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The. licensee's corrective actions included the following:

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a Th'e space' heaters for the affected motor operated valves (MOVs) were

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t de-energized. Engineering Change Notice Package (ECNP) (four in -

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No. 89-L-0038-A~ required the heater leads.to eight MOVs

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Unit I and four in Unit 2) to be ~ disconnected and identified as spare

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leads. The drawings were revised.- The installation records provided.

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A, instructions for cable, raceway, and terminations installation. The o'

inspector _ verified that the design change checklist was properly'

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LThe affected MOVs were evaluated to determine the impact of the j

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energized space heaters. The review determined that t N energizedT

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space heaters would'not have significantly increased t a temperature l_

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no adverse. impact on the qualified life or operation of any component because of the e::istence of energized space heaters in the DC power L

H Class IE motors.- The valve operator vendor confirmed that no adverse effect on the intended safety function for either the valves or motor p

operators would occur if the heaters were connected and operated.

"l The design was reviewed to determine other instances where space

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heaters had been energized inappropriately. None were identified.

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adequate. This' item is closed.

(Closed) Violation (498;499/8917-01):

Falsified N re Watch Loos

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Entry" ' Units 1 and,2,

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.An individual assioned to fire rotection system compensatory hourly fire

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' watches initialed a fire watch og indicating that he had passed through

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the. required area checking for signs of fire when, in fact, he had not '

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passed through the required areas within the required time interval.

t The_ licensee-determined that the root cause of the failure to perfonn fire watches'and the falsification of the fire watch logs was dereliction of duty by the individual. A contributing: factor was the established practice of leaving a blank space on the fire watch log sheet to highlight that a fire watch round had been missed (" skipped").

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The licensee's corrective action included:

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Conducting:an' investigation to determine the extent of the

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irregularities in the fire watch program. No other findings y

involving other personnel were identified during the investigation.

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n The individuallidentified as having falsified the fire watch logs was-d l

removed from fire watch duty.

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Giving written instruction to fire watch personnel (Office Memorandum

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- Station Problem Report (SPR)89-318, dated June 15, 1989)-to make

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all entries on the fire' watch logs consecutively and to " skip" no _

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lines. Paragraph 5.7.2.4, Procedure OPGP03-ZF-0013. " Fire Watch Program,"wasrevised'(Revision 3)toinclude," NOTE All entries on

4 the " Fire Watch Log" shall be continuous. N0 lines shall be skipped.

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This statement partially fulfills the commitments in response to SPR-890-318..

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Reinstructing' fire watch personnel on the consequences of falsifying

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. records. Training included a review of NRC-Information Notice j

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No. 89-18, " Criminal-Prosecution of Wrongdoing Comitted By Suppliers-

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.of Nuclear Products or Services,"' dated February 22, 1989.

Attendance at this' training was verified by the inspector. from the i

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attendance sheets for course title'"SPR 89-318 - Fire Watch

. Retraining."' New personnel receive this training through general employee training (GET) Category I and other personnel. receive this.

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training through the GET requalification program.

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The inspector determined that the licensee's corrective actions were adequate.. This item is closed.

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t (Closed) Unresolved ~ItemL(498/8873-03): " Installed Plant' Instrumentation

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' Calibration Verification Program" - Unit 1 L

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The failure.to' complete'd'ocumentation of an "as-found," out-of-tolerance

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. condition was comidered unresolved pending further inspection to

determine whether, proper use and distribution of the calibration deficiencyireports were maintained.:

a The inspector did not find evidence to ~ conclude that'the failure to I

complete documentation ~of the "as-found," out-of-tolerance condition was nere than an ' isolated incident.

The licensee's actions regarding this matter included the following:

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a Procedure OPGP03-ZM-0016, " Installed Plant' Instrumentation Calibration Verification Program," was revised (Revision 1) to

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require problem? reports to be written if instruments within the scope t

of the procedure were out-of-tolerance.

FCR 89-0054 and FCRs 89-0058 through 89-0064 were prepared to revise

applicable _ generic plant proceaures. These procedures delineated programmatic controls for calibration and status verification / notification of installed permanent plant instrumentation. The procedures are applicable to

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quality /nonquality-related display instruments used in routine I

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operating procedures or surveillances., Procedures. revised include.

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.the following:,

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OPMP08-ZI-0134,." Generic QDPS Loop Calibration," Revision 0,

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page'9

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0PMP08-ZI-0002, " Pressure Transmitter or Differential Pressure

Transmitter Calibration," Revision.1, page' 6'

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OPMP08-ZI-0007, " Generic Recorder Calibration," Revision 2 o

page 5'

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~0PMP08-ZI-0002 " Pressure and Differential < Pressure Switch-

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Calibration,"- Revision 3, page 6

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y OPMP08-ZI-0011. " Generic Temperature Switch Calibration,"

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Revision 5, pages 6 and 11

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H OPMP08-ZI-0016,l" Generic Indicator' Calibration," Revision ~ 2,

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pages 4 and 6 OPMP08-ZI-0016,'"beneric 7300 Loop Calibration," Revision 0,-

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pages 8 and 11

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OPMP08-ZI-0203, " Pressure or Differential Pressure Indicator i

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L-Calibration,": Revision 4, pages 7 and 10

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The. inspector determined that the procedures developed and revised to

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preclude recurrence were appropriater The inspector had no further

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questions..This item-is closed.

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.(Closed)OpenItem(498/8755-01):

Power Operated Relief Block Valves -

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During,al tour of th'e re'ac' tor containmedt building;(RCB) it was noted that

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the pressurizer power operated' relief block valves exhibited body-to-onnet

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leakageu Further investigation by;the licensee revealed that the

body-to-bonnet fasteners were torqued below their, specified values.

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' The' licensee's investigation determined the following:

The valve outlet centerline to inlet flange face dimension tolerance.

of 1/4-inch to mirus 1/B-inch may contribute to permitting leakage

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when replacement valves are installed.

.4 The second valve installed in May 1987 had been machined on the inlet

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face approximately 1/8-inch to remove surface scratches.

The discharge side was torqued (after the valve was "placed" on the inlet ~ flange with quality control (QC) verification of gasket

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material) before the inlet side was final torqued. This, in j

conjunction with the machined surface, could have caused inlet piping

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movement / induced. stresses.-

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The criteria (1'.e., 3/64-inch per foot of flange face diameter) may

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not' have been appropriate to this unique valve to pipe flange

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L assembly. The inlet flange gap closure criteria was-revised and R

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-included in the Crosby Safety Relief. Valve Manual..

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TheWestinghouseinstallationinstructions(ST-WN-YB-1142)were

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incorporated into the construction process Sheet No. 0266,

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Revision-0, and documented to be satisfactory by the Ebasco QC -

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'wasinitiated).

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One of the leakingtvalves, 1-RC-PSV-3461; 12-inch outlet flange

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bolted to a 12-inch ring header, was-found with excessive T

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misaligninent. This required weld repcir on the discharge flange of the header assembly. Final-assembly of the valve subsequent to weld repair (Field Weld FW-PVL-0004) required a vertical and lateral pull.

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Nonconformance Report (NCR) No.87-160. permitted cold springing of q

the inlet line,to align valve discharge flange to pipe flange. The

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force on the valve was measured with a load; cell. The' force. required

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The valve, 1-RC-PSV-3451, was reinstalled on the inlet pipe-and-torqued to-500 foot-lbs. Westinghouse completed the as-installed analysis of this. valve and loop seal. A verified copy of this I

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analysis is maintained on Westinghouse Gas Turbine Service Division r

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'r.uditable file as a permanent plant record.- This' analysis verified that the installation and. operation of the valves met ASME Code,Section III, requirements.

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The final as-built analysis and evaluation-of the pressurizer safety il

and relief line piping system evaluation (PSARV) was completed. The-S piping and support results met ASME Code requirements. The ring

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H header assembly 1(pressurizer nozzle, inlet piping, safety valve,

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dischargemanifold,~andsupportsystem)wasdemonstratedadequateand J

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- met ASME Code,:Section III, reouirements.

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The valves were reinstalled using.new gasket material. The inlet (

a ~~(CCHPs), a loss of suction pressure to the CCHPs occurred on " swap p~~

~~over"'from the volume control tank (VCT) to the refueling water storage L~~

~~tank (RWST).~~

~~1;,~~

~~L EThe licensee's investigation determined that the cause of'this event was a-~~

'

~~design error in the piping configuration from the RWST to the CCHP suction which resulted in a high point that released entrained air. Air collected i~~

~~and formed a pocket when the RWST level was reduced below elevation 25' 6" (Unit I was in Mode 5 at the time of discovery).~~

~~The piping configuration on Unit 2 is identical to the piping configuration on Unit 1.~~

-

~~A creview of this event was performed by engineering department personnel.~~

~~The review determined that the RWST volume required by the Technical~~

)

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~~Specifications (TS) was adequate for Modes 1, 2, 3, 4, and 6.~~

~~However,~~

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~~the Mode 5 minimum volume requirement of 122,000 gallons was below the I~~

~~loss'of net pump suction head (NPSH) level. At the timt of discovery, the RWST volume was greater than 458,000 gallors; therefore, the licensee met~~

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~~TS requirements.~~

~~L j=~~

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~~spn~~

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UV The licensee completed the following corrective actions for Units 1 and 2:
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Operating procedures were revised to require a ' minimum volume of

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378,000 gallons in the RWST when the RWST is used as a boration-source in Mode 5 (TS' permits RWST te be out of service when in

~~Mode 5).~~

U A review of suction piping configuratm to other safety-related and
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. Class 7 pumps was performed. No suction line configurations were
'
found that could.cause 'airf binding similar-to the CCHPs line from the
'
RwsT.
%..
K The.RWST to CCHP 8-inch piping was rerouted in accordance with Plant

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Modification No.- 88246,' Contractor Work Request (CWR) No. 005393,"CCP No. 2-M-FST-0285, and final Calculation Package 2-A-30-HNC-0001,
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"RWST. Verification of Level," Revision 0.
,
Startup field Report.PFM M19: verified that, after the modifications

were completed, no loss of suction cccurred under the conditions specified.'
,
.The inspector.'s~ review of the licensee's investigation,' corrective
~~actions..and modificationsi developed to preclude recurrence indicated that.~~
,
1the licensee's-actions were appropriate. The inspector had no further e
, questions'. This item is' closed;^
'
~~(Closed) Unit 1 LER 88-45i " Reactor Trip Due to Personnel-Error While~~
"
. Troubleshooting Qualified-Display Processing System" - Unit 1
.
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! L-On July 19, 1988, Unit I was in Mode'1. A reactor trip occurred as a
. result of'a troubleshooting ' activity when a technician reset the wrong-y
_. qualified display processing system (QDPS) processor.
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-The licensee's investigation determined the following:

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On July' 18, 1988, two technicians were calibrating the reactor.
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. coolant-system (RCS) hot leg ~ temperature inputs to QDPS. The

technicians completed the: calibration of the RCS Loop B inputs to
"

QDPS: Processor D.
The. procedure did;not contain specified steps to
!
[N ensure proper reset. The technicians were unaware of the proper
.
I.
method to restore the system. The technicians subsequently reset RCS M
. Loop B instrumentation channel with QDPS Processor D " locked up."
hX After the shift change, another technician was assigned to continue

k the troubleshooting.' While troubleshooting,(the technician L
incorrectly placed QDPS Processor B in test the QDPS processor L
associated with the RCS Loop B temperature was Processor D). The K
technician then informed the control room operator that the.
L technician was going to reset the QDPS processor associated with the
'a p'
RCS Loop B.
The technician then returned the QDPS processor to u
normal and operated the reset switch.
Reset of QDPS Processor B
[
caused the calculated value of T-hot to momentarily go to zero while Jr p
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reactor trip system (RTS) Channel-II was in' test., This resulted in a'
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> reactor trip on over temperature / delta temperature.
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_ The root causes of the reactor trip were selection of the wrong QDPS

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cabinet by a-technician and-a technician returning the RCS Loop B hot q"
"r '
41eg temperature channel to service with the QDPS Processor D " locked.
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up."
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The licensee's' corrective actions included:
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Holding training imediately after the event, for the technicians to
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review the event and identify _the need to exercise care when
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resetting QDPS. Course attendance records were reviewed by the
inspector.,
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Revisis Procedures 1 PSP 02-RC-0410, " Delta T and T Average Loop 1
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Set-1 ACOT'(T-0410)," Revision 2;,1 PSP 02-RC-0420. " Delta T and
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T Average _ Loop-2; Set 2 ACOTs(T-0420) " Revision 2;'1 PSP 02-RC-0430, j
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" Delta T and T Average Loop 3 Set 3 ACOT (T-0430)," Revision 2;' and M
1 PSP 02-RC-0440, " Delta T and >T Average Loop 4 Set 4 ACOT (T-0440),"
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Revision 2, which required QDPS: processors.to be reset during
.
restoration.- to include specific steps to ensure proper reset during-q-restoration. Procedures for Unit 2 were. prepared and are same title-Y and number. except for Unit 2 designation and Revision 0.
EInstructing' Instrumentation and Control- (I&C) foreman to verify the

actions;of technicians _which could affect'the operation of the RTS.
Evaluating.the maintenance procedures which controlled work perfonned
)

on equipment which could cause spurious trips for the inclusion of buman factors aids..
Adding clarification of independent verification requirements and M

additional requirements to be observed during troubleshooting of RTS-
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and engineering' safety features (ESF) equipment to i
Procedure 0PGP03-ZM-0021, " Control of Configuration Changes During
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Maintenance or Troubleshooting."
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Placing.a board, labeled with four color coded protection channel key
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sets for protection Channels I,-II, III, and IV (separation Groups A, e
D, B, and C, respectively), in the Units 1 and 2 control rooms. The-key sets included keys for.the QDPS and Westinghouse 7300 process control-(reactor protection system) and are: tagged using separation.
group colors to match the respective cabinets. Personnel have access F
to only one set of keys at a time. A key. log is maintained.
<
The inspector verified the location (east wal; acar north wall) and labeling of the. key boards and keys in the control rooms of Units 1 and 2.
.e The' inspector observed a technician sign out the key in Unit i for the auxiliary shutdown panel (ASP) and accompanied the technician on an L
inspection of the Unit 1 ASP.

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The inspector's review of the licensee's investigation, corrective
.*
actions, and modifications developed to preclude recurrence. indicated that-the licensee's' actions were appropriate.. The inspector had no further questions. This item is closed.
.;
--(Closed) Unit 2'LER 89-01:
" Partial ~ Loss of Offsite Power Due to a Fire-m Protection System Actuation" - Unit Z
'
On January 6,1989, Unit 2 was in Mode 5 prior to initial. criticality. An
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. actuation of the deluge system occurred during restoration of an equipment-clearance on a fire protection panel following replacement of the Unit 2
.
standby transformer heat detectors.
Immediately following the actuation,
. an arc over occurred on Phase B 345 KV transfonner bushing. This arc over was followed by a transformer iustantaneous phase overcurrent and lockout
~!
relay' actuation, which resulted in automatic protective action to clear
,
the switchyard south bus and de-energize the faulted line. The licensee-verified that no fire occurred and actions'were taken to secure the deluge
,
system and restore the plant to normal operating conditions.
'
.Deenergization of the Unit 2' standby' transformer resulted in loss of the i
normal power supply to the'ESF Trains B and C.
The diesel generators:(DG)
,
. started and power was restored to the ESF equipment. The' effects of the
'
switchyard voltage transient associated with the fault were observed in'
Unit.1.. Some' HVAC equipment-in' Unit 1 tripped off. and was subsequently s
restarted.
'
The licensee's investigation determined that~no panel malfunctions existed and that the standby transformer thermal. detector alarm signal was present
'
at the time the clearance was being released. An unlicensed reactor plant operator installed the fuses to restore the actuation circuit. The operator was.not familiar with the fire protection panel. This operator
'
did not-reset the panel to clear the alarm signal. The operator did not realize that the panel must be~ reset prior to reactivating the actuation
'
circuit. Th'e licensee identified a lack of procedures in the area of fire
protection panel restoration. The licensee determined that the
. positioning of the standby transformer delugeLnozzles resulted in water
,
spray on the Phase B bushing, i
The licensee's corrective actions included:
q Developing a procedure (OPEP03-FP-0030), " Transformer Deluge Water H
l

Spray: Actuation Verification and Valve Reset," Revision 0, dated j
,
August 1, 1989, to control the restoration of fire protection panels.
Developing a procedure (OPOP02-FA-0001), " Fire Detection System,"

l Revision 0, dated May 30, 1989, to provide guidance for fire j
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detection system operation.
Repositioning the delu e nozzles to reduce impingment on the Unit 2 L
standby transformer bu hings. The n'zzle configuration on the deluge r
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' systems for.other large transformers were. visually examined to ensure.
that nozzles are not directed at the transformer bushings.
lant operator-(RPO)
. Incorporating this event'into the reactor p(attendance sheets verified
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.requalification program. Shift briefings

training'on " Fire Detection Instructionsefor Closure of SPR-89-0001")L were held on fire protection panel operation (actions for; fire,
'

trouble detection or alarm activation).
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>The inspector's' review of the licensee's investigation, corrective
.
. actions, and procedures developed to preclude recurrence determined that
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the' licensee's cactions were. appropriate.
The inspector had no further-questions.. This item -is elosed.
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L (Closed) Unit P. LER 89-20: :" Reactor Trip Due to a Simultaneous Trip of
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V Three Feedwater Pumps" - Unit 2
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On August 29,;1989, Unit 2'was in Mode 1. J All three operating. turbine-J driven feedwater pumps (TDFW) tripped. The licensed reactor-operator immediately(SG): level, tripped the-reactor in anticipation off low steam

'
generator-i The' licensee's investigation determined that:
~~The TDFW pumps tripped on overspeed following a momentary'~~
interruption of control-power from'an inverter. 'The overspeed
.'
circuitry was designed to fail to-the tripped condition on loss of control power. This protection circuitry was added after,a r
destructive feedwater pump turbine overspeed which occurred on Unit 1
,7
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on May 25, 1988.
The interruption'in power from the inverter was attributed to a

component failure in the static transfer switch circuit of the inverter which feeds the local centrol panels for the SG feedwater pumps. ' A contributing cause was the design of the feedwater pump overspeed protection. circuit which was not designed to remain in
.
operation during a momentary loss of control power without tripping j
the pumps.
The SG feedwater pump overspeed protection circuitry was changed h

~~-(WRs LP-90891, LP-90892, and LP-90893 and ECNP 89-J-0279 for Unit 2~~

and WRs<LP-84062, LP-84068, and LP-84057 and ECNPs 89-J-0278 and
,
.
89-M-0283 for Unit 1)~to an " energize to trip" scheme on Units 1 and 2.. A mechanical overspeed trip for the overspeed protection for the feedwater turbines does not require a power source.
In addition, loss of the electrical bus feeding the feedwater pump turbine electrical overspeed trip also results in a loss of control power which causes the feedwater turbine control valves to close.
The inspector's review of the licensee's investigation, corrective actions, and modifications installed to preclude recurrence indicated that
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m Lthe licensee's actions were appropriate. -The inspector had no further
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~ questions. >This item is-closed.-
(C1' sed) Uriit 2 LER 89-21:
" Reactor Trip Due to a Defective Feedwater.
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Pump Speed Controller Card Edge Connector" - Unit 2
,
On Septeinber 5~,1989, Unit 2lwas in Mode 5.
Control room operators
"'
observed speed oscillations 'on turbine driven TDFW 21. The TDFW subsequently tripped on overspeed. The resultant loss of SG level caused-a reactor' trip and auxiliary feedwater system actuation. The turbine
'

tripped following the reactor trip. The ;feedwater isolation valves closede
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< on low, reactor coolant system average temperature.
ca LThe licensee's investigation determined that:
,
Prior to this event, TDFW 22 had exhibited erratic speed control.
~~The speed control circuitry for all three TDFWs is housed in a comon card frame. -Technicians were_ troubleshooting the speed control~~
,
problem. One of the printed circuit (PC) cards for TDFW 22 was removed.
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During removal of the PC card for TDFW'22,.a defective edge = connector on a circuit card for TDFW 21 caused erratic speed controller output and the erratic TDFW speed oscillations. During the posttrip:

-troubleshcoting, the effects of disturbance of the PC card frame on TDFW 21 speed controller output was verified.
. The defective PC card edge connector on TDFW 21 was repaired. The
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card frame alignment was-checked and the remaining printed circuit cards and edge connects were inspected. No other defective edge l'
connectors were found.
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.The cause.of'this event was the defective TDFW 21 speed controller PC
,
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- card edge ' connector which was inadvertently. moved during troubleshooting of a PC card for a different TDFW in the same card frame.
e The inspector's review of the licensee's investigation, corrective
, actions, and procedures developed to preclude recurrence determined that the licensee's actions were appropriate. The inspector had no further y.
questions. This item is closed.
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'(Closed) Unit 1LER88-28:
" Leakage of Aluminum-Bronze Essential Cooling Water System" - Unit I ~
On April 1,1988, Unit I was in Mode 3.
Plant operations personnel observed slight leakage occurring at a number of locations in the aluminum-bronze essential cooling water (ECW) system. This event dio not require that an LER be submitted by the licensee. The licensee submitted Unit 1 LER 88-28 as a voluntary LER.
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TheJ11censee'sinvestigationdeterminedthatsomesmallbore.(2-inch-
'
' diameter and less) fittings and valves-in the' ECW system had undergone
+
crevice corrosion--(dealloying). The dealloying permitted through-wall-seepage'of ECW.
The licensee's corrective actions included the following:-
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Specifying,the material 1for the replacement fitting and. valve
'
components:-to' meet'ASME Specification SB-16g Alloy C61400 (wrought
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aluminum-bronze) requirements.
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Replacing cast aluminum-bronz'e components with fittings and valves
-.
machined from _ wrought aluminum-bronze. :
Issuance of.FCRs by design engineering to delete or replace the cast
'
aluminum-bronze components.
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Tracking the disposition of each cast aluminum-bronze component by
- engineering on the master computer list.
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~~The inspector selected five components and reviewed the engineering and~~
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receipt' inspection, documents associated with the disposition and
~!
-replacement of.the aluminum-bronze cast components. The certified i
material test: reports for the four fittings 'and one valve selected for review were consistent with the ASME requirements for Alloy C61400 wrought aluminum-bronze. iThe inspector reviewed the Bechtel Engineering i
systemmatic method for assuring that all safety-related small bore-aluminum-bronze cast components were identified on a master computer list.
This ovent for Unit 2 was reported as Independent Review Committee Report (IRC) IRC-433, " Leakage of_ Aluminum-Bronze Essential Cooling Water
!
-System" Unit 2.
An inspector reviewed the corrective actions for

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aluminum-bronze fittings and valves in Unit 2 and closed IRC-433..in NRC j
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H Inspection Report 50-498/88-79;'50-499/88-79.
The' inspector reviewed the licensee's investigation, and corrective
actions, to preclude recurrence. The review determined that the licensee's actions:were appropriate. The inspector had no further

questions. 'This item is closed.
j i
6.
LMonthly Maintenance Observations (62703)
The inspector observed selected maintenance activities to verify that'the activities were being conducted in accordance uith approved procedures and
.TS.
The activities observed included:
L
,
MWR PD-71342; troubleshooting of the 4.16 KV electrical breaker which supplies RCB Chiller 11A.
Improper functioning of the breaker interrupt racking mechanism was causing the elevating fuses to fail
'open on overload wheneur an operator attempted to rack the breaker in or out.
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Preventive Maintenance (PM) Mi-1-HF-89002343; lubrication and i
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L inspection of Train C fuel handling building-(FHB) booster fan outlet j
E damper.
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~~MWR AF-83553; realignment of-the pump to motor on Auxiliary~~
~~i Feedwater Pump No.13.~~
The inspector verified that the activities were conducted in accordance with approved work. instructions and procedures, test equipment was within its current calibration cycle, and housekeeping was maintained in an l acceptable manner..

e No-violations or deviations were identified in this area of the inspection.
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7.
Monthly Surveillance Observ$tions - (61726)
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Selected surveillance activities were observed to ascertain whether the surveillance of safety significant systems and components were being
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coaducted.in accordance-with'TS and other procedural requirements. The-surveillance' activities observed included:
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2 PSP 02-RC-0419, "RCS Flow Loop 1 Set 3 ACOT (F-0419)," Revision O '

~~'1 PSP 03-AF-0007, " Auxiliary Feedwater Pump 14 Inservice Test,"~~

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Revision 5
^
.2 PSP 03-SP-0006S, " Train S' Reactor Trip Breaker TADOT," Revision 0

Specific items inspected included verifying that as-left data was within acceptance criteria limits, the acceptance criteria listed in the procedures agreed with values listed in desige documents or instrument
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setpoint indexes, and test equipment used'was within its current calibration - cycles.
Following observation of the surveillance activities,
'
the procedures were reviewed' for technical accuracy and conformance to TS requirements.
Items noted and discussed with the licensee included:
-Procedure 2 PSP 02-RC-0419 by I&C technicians.to verify that the

accuracies of the reactor coolant low flow trip logic circuitry were within acceptable limits. Several minor observations were reparted to the licensee: Step 6.2.d called Flow Indicator FI-0417A " Loop 1 Flow," but the indicator was labelled," Loop 2A Flow" on the Unit 2 g
control panel. On page 3 of 3 of the Analog Channel Operational
,
Test (ACOT) data package, the values-listed fo; the TS allowable s
value and Comparator Input Trip Voltage varied slightly from values listed in Instrument Loop B2RC-F-0419. Revision 2.
The TS allowable value listed in the procedure was 3.296 vde, but was listed as 3.295 vde in~ the instrument loop. Additionally, the comparator input trip voltage was listed as 5.853 vdc in the procedure, but was 5.852 vdc in the instrument loop. The differences in value were due
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to. rounding'off calculated values.. The values listed in the procedure were more conservative than required by the instrument
.
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loop.
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- Performance of Procedure 1 PSP 03-AF-0007' by Unit 1 operations l
personnel to verify operability of Auxiliary Feedwater Pump 14.-
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associated valves, and flowpaths. Typographic errers were cbserved instep 5.12.2(incorrectnameforCrossConnectValveFV-7518).and'

m
indatasheet.(-2)onpage21(Step 5.12.3shouldhavebeen
.
.
Step 5.12.2andthe"y"ofword"by"wasmissing'fromStep5.12.2).
!
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Step 5.9.5 in the procedure was revised by~FCR 88-1739. The change
'
was made to verify whether Valve AF0011' had returned to aifull flow
,
position or notC A review of FCR 88-1739 was performed. 4A'
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description of the change made to Step 5.7.5-was not included in the
!
" Description of Changes or Reason For Changes" blants of the.FCR.
d Although Step 5.7.5 was revised in the body of the procedure, the'
. intent of the step apparently~ did not change.
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Step 5.7.1.6 of Procedure IPSP03-AF-0007 instructed the' operator to-
. start Auxiliary Feedwater Pump 14 by opening Turbine-Trip / Throttle Valve MOV-0514. The pump failed to start on the first two attempts.
A mechanic was dispatched locally to Valve MOV-0514 to determine why
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.the valve was shutting immediately after travelling to the full open
'
position. With a mechanic present, two more starts-were
'
- unsuccessfully..The mechanic determined there was too much play in the trip mechanism and made a manual adjustment. On the fifth attempt, the pump turbine started and the valve remained open as required'by the procedure. A review of the procedure and vendor

. manual' did not reveal any limitation on.the number of pump or turbine
.
starts within a period.'
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Performance of Procedure 2 pap 03-SP-0006S by Unit 2~ operations
.personne.~. The test simulated the trip functions of the Train-S
'
l Reactor Trip Breaker.. The inspector observed'the test being
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performed and~ compared the procedure to Vendor Manual- 0387(2)00002-KWN, " Westinghouse Three Train SoliJ State Prctection System Technical Manual," Section 6.0, " Testing and Troubleshooting." Several differences between the manual and the procedure' were observed. Fec example, procedure Steps 7.7.20 (turn
,
logic switch eff) and 7.7.21 (reclose trip breaker) were reversed in L
d,
- the vendor manual. The licensee stated that the revarsal of steps had no co7 sequences in the logic test sequer.ce. Step 7.7.36.a l-instructed the test performer to verify that a meter reading was less

than 6 volts, but the vendor manual instructions used a value of 1 volt, not 6 volts. The licensee stated that the 6 volts reading
,
L was based on a meter tolerance of + 5' volts, therefore,1 volt plus L
the tolerance of 5 volts established an upper limit of 6 volts.
L Step 7.8.5 instructed the test performer to verify a meter reading of
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43 + 5 vdc. The vendor nenual instructions used a value of 43 +
2'vic. Again, the tolerance of 5 volts was used by the licensee for h
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' the. meter, notL the value (2 vde) listed by the manufacturer. - Also,
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several final checks listed in the vendor manual were:not performed R
in the procedere, includir:g verifying' that urgent alann light goes i
'1 off and.the' test indicator lights come on ':The licensee stated that
' verification of these final indicstor lights status was determined.
'
not to be. required.
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i No' violations or-deviations wereiidentified'in this area of the i
inspection.
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' 8.
Operational Safety Verification (71707)
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-The purpose of this-inspection was to ensure that th'e facility was being-
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~~operated safely'and_ in conformance with licensee and regulatory~~

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requirements. This inspection also included verifying that selected
,
-activities of the licensee's radiological protection program were being
~~implemented in conformance with requirements and procedures, and that the licensee was in compliance with its approved physical security plan -~~
,
The inspectors visited the control room on a daily basis when o'nsite and j
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verified that control room staffingi operator behavior, shift turnover,
~~adherence to TS limiting conditions for operation, and overall control~~
.'
room decorum were being conducted in accordance with requiremerts-
~~Tours,were conducted'in various locations of the plant to observe work and~~
'
operations:to ensure that the facility was being operated safely and in-
.
conformance with license and regulatory requirements. The following items ws.re observed and discussed with licensee representatives who took appropriate corrective action.
'
The-inspection and review cf the Vibr.dion and loose parts monitoring
-
! system (VLPMS) and associated procedures were performed during this j
einspection period. The VLPMS consists of permanently. installed sensors i
that monitor selected locations on the reactor vessel'and S0s, and the o
control room electronics that receive the sensor's' signals. The system is-designed to detect excessive motion of core internals and noise resulting.
from loose aarts in the'RCS. The system is..nonsafety-related and is not-listed in.tle TS. However, the syster is described in Section 4.4.6.4 of
,
the Final. Safety Analysis Report (FSAR) and was designed to meet the
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guidanceofRegulatoryGuide(RG)1.133," Loose-PartDetectionProgramfor the Primary System:of-Light Water Cooled. Reactors."
Electronics panels are loc'ated in.both the Unit I and 2 control rooms.
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The electronics in the panel. ncluae a. vibration and loose parts monitor, i
tape recorder, and loose parts locater printer. The monitor provides alarm indication, channel ' selection switches, a speaker for audio
' listening cf selected channels, a deciblemeter to display the vibration level, and power and test lamps. The. tape recorder is useo to simultaneously record those signals or combination of signals in the event
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of an alarm that is not readily interpretable, or for which additional analysis or permanent record is desired. The loose parts locator is a
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Emicroprocessor-based subsystem that assists the operator in locating loose p;
. parts 6tected by the VLPMSL The locator will automatically. print n4ssages with a. printer mounted in the electronics. cabinet. A separate
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spectrum analyzer can be. connected to the VLPMS circuitry to provide o
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-permanent signature analysis plots.
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As part of the VLPMS, inspection, the following procedures were reviewed:
OPEP02-IB-000lb" Vibration and L'oose Parts Monitcring System Data

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manually test the VLPMS indicator lights by pressing the TEST, then RESET, buttons. Step 5.5 provided instructions'to " place the MANUAL-AUTO switch in the AUTO position." Per the vendor manual, "the MANUAL / AUTO switch
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~~should be toggled to MANUAL then AUTO efter the reset." Step 5.5 could~~
- have been more specific to ensure that the circuitry was proporly placed-in-the AUTO condition. ' Steps 6.4 and'6.5 orovided instructions to manually. record data on the selected tape recorder channels.- The
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procedure failed to specify which of,16 chennels to record. Additionally,
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the procedure failed to specify that the AUT0/ MANUAL switch had to:be in MANUAL position for manual recording. The restoration of the recorder to-
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. automatic operation following. manual operation'was also not mentioned in-a the' procedure.
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Procedure OPGP03-ZE-00019 describ'ed the VLPMS program, including.

responsibilities, data collection requirements', periodic inspections and-tests, and alarm response instructions.
Section 8.0, " Alann Response,"
4 -
provided detailed.information on how to respond to thetalarm LOOSE PARTS
,
SYS TRBL-located on Control Room Panel CP-009. The responses included.
E notifying the shift technical advisor (STA), determining that the initial-
'and subsequent alarms actuated, and recording data if necessary. Alarm'
Response Procedure IPOP09-AN-09H1 for this alarm (Location B-8) did not

mention any of the requirements estabitshed by OPGP03-ZE-0019.
The alarm response. procedure, as a minimum, should have referenced OPGP03-ZE-0019, Section 8.0.
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As a result of discussions with licensee personnel and:through visual
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observations'of VLPMS operation, the inspector noted, that the system was
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L not being maintained in its normal mode.of operation. Normally, the system should be'in the AUTO mode of operation.
If an alarm was received,
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a first out-indication light would energize, the Alarm B-W on CP-009 would
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actuate, and the tape recorder would automatically-start to record the
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l alarmed channels. However, the licensee routinely left the VLPMS monitors in MANUAL' mode of operation, which defeated the purp:se of the AUTO functions. - On the day of inspecticn, both the Unit I and Unit 2 VLFMS monitors / recorders were in MANUAL modes. The Unit 1 monitor-had 10 of-16 t
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alarms present and the Unit 2 monitor had 7 of-16 alarms present.
Additionally, the loose parts locator printer was supposed to be in
. service with the OFF LINE and TEST buttons in the out positions. The n
L Unit 1 printer was found without paper and with the UFF LINE butten depressed in the IN position.
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At the request of the inspector, the licensee placed the Unit 1 VLPMS K
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L, system in AUTO mode of operation. Within minutes, the system automatically. started up on a high vibration alarm. The system worked as i
c designed: the first.out indicator illu.ainated, the control room alarm actuated.and the tape recorder started. Apparently, the alarm, like all L,
others previously actuated, was spurious and was not indicative of.
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excessive vibration or-loose parts. The. system was then returred to MANUAL to reset the tape recorder tape position. The inspector questioned licensee personnel on the use and current condition of the VLPMS monitors.
The inspector believed that the system was inoperable or incapable of performing its intended functions, for several reasons:
the monitors were left in MANUAL, the recorder was not being utilized, the VLPMS sensor setpoints were apparently incorrectly established, and operations i P
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personnel were not routinely respondir.g to the VLPMS alarms.,The licensee i
offered several:cocnents regarding tu system including:. (1) daily and

weekly channei checks were being perfomed-and.if a -loose part was
.
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p present,itprobablywouldbedetectedduringthese. checks,(2)despite
.I efforts to recalibrate the sensors, high background noises attributable.to reactor. coolant' pump seals resulted in spurious' alarms (3) _overall, changes.in alarm setpoints were being contemplated, and (4) the licensee
- was aware of the problem and was considering long-term solutions,
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including replacing the current VLPMS with a.different system of

detection.
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System operability is described in FSAR Section'4.4.6.4 and'RG.I.133.
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FSAR specifically describes the' system as operable in Modes 1 and 2.
- RG'1.133 states that the VLPMS systems should be maintained in the'

automatic mode for continuous, online detection of;1oose parts. = Also, the manual mode is to be used only periodically, for testing loose parts,.
system calibration, channel checks, detecting trends, and' diagnostic purposes. The operability of the VLPMS.and the licensee's long-term
~
corrective actions will be monitored by the inspectors. This is an open
item-(498;499/8941-01).
No violations or deviations were identified in this area of the
. inspection.
l 9.
Installation' and Testing of Modifications (37828)
Minor plant modifications were selectively inspected to ascertain whether the modification activities were in conformance with established
"
procedures and regulatory requirements. The inspection consisted of direct' observation of the modification work in progress, testing of the modifications following installation, and a review of the completed work packages. The three modifications inspected were on Unit 1.
The
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modifications examined and reviewed included:
ECNP No. 89-L-0011, Modification of 125 vde power supply breakers
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ECNP No. 88-J-134. modification of the main turbine trip system
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~~cabinet power supplies, and ECNP No. 89-F269, modification of the feedwater isolation valve~~
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control circuits ECNP No. 89-L-0011 was developed to comply with a licensing commitment'to the NRC.- The modification consisted of changing the control power for a standby transformer 13.8 kv incoming breaker to a second power supply to maintain independence from offsite power sources. Two sources of DC power
.were required for independence from offsite power sources (the incoming breaker from Standby Transformer 2 would be separated from the power to
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the remaining incoming breakers) to meet the general design criteria of SRP (standard review plan) Section 8.2.1.
The work consisted of disconnection and deletion of selected cables, remos;al and replacement of n"
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.several circuit breakers, installation of. several new cables a'nd ~one '

conduit,. removal of transfer switches, and-reworking three transfer switch
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panels. -~ All-work specified in ECNP 89-L-0011 was verified to be completed, the ECNP was reviewed, and noLconcerns were identified-by the
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. inspectors.-
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ECNP No,'88-J-l'34'was. developed to provide two independent power supplies
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'to the main turbine' trip system cabinet in lieu of two power supplies from one distribution panel. This modification was in response to the licensee's program to minimize single component failure modes which could cause a trip of the main turbine or challenge the primary systems. This
'
modification was designed to prevent a turbine trip due to:the loss;of a
, single non-essential inverter power circuit or a single turbine trip-
, control circuit, thus enhancing' the system reliability.
,
3The work-consisted of installation of.a new branch switchboard unit.
-removal of1 fuses from one distribution panel (DP002) and placement in a
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.second distribution panel (DP001), rerouting of a power supply. cable',.
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installation of a conduit, and rewiring two trip solenoid inverter
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cabinetsi : Portions of the installation were witnessed by the inspector while in progress and:all: modifications were verified correct per the ECNP instructions following installation completion.
Several observations were reported to.the licensee ^ for corrective action, including:
the incorrect
. cable code was noted on two EE580 cards (electrical termination
.
, installation cards); temporary test jumper connections were noted installed in the. trip solenoid inverter cabinets that should have been
,
removed following test completion;' relays in the trip solenoid inverter cabinets were labelled differently on the drawings in the ECNP and on the relays in the field (for example, point "AC(Y)" on the documents was labelled "7" in the field); and the Local Panel DP001 nameplate had six
revisions (plaques or tape over olit names). required a seventh, but should A
have been replaced.
.
ECNP 89-J-269 was generated to add 'a relay to each Feedwater Isolation t
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Valve (FWIV) control circuit. The modification was designed to prevent
'
plant trips during partial closure tests at full power in the event of l
l limit switch malfunctions. The modifications were made in responte to L
Unit 2'LER 89-019. Unit 2 tripped off line when FWIV "C" unexpectedly (
fully closed during partial stroke testing.
The work consisted of relay panel modifications, installation of 4 new
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relays, adding wiring for the 4 new relays, and resettS g the time
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~~settings on 4 other relays. The work was monitored and postmaintenance l~~
testing was witnessed. Observations made and reported to the licensee L
included:
the postmaintenance testing requirements of the ECNP listed the wrong surveillane test, the ECNP required a change to reflect that j
L different relay model numbers are used than those described in the H
engineering documents, and two relays were mislabelled. Tne relay
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l labeling errors were subsequently corrected.
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testing was completed, and~no safety concerns were identified.
!l-No violations or deviations were identified in this. area of the

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inspection.
10.
Engineered Safety Feature System Walkdown (71710)
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A complete walkdown of the ECW system for Unit 1:was performed to
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independently _ verify the status of this ESF system. : The walkdown was
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performed in conjunction with Section 12 of this inspection report (plant (
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'startup following refueling outage) to independently ascertain whether the '
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ECW system'was returned to service in accordance with approved procedures.
H The inspection consisted.of-an operating procedure review, comparison of >
' the operating procedure to plant drawings, and a complete walkdown of the system to verify whether the system was in a position to support plant J
operation.
Specific-items inspected in the field-included determinations l7
'that: valve, switch, and breaker. positions were correct, housekeeping was
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E adequate, and support-systems were in service.
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- A review of Operating Procedure IP0P02-EW-0001, " Essential Cooling Water
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L Operation,"-Revision 8.,was performed. Theprocedurelineups(valve, l
andinstrumentdiagrams'(P&ID)ypositions).werecomparedtosystempiping switch, electrical power suppl l
Observations made in the technical review l
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included:
' Two P& ids were' missing from the references section of the procedure

(P& ids 9F20005No.Iand1F00032). The prerequisite section of the
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~~procedure did not ensure proper operation of support systems (such as~~
-instrumentair).
' Numerous differences in the required positions of valves as shown on
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the P&ID and as listed in thezvalve lineups.were identified. These
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position differences. included normally open versus normally closed and locked tull open versus locked in. position (throttled).
Additionally, some valves did not specify a required position on the o
P&ID.
Butterfly.. valves on 'P& ids require letter designators
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signifying their required position because all butterfly valves are drawn the same. These valves did not have letter designators specifying a. required position. The affected valves included Train A (EW-0188, EW-0001 EW-0020, EW-0259, EW-0031, EW-0280, FV-6914, and
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L FV-6935), Train B (EW-189 EW-0003, EW-0056, EW-0260, EW-0203, EW-0281, FV-6924, and FV-6936), and Train C (EW-0002, EW-0261, EW-0206, EW-0282, EW-0093, EW-0190, FV-6934, and FV-6937) valves of the ECW system. Two valves of the Essential Cooling Pond Makeup
, System EP-0009 and EP-0010, were also affected.
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Typographical errors were observed in the procedure, includin the

wrong power supply was listed for Screen Wash Valve FV-6924 (g:
the procedure checklist (-2)' listed FV-6924 power supply as CPB-135, but
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actually was DPB-335), and the name of Valve EW-0095 was incorrectly l
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was the PI-6898 isolation valve. Labelling discrepancies were observed. - For-example, Fan FN001 'was named "1A-ECW HVAC VENT FAN -
FN0001"'in the procedure but was labelled "RM 105 FAN-11B" on Control-l
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Vent and drain valves are"normally shown as nonnally) closed valves on ~
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P& ids with letter designators (such as "D" for drain to identify
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what;the valves are. The following valves were noted to be' missing
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their letter designators:
EW-216, EW-220, EW-0376, and'EW-0377.:

Valve 1-EW-0367 was deleted from:the system and-the procedure
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checklist (-2)~ 1n accordance with FCR-2232, but the-valve was not e-deletedfrnmtheventchecklist(-6).
Valve EW-268 was deleted from
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the vent checklist (-6) in accordance with FCR 89-2507, but was not M
deletedfromthevalvechecklist(-3).-
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A walkdown of the ECW system was performed, observations included:
g The identification tag for EW-0375A was badly rusted and could not be
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a read. Valve'EW-0139 was missing 1t's identification tag. The following pairs of valves were noted to be tagged in reverse in the
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field:' EW-0235.and 0234, EW-0289 and -0290, EW-0237 and -0236,-
. EW-0299 and :-300, cod. EW-238 and -239.
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Several nonsafety-re1ated components were found to be missing from

-the procedure checklists (but were in the correct position to support
,
system operation): Distribution Panel DPA-335, Breaker No. 5, Train'
A Room Intake / Exhaust' Air Dampersf(ventilation power supplies supportingEWsystemwereconsideredpa.rtofthesystemprocedure);
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DPB-335, Breaker No. 5, Train B Room Intahe/ Exhaust Air Dampers;
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DPC-335,: Breaker No. 5,; Train C' Room Intake / Exhaust Air Dampers; l
Valve EP-0011, ECW Sump Pump B' Discharge' Pressure Isolation; and
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L Valve EP-0012, ECW Sump Pump A Discharge Pressure Isolation.
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The valves, EW-100 and EW-096, were deleted in the field but were

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" still listed in the-procedure checklist (-3). Several valves needed
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lP to be added to the~ valve checklists (valves added by addition of
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corrosion moni*oring, equipment) that were recently installed:
EW-475 o
E and EW-476.
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Several components were found in the incorrect position in the field.
Valve EW-0369A, ECW Pump 1A Supplementary Lubricating Water
' Isolation, was found open but1 should haveLbeen shut (Train A was not i
L in. service at that time). A spare breaker, Breaker No. 3 at i
!
Distribution Panel DPA-335, was found ON'but should have been 0FF.
Two powe'r supplies (Motor Control Center E1A4, Cubicle F1, and Distribution Panel DPA-135, Breaker 7) were found 0FF but should have L
been.0N. The: power supplies provided power to the Essential Chiller
,
l-Condenser 12A outlet valve 'and associated servoamplifier. Two power L
supplies (Distribution Panel -DPB-135, Breakers 7 and 8) were found i
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the Essential Chiller Condenser llB and 12B outlet valve
servoamplifiers. Althoughjthese components were incorrectly--
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positioned, a safety concern did not exist.. The power supplies to!
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the essential chiller condenser outlet valves had been disconnected
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' from the valves by.a temporary modification. the spare breaker was
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anot connected to any; component, and the open valve would have simply i
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supplemented the lubricating water flow to the:ECW Pump'~1A.
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. A: fire cabinet in Room 67E (essential chiller room).of.the mechanical L

b Lauxiliary building, Unit 1,"was found open. lAn inspection'of the.
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~~cabinet identified a thick layer of spilled oil in the bottom of'the cabinet. This condition was.' reported promptly to the Unit 1~~
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operations manager for resolution.
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Thecableguard'onnonsafety-relatedCableN1XM1ARX258.(connectedto
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E Chlorine Analyzer AE-6952) was pulled loose:from.an endpoint
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l connection and was exposing the enclosed insulated conductors.-
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AllLitems were reported to the licensee for resolution. The licensee-
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~~stated the operating procedure was in the process of revision end.the L~~
observations would be incorporated into the revised procedure. The-
' observations'made were not considered safety significant.and none.would E
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have3 prevented the system from~ perfonning its intended function.
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NoLviolations or deviations were identified in this area of the L
~~inspection.~~
~~11. -NonroutineLReporting Program- (90714)~~
~~An inspection-was conducted of specific portions of-the licensee's.~~
i
E nonroutine reporting program to ascertein whether responsibilities have

L been assigned for review and evaluation of-off normal operating events and-
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~~whether vendor bulletins and circulars were reviewed for applicability and L-incorporation where appropriate. This inspection also included the h~~
controls for recognition of applicable events that-are covered by p
10 CFR Part 21. The following documents were reviewed in the course of l:
this. inspection:
,
Interdepartmental Procedure IP-1.8Q, Revision 7. " Control of Vendor

Documents"
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InterdepartmentalProcedureIP-1.15Q, Revision 0,"STPEGSReporting
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Program?
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'Iriterdepartmental Procedura IP-1.03Q, Revision 1, " Reporting 10CFR21 e
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Deficiencies to the NRC"
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Interdepartmental Procedure IP-1.45Q, Revision 4, " Station Problem Reporting"
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' Interdepartmental Procedure IP-1,22Q,-Revision 5 Operating

Experience Review"
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The inspector determined that these procedures provided adequate controls
m for:the following:
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Receipt, review,L statusing, and distribution of. vendor supplied

design, technical, and QA documents applicable to this facility, q
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Identifying and evaluating conditions which could pos'sibly affect the
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safe: operation of this f acility.
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Reporting deficiencies, defects, and noncompliances to the NRC in
'
.accordance with 10 CFR Part 21.
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~~Ensuring that the reporting requirements applicable to the facility~~

-are effectively addressed.
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? Establishing uniform requirements for the management"and y
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administrative controls for correcting conditions that may not L
. conform to established requirements'.
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Establishing a uniform method for the screening and' review of l
industry and internally generated operating experience information.
'
a Tne inspector' performed an in-depth review of the licensee's Vendor
!
Equipment Technical Information" Program (VETIP) by: verifying that selected j
'
. bulletins'had'been incorporated into the following vendor manuals:
k Westinghouse Steam Turbine Manual - Volume II

~~Cooper Energy Services Emergency DieseldGenerator Service Manual~~

Additionally, the following bulletinscwere reviewed with respect to
_l timeliness and incorporation:
!
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Westinghouse Bulletin 87-01
!
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~~Westinghouse Bulletin 87-01, Revision 1~~

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~~Westinghouse Customer' Advisory Letter 89-03 l~~
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Foxbero Company Bulletin 89-01'
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~ Johnson Controls (ITT/Barton)Bulletin 86-01
'"
The inspector also reviewed licensee actions taken in response to a 10 CFR-Part 21 Notification' issued by_ Limitorque Corporation on November 3,1988. This notification involved reported failures of Melamine torque switches in Limitorque supplied SMB-00 and SMB-000 valve

actuators. The reported failures were determined to represent a common mode failure resulting from the postmold shrinkage of Melamine.
i Limitorque reported that SMB-000 actuators with serial numbers lower than 354839 and SMB-00 actuators with serial numbers lower than 233218 could have been supplied with Melamine torque switches. HL&P initiated a document search and determined that no SMB-00 actuators with serial
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. numbers lower than 233218 had been purchased. The search identified 13 SMB-000 actuators in Unit I and 11 SMB-000 actuators in-Unit 2 with serial-s <
numbers lower than 354839. HL&P1 considered this to be a potential environmental qualification (EQ) issue because. reference to time and
~~temperature' degradation 'of the Melamine switches was made in the notification. - Therefore, the suspect torque switches were further~~
'
reviewed to determine if any were utilized in EQ applications.- This
,
. review disclosed that none of the suspect torque switches were installed inside either containment buildings and that only one switch was installed n
in the isolation valve cubicle (IVC) (a harsh environment) of each unit. -
e'
Maintenance work orders were written (AF-61299 and AF-62197) to inspect
.
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MOVs DIAFHOV0514 and D2AFMOV0514 in the Units I and 2 IVCs. The results of these inspections were that Melamine was not used in the torque
,
switches for these two M0V actuators. The remaining suspect actuators were addressed by revising Procedure OPHP05-ZE-0300, "Limitorque M0V Motor.
Inspection and. Lube," to include inspections,for Melamine torque switches
~and replacement of any Melamine with Fiberite torque -switches. The remaining 22 actuators identified in the Part 21 notification were
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reviewed. Thirteen of the valves were butterfly valves which do not use a torque" switch, 4 were rad monitor system ball valves which do not use a torque switch, 2 were chilled water systems valves which had been deleted.
fromLthe system, and the remaining 3. valves were inspected as part of r
routine preventative maintenance and verified'as not having Melamine.
This procedure is applicable to all MOV actuatorsLduring their scheduled preventive maintenance and, therefore, represents a search program not
.
limited to the, actuators identified in the Part 21. notification. The licensee's decision.tn inspect all MOV actuators proved prudent in that, during the : application of this procedure,- one 'SMB-00 actuator, outside the
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scope of the Part'21 notification (Serial No. 339924),wasidentifiedas

L having a Melamine torque switch. The valve in. question was part of the L
chemical volume & control system (CVCS) and was not located in a harsh-L; environment. A problem report was issued'and the switch was replacad.
L Further investigation was inconclusive as to whether the valve actuator
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was originally provided with a Melamine torque switch by Limitorque. To
. preclude the use of Melamine torque switch usage, inspections were made of r
the spare parts and components located in both the HL&P and Bechtel warehouses, No other Melamine. torque switches were found. Additionally,
.the restricted components list (Specification SA23 HGS 0001) was revised to
. prohibit the purchase or use by any supplier of any Limitorque motor operators utilizing Melamine torque switches, t
. This inspection determined that the licensee's programs were effective and
.'
were being implemented in conformance with. requirements.
No violations or deviations were identified in this area of the inspection, l
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t 12. ' Plant Startup from Refueling (71711, 61705, and 61707)
j Following the; Unit I refueling outage and prior to unit startup, an=
inspection was performed to ascertain whether systems disturbed during the
. outage were returned to an' operable status and to determine if the plant
-
startup and core physics tests were conducted in accordance with approved j
~ procedures. The inspection consisted of observation of control room

Lactivities, plant walkdowns of-two systems worked on during the outage,
.and witnessing of; selected tests. Control room activities were inspected to ensure that the startup procedures and' plant personnel adhered.to TS-and procedural requirements.. The safety-related portion of the feedwater (FW) system and the entire ECW system were walked down.; The systems were inspected to ensure that they were returned to ' service in accordance.with plant procedures following disturbance during the outage.,
Additionally, the performance of1three surveillance tests and one.
temporary procedure was witnessedLto ensure that they were performed correctly and the requirements of TS were met.
'

The safety-related portion of the FW system was-inspected to verify that:
the system was-correctly lined up'to support plant operation. - The FW valve checklist, IPOP02-FW-0001-4, Revision 5, was compared to the system
'P& ids and to the as-found positions in the plant. ' All valves were noted to be in the correct positions.to support plant operation. Several
~ observations were made that were reported to the -licensee for resolution.
For example, Valve FW-0625 was-incorrectly labelled FN-0625_on.
P&ID SS139F00063-1, Revision 13,;"Feedwater System." The_ hydraulic skid for Isolation Valve FV-7143 had three meters with damaged faceplates.~
,
Valve 1-FW-0576 was found in the checklist for safety-related valves, but 1-FW-0576 was nonsafety-related. The'ECW system was also inspected in i
detail. The results of-the ECW inspection are discussed in Paragraph 10-of this inspection report.
,
.The procedures that were witnessed and reviewed by the inspector included (all. observations were reported to the_ licensee for resolution)i OPSP10-DM-0001, " Rod Drop Time Measurement,"' Revision 1

,
OPSP10-ZG-0003, " Shutdown Margin Verification - Modes 3, 4, and 5,"

Revision 4 L
"
OPSP03-NI-0001, " Daily Power Range NI (Nuclear Instruments) Channel

Calibration," Revision 2
OTEP07-MS-0002, " Main Steam Power Operated Relief Valve Inservice
'
L Operability Test," Revision 0 l
Procedure OPSP10-DM-0001 was performed to ensure that all shutdown and l
control rod drop time's were in accordance with TS (less than or equal to,
2.8 seconds). The licensee's biennial review of the procedure was performed late. The review should have been completed by June 23, 1989, but was
,
actually completed in October 1989. This was previously identified by the L
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.t fliin.ee and was documented in SPR 89-0610.--- Step 5.13 instructed the Joperator to open lift disconnect switches for rods not being tested.
There were no steps 1n the procedure to ensure that switches opened were
~
'reclosed. Two steps were missing from the procedure, requiring data
"
- to be recorded at Rod Position "0."
Step 5.28 refers the test performer-
'
to an incorrect addendum number. The note prior to Step 5.28.1.3-permitted;the use of a capacitor without specifying what size capacitor to use. -. The acceptance criteria, Step 6.1. quoted Section 3.1.3.4 of TS, but
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the step didinot' list all information provided in TS 3.1.3.4 (specifically,
,
the minimum temperature and number of reactor coolant pumps required to be
.
operable were not listed). The incomplete acceptance criteria had. no t
r effect on the final test'results.. A11' acceptance criteria reviewed by the inspector met TS requirements.

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The licensee performed Procedure OPSP10-ZG-0003 to determine the shutdown
.
margin and to verify the shutdown margin present was within the
,
.
requirements of TS. No specific concerns were identified with this
,
E procedure.
.
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t The licensee performed Procedure OPSP03-NI-0001-to determine, and readjust if necessary, the power. range nuclear instrument channels. This-test was
.
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performed to meet TS surveillance requirements.
Section 2.0, " Test
?
Equipment," included the use of a barometer, if available, to measure atmospheric pressure. Step'6.1.7.5 required the' barometric pressure to be taken,and recorded on the data sheet, or to use a= standard value of
,,
~14.7 psia if a barometer was not available. Two barometers were onsite
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and were available for use, but the standard value was'used instead. The L
procedure allowed the use of computer generated facsimiles to be L
substituted for the applicable forms contained in the procedure. The computer generated forms were similar to the ones in the procedure, but some of the columns for data recording were missing. Calculation sheet I
(-4) was missing the " Flow Nozzle Differential Pressure" column and the calculation sheet (-5) was missing the " Steam Generator Saturated Fluid Enthalpy" column. Neither of the missing columns had any effect on final ll results or acceptance criteria. Step 6.1.7.3 instructed the test performer
'
.to record feedwater temperature three times. The wording of the step did not clearly specify that four temperature readings were to be recorded three times (Steps 6.1.7.1 and 6.1.7.2 were more specific).
The licensee performed Procedure TEP07-MS-0002.to verify-that the main L
steam power operated relief valves (PORVs) would operate under full steam pressure and flow. This temporary procedure was writcen in response to a Justification for Continued Operation (JCO) concern on potential inadequate thrust to open the steam generator PORVs. Three'page numbers on the table
of contents page were incorrect. The procedure required stationing an operator outside the building housing the P0RVs to quickly respond in case a PORV failed to close. The test was performed with the operator inside the building, observing the valve. The note prior to Step 5.3 stated that-procedure Sections 5.6, 5.7, 5.8, and 5.9 could be performed in any order.
Section 5.9 was missing a step (allow plant to return to steady state
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conditions following'PORV testing)'that was present in the other three
'
in the other three sections.. The missing step suggested that Section 5.9
.should have been performed last.-
.
i No ' violations' or deviations were identified in this area'of the y!
,
h inspection.
13.
Exit Interview (30703)
<-
cThe inspectors met-with licensee representatives 3(denoted.in paragraph 1)-
' on November'1, :1989. The inspectors summarized the scope andl findings of-
,.
l the: inspection. The licensee did not identify as proprietary any of the
,
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Information provided to, or; reviewed by, the inspectors.'
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v t e Inspection Report - South Texas - 1989041
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1989 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 30(0) 31(0) 32(0) 33(0) 34(0) 35(0) 36(0) 37(0) 38(0) 39(0) 40(0) 41(0) 42(0) 43(0) 44(0) 45(0) 47(0) Security Operator Exams Emergency Planning Other

IR 05000498/1989041

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