ML20058M250

Insp Repts 50-445/90-22 & 50-446/90-22 on 900606-0703. Violations Noted.Major Areas Inspected:Plant Status, Operational Safety Verification,Startup Test Results Review, QA for Startup Test & Onsite Followup of Events

ML20058M250
Person / Time
Site:	Comanche Peak
Issue date:	08/03/1990
From:	Chamberlain D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20058M247	List: IR 05000445/1990022 ML20058M244 ML20058M248 ML20058M250
References
50-445-90-22, 50-446-90-22, NUDOCS 9008100026
Download: ML20058M250 (22)

See also: IR 05000445/1990022

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APPENDIX B-

U.S.- NUCLEAR REGULATORY COMMISSION

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' REGION IV

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NRC Inspection Report:

50-445/90-22

50-446/90-22

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Dockets:

50-445

Unit 1 Operating License:

NPF-87

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50-446

Unit 2 Construction Permit: CPPR-127-

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Expires:

August'1, 1992

Licensee:

TV Electric

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Skyway Tower

400 North Olive Street

Lock Box 81

Dallas, Texas 75201

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Facility Name: Comanche Peak Steam Electric Station (CPSES), Units 1'and 2

Inspection At

Glen Rose, Texas

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Inspection' Conducted:= June 6 through July 3, 1990

Inspectors:

W. D. Johnson, Senior Resident Inspector

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R.-M. Latta, Senior-Resident Inspector

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S. D. Bitter, Resident Inspector-

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H. F. Bundy, Reactor Inspector, Region IV

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D. N. Graves, Resident Inspector

A.' T. Howell, Resident Inspector

D.=L. Kelley, Reactor Inspector, Region IV

M. F. Runyan,. Resident Inspector

Approved by:

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D. D.//hamberlain, Chief,' Project Section B

Date

Division of Reactor Projects

Inspection Summary

Inspection Conducted May 3 through June 5,1990 (Report 50-445/90-22;

50-446/90-22)

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Areas Inspected:

Routine, unannounced inspection of plant status, operational

safety verification, startup test results review, quality assurance for the

startup test program, onsite followup of events, monthly maintenance

observation, monthly surveillance observation, startup test witnessing,

followup of licensee event reports (LERs), followup of previously identified

items, followup of NRC Bulletins, Information Notices, and Generic Letters, and

Unit 2 walkdowns.

9008100026 900503

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-Results: Within the areas inspected, strengths were identified in the areas of

performance of a reactor startup and conduct of the remote shutdown test.

Interdepartmental coordination, communications, and professionalism.were

excellent.

In addition, the startup test program was being accomplished in a

. professional manner by knowledgeable personnel, and radiation protection

support-of'the incore detector thimble tube cleaning effort was considered to

be very good.

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Areas of' concern included an incomplete technical evaluation dealing with the

potential impact of debris on the operability of the feedwater preheater bypass

valves and reliance on system flush data from 1982 to establish a basis for

feedwater system cleanliness.. In addition, the number of. incidents involving

clearances.and work control during the outage caused concern.

Three violations were identified. A slave relay test procedure was considered

to be inadequate (paragraph 6.b).

The second violation involved work on a

Unit 2 component without work order authorization (paragraph 7.i). .The third

violation involved an initial engineering evaluation of a flange misalignment

which was considered to be inadequate (paragraph 7 k).

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DETAILS

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persons Contacted

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• M._L. Bagale, Manager, Startup

• 0. Bhatty, Issue Interface Coordinator

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• M. R. Blevins, Manager of Nuclear Operations Support

• H.:0.-'Bruner, Senior Vice President

• C._B. Corbin, Licensing Engineer

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• G. L. Edgar, Attorney, Newman and Holtzinger .

• J. K. Ettien, Acting Unit 2 Operations Manager.

• J. L. French, Independent Advisory Group

• W. G. Guldemond, Manager of Site Licensing

• J. C. Hicks, Unit 2. Licensing Manager

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• J. J. Kelley, Plant. Manager

• F. W. Madden,' Mechanical Engineering Manager

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• D

M. McAfee,. Manager, Quality Assurance'(QA)

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'J. W ' Muf fett, Manager of Project Engineering

• E. F. Ottney, Project Manager, CASE

• S. S. Palmer, Stipulation Manager

• H.' S. Phillips, Consultant, CASE

• C, W. Rau,-Unit 2 Project Manager

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• A. B. Scott, Vice President, Nuclear Operations

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• M. D. Spence,-President-, Generating Division

• P.,B. Stevens, Manager of 0perations Support Engineering

• J. F. Streeter, Executive Assistant, Generating Division

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• C. L. Terry, Director of- Quality Assurance

• 0. W. Thero, CASE

• J. R. Waters, Site Licensing

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• Present at exit interview,

In addition to the above personnel, the inspectors held discussions with

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various operations, engineering, technical' support, maintenance, and

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administrative members of'the licensee's staff.

2.

' Plant Status - Unit 1-

(71707. 71715)

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The unit was at the 50' percent testing plateau at the start of this

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inspection period.

Following the remote -shutdown test on June 6,1990,

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the unit was cooled down to Mode 5 for a maintenance outage. Mode 1 was

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entered again on June 24, 1990, and power was subsequently raised to the

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75 percent testing plateau. The NRC resident inspectors provided

24-hour per-day site coverage during the power ascension. On July 2,

1990, power was reduced to 60 percent due to problems with a main

feedwater pump.

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3.

Operational Safety Verification (71707, 71715)

The objectives of this inspection were to ensure that this facility was

being operated safely and in conformynce with regulatory requirenients, to

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ensure thatithe. licensee's management' controls ~were effectively

discharging-the licensee's responsibilities for_. continued safe. operation,

to assure that selected activities of the licensee's radiological

protection programs are implemented in conformance with plant policies and

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procedures and-in compliance with regulatory requirements, and. to inspect-

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the licensee's compliance wi_th the approved physical security plan.

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.The inspectors conducted control room observations and-plant inspection

tours and reviewed logs and licensee documentation of equipment problems.

Through_in plant observations and attendance of the licensee's

plan-of-the-day meetings, the inspectors maintained cognizance over-plant-

status and Technical Specifications (TS) action statements in effect.

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a.

Work Control and Danger-Tag Clearance Problems

During a routine review of licensee operations notification and

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-evaluation (ONE) forms, the inspector noted that several ONE forms,

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which were originated during the scheduled: outage of' June 7-20,-1990,

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related to work-control and danger-tag clearance problems. These

problems-included:

failure to identify the' correct compone.nt to be

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repaired, failure to danger-tag the correct components for the

establishment of an adequate isolation boundary, failure to obtain

the proper authorizations prior to commencing work, failure to remove

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tags prior to removing tagged equipment,-failure to tag all the

necessary components for an adequate isolation boundary, and failure

to adequately determine the effect of clearance isolations on plant

equipment and systems.

The inspector noted that approximately one-half of-these ONE forms

were related to problems associated with the danger-tag clearance

process.

Several of-the ONE forms documented instances in which the

original clearance request required certain components to be tagged,

but the actual clearance (Clearance Report - Tag (s) to be Placed) did

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not reflect all the tags listed in the clearance request.

Discussions

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with licensee personnel revealed that, in some cases, the " Clearance

Report - Tag (s) to be Placed" provided an adequate clearance boundary

even though they differed from the clearance request', while in other

cases, errors were made in the generation of the Clearance Report -

Tag (s) to be Placed.

In most cases, however, where errors were made,

the errors were detected by craftsmen in accordance with Station

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Administrative Manual Procedure STA-605, Revision 8, " Clearance'and

Safety Tagging." The licensee formed a task group to review these

recent examples of work control problems.

It issued a report

detailing the root and contributing causes of these work

control / clearance problems as well as corrective actions.

The inspector reviewed this report and found it to be comprehensive;

however, work control and clearance problems were still occurring as

evidenced by additional ONE forms and the identification by an

inspector of licensee maintenance personnel performing maintenance on

plant equipment that differed from the equipment authorized by Work

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Order C90-3176 (paragraph 7.1).

The inspectors will' continue to

' followup on the_ licensee's implementation of corrective actions

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future inspections.

Paragraph 6.a. also discusses a problem with a

clearance. tag-out and inspection followup of this area will_be tracked <

as Inspector Followup Item (IFI) 445/9022-01.

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b.

Reactor Startup

On June 20, 1990, the inspector observed a Unit I reactor startup

conducted in accordance'with Procedure IP0-002A, Revision 6,'" Plant

Startup from Hot Shutdown to Minimum Load." The unit supervisor

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-briefed the-crew prior to the approach to criticality. . Operator

actions during_the evaluation were ci liberate and were carefully

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performed.

Communicattuns between the operators and.the unit

' supervisor were formal and. professional.

Reactor engineering

personnel were:present to perform inverse count rate ratio

monitoring. Overall crew performance during this startup was

excellent.

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c.

Containment' Sump Level Instrument

It was observed during a-control room tour that a limiting condition

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for operation (LCO) for containment sump level had been exited as a

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result of-. a technical evaluation of containment sump level-

Indicator 9-LI-5160. The indicator had been indicating erratically

and the action statement for TS 3.4.5.1 was entered on May 30, 1990.

Technical Evaluation (TE) SE-90-1744 stated that the faulty level

indicator was not part:of the sump level and flow rate monitoring-

system. required by TS 3.4.5.1.

The LC0 was terminated on June 3,

1990.

Discussions with 1.he system engineer and reviews of the TS.

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bases, surveillance requirements, and Regulatory Guide 1.45, " Reactor

Pressure Boundery Leakage Detection Systems," indicate _that the

faulty containment sump level indicator is not part of the equipment

required by the specification.

d.

Containment Inspection

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Following;the completion of the planned maintenance outage described

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in paragraph 2, the inspectors performed a walkdown of the Unit 1

containment building using procedure OPT-305, " Containment Closeout

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Inspection," as a guide.

In general, the observed housekeeping-and-

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material- controls were determined to be very good. However, the

following minor discrepancies were identified and were reported to

the shift supervisor.

The latch was broken on a ventilation duct inspection door near

Damper 02-VD-042 and the door would not close.

Scaffolding was erected in several locations.

Several components had boron buildup on them or other

indications of leakage.

These included reactor coolant

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system (RCS) Flow Transmitter Nos.1-FT-424, -435, and -444;

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and Valves 1-05-038, 1-WB-016, 1-LCV-460, 1-8000A and -B-

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(power-operatedreliefvalves[PORVs]blockvalves).=

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The regenerative heat exchanger room had small pieces of-

concrete on the floor beneath,the heat exchanger,

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The-open-ended pipe for the containment. particulate, iodine,'and

gas (PIO)' monitor air return (downstream of'1-RM-001) had a

plastic bag taped to'it to catch oil dripping from the open

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-pipe.

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Subsequent to the identification of these items,.the inspector

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determined-that the scaffolding which was not required or approved

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for placement inside containment during operation had been removed

and that the remaining items were addressed or had been previously

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addressed by the licensee,

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Main Steam Isolation Valves

The inspectors observed control room operations during the ascension

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to 75 percent power.

One' problem observed included Main Steam

Isolation Valve-(MSIV) Nos, I and 3 hydraulic pressure decreasing as

-a result of the hydraulic pump on the actuators not functioning

properly, 'If hydraulic pressure dropped below hemisphere (a static

nitrogen supply on the valve actuator) nitrogen pressure,'the valve

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could drift closed. As hydraulic pressure decreased, nitrogen

pressure was reduced to ensure that hydraulic pressure remained

higher.

Nitrogen pressure was not reduced-below 2100 psig, which was

well above the 1850 psig minimum required-for valve operability.

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The hydraulic pumps are air operated. The pump on the No. 1 MSIV was

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restarted by removing the exhaust muffler, covering the pump air

exhaust by hand, and then releasing the exhaust to provide a surge

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through the pump.

This appeared to restore the pump's operation.

The pump on the No. 3 MSIV was tapped with a rubber mallet and began

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to function properly. These actions were performed under Work

Order C90-3109.

The licensee planned to replace or repair the hydraulic pumps when

plant conditions and parts availability permit.

4.

Startup Test Results Review (72580, 72582, 72583, 72600, 72608)

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Results of selected startup tests were reviewed by the inspector to

determine that each had been properly conducted, reviewed, and approved by

the licensee.

Among the characteristics checked by the inspector were

entry of all required data, appropriate disposition and retesting for all

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. deficiencies, and' documentation of. appropriate approvals.

Test results

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reviewed were as follows:

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ISU-222A, Revision 4', " Turbine Generator Trip with Coincident Loss of -

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Offsite Power"

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ISVa020A, Revision 4 "Startup Adjustments of Reactor-Control Systems' -

(30-and 50 percent Plateaus)"

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ISV-231A, Revision 3, " Design Load Swing Test (30 percent Power

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ISV-231A, Revision 3, " Design Load Swing Tests (50 percent Power

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Execution)"

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EGT-788A, Revision 1, " Precision Secondary Side Power Calorimetric

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(30 percent Power Plateau)"

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ISV-023A, Revision 4, " Reactor Coolant Flow Measurement (50 percent'

Power)"

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EGT-709A, Revision 0, " Reactor Coolant System' Total Flow Rate _ Test"

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ISV-203, Revision 3, " Automatic Reactor Control System"

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ISV-223A, Revision 2, " Remote Shutdown Capability Test"

ISU-226A, Revision 3, " Operational Alignment of Process Temperature-

and N 16 Instrumentation" (50 percent power plateau)

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ISV-205A, Revision 4, '! Dynamic Steam Dump Control"

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ISV-202A,-Revision 3, " Calibration of Feedwater and Steam Flow

Instrumentation at Power"

ISU-204A, Revision 4, " Operational Alignment of Nuclear

Instrumentation" (30 percent and 50 percent power)

The completed test packages reflected satisfactory completion of the

associated tests, including disposition of test deficiencies and

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appropriate retesting. The inspector noted that several core exit

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temperature data sheets beginning at 9:27 a.m. (CDT) on May 22, 1990,

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included in the results package for ISV-222A, were signed by a person

whose-name was not included on the signature sheet in accordance with

administrative requirements.

The test coordinator subsequently had this

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name added and signed on the signature sheet.

The inspector concluded that the startup testing program was progressing

smoothly.

The startup program control and performance was professionally

accomplished. The test personnel appeared conscientious and knowledgeable.

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Quality A'ssurance (QA) for the Startup (SV) Test Program (35501)-

As a followup to NRC Inspection Report 50-445/89-67;.50-446/89-67, the

inspector reviewed the'following audit and. surveillance' reports:

-QAA-90-002, " Initial Startup (ISU) Program," performed February 5-16,

.1990

QAA-90-005, "TS 3/4.9,8.1 and 3/4.9.8.2 - RHR and Coolant Circulation

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High and Lnw-Water Levels," performed March 8-13, 1990

QAA-90-009, "TS 3/4.4.8.3 - Overpressure Protection Systems,"

performed March 13-23, 1990

QAA-90-013, " Shutdown Margin --Tave 5 200 F," performed April 4-24,.

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QAS-90-454, " Unit 1 Initial Startup Activities," performed

. March 12-18, 1990

QAS-90-472, "ISU Procedure QA Hold Points, Testing in Accordance with

Nuclear Engineering and Operations Procedures, and Plant Operations,"

performed April 2-8, 1990

QAS-90-475, "ISU Procedure QA Hold Points, Testing in Accordance with

Testing and Operatiens Procedures, and Plant Operations," performed

April 9-15, 1990

QAS-90-480, "ISU-Procedure QA Hold Points, Testing in Accordance with

Testing and Operations Procedures, and Plant Operations," performed

-April 16-22, 1990

QAS-90-482, "ISU Procedure QA Hold Points, Testing in Accordance with

Testing and Operations Procedures, and Plant Operations," performed

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April 23-29, 1990

QAS-90-485, "ISU Procedure and QA Hold Points and Plant Operations,"

performed April 30 through May 6, 1990

QAS-90-486, " Evaluation of Performance of ISU Activities," performed

May 7-13, 1990

QAS-90-492, " Unit 1 ISU Testing and Operations as Performed by

Nuclear Operations personnel," performed May 14-20, 1990

QAS-90-495, "ISU Procedure QA Hold Points, Testing in Accordance with

ISU and Performance and Test Procedures, Operations Shift Turnover

and Plan-of-the-Day Meetings, and Plant Operations," performed

May 21-27, 1990

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. The; inspector verified that appropriate corrective actions had been-

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initiated for deficiencies identified during the'.above audits and

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surveillances. Also,. requested responses and dispositions by QA had been

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reviewed. The activities examined during the above audits, and

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surveillances' reflected the principal attributes of the startup. test

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program.

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The-inspector also witnessed portions of QA Surveillance (QAS)-90-501

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which involved the performance of Test ISU-223A,' " Remote Shutdown

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Capability: Test." Three QA. specialists participated in this surveillance

to comprehensively cover significant test activities.

Through interviews,

the-inspector determined that all three QA specialists were knowledgeable

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-of1the> test objectives and methods as well as QA requirements.

QA hold

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' points-were appropriately witnessed and' signed off. The inspector

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attended a posttest debriefing with the QA specialists to ascertain their

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approach to deficiency' followup. The test deficiencies observed were

minor and did not impact acceptance criteria. . The QA specialists

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appropriately coordinated followup actions with engineering and operations.

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. personnel. Apparent deficiencies discussed were as.follows:

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An apparent labeling discrepancy for one' transfer switch;

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Attempts by an operator to trip a breaker by-depressing.a push button

which would only function in the test position; and

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Availability of step stools or ladders to assist operators in

operating upper tier electrical breakers.

The overall performance of the test by operations and' performance and test

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personnel was discussed and no further problems were identified. The

above noted deficiencies did not affect the acceptability of the test.

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Onsite Event Followup (93702)

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Inadvertent Transfer of Reactor Coolant to the Refueling

Water Storage Tank

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On June 12, 1990, at 8:13 p.m. (CDT), with the plant operating in

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Mode 5 (reactor coolant system pressure approximately 350 psig, and

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temperature approximately 140 F), the pressurizer high level

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annunciator in the control room cleared. An operator in the control

room determined that reactor coolant inventory was decreasing as

evidenced by decreasing pressurizer level. At the time, the licensee

was conducting a leakage test of RCS pressure boundary isolation

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valves in accordance with Testing Manual Procedure EGT-712A,

Revision 5, " Reactor Coolant System Pressure Boundary Isolation Valve

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Leakage Testing." The operator stopped the decreasing pressurizer

' level by closing air-operated safety injection system test header

Isolation Valve 1-8890A.

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TheLlicenseel determined that a direct flow' path from the RCS through

the residua 1Lheat removal-(RHR)' system to the refueling water: storage

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tank (RWST) was created:when RHR System Isolation Valve 1-8809A

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'(RHR'to CL-1/2 INJ ISOL) wasiopened with Valve 1-8890A: subsequently-

opened per Procedure EGT-712A. This' allowed approximately

260 gallons'of reactor coolant to drain to the RWST before

-Valve;1-8890A was closed by an operator.

RHR Valve 1-8809A was

required;to be closed during.the test, but~it.had been'previously

opened to-its normal position following'the release of Danger-Tag

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Clearance 1-90-1221.

Discussions with licensee personnel and a review of licensee records

revealed that1 Valve 1-8809A was checked closed in accordance with-

.EGT-712, Attachment 26, prior to the conduct of the test; however.

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Valve 1-8809A was still danger-tagged shut as required by

Clearance ~ 1-90-1221 when it was checked.

Shortly after 5 p.m. on

June 12, 1990, Clearance 1-90-1221 was released and Valve 1-8809A was

-restored.to its normally open position.

Discussions with licensee

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personnel revealed that operations department personnelLhad asked the'

shift' test director if. releasing Clearance 1-90-1221 would affect.the

test lineup and the operators were informed that' removal of

Clearance 1-90-1221 would probably not affect the test lineup.

Subsequently, the licensee began leak testing of RCS cold leg

injection outer Isolation Check Valve 1-8818A which requires-

Valve 1-8809A to be closed, but Valve 1-8809A had been repositioned

. opened after Clearance 1-90-1221 was released. As a result, when.the

licensee opened safety injection test header Isolation Valve 1-8890A,

in accordance with Procedure EGT-712A (Step 11.7.8), a direct flow-

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path 1from the RCS to the RWST was created.

The inspector reviewed Procedure EGT-712A and determined that there

were adequate provisions for verifying the test lineup and reviewing

the clearance report index prior to commencing leak testing.

However, problems in effective communications between operations and

performance and test personnel appeared to have' contributed to the

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event. -As of the end of this inspection period, the licensee was

still evaluating the cause and-corrective actions for this event.

The inspectors will followup on the licensee's actions af ter this

incident has been resolved by the licensee in accordance with Station

Administrative Procedure STA-422, Attachment 8.B.

The inspector

will continue to monitor licensee action in the clearance tag-out

area as IFI 445/9022-01 as noted in paragraph 3.a.

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Inadvertent Actuation of Auxiliary Feedwater ( AFW)

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At 3:16 a.m. on June 13, 1989, with Unit 1 in Mode 5, an

inadvertent actuation of Train A of AFW (motor-driven pump only)

occurred when control room personnel placed the mode selector switch

on the Train A solid state protection system (SSPS) output cabinet to

the normal position during the performance of Procedure OPT-467A,

" Train A Safeguards Slave Relay K609 Actuation Test."

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Prior to the event, the plant ~was in Mode'5'with the Train _A Mode 5/6'

switch (S604) in the Mode 5/6 position-and the Train ~A mode: selector

switch in:the test position.

These switch positions bypassed all

- normal solid state protection system automatic actuations with the

' exception:of_ the flux doubling circuitry and the containment .

ventilation isolation circuitry. When_the' operator performed

- Step 9.13 of OPT _-467A, which should have caused a number of

. components,to actuate, none of the component actuations occurred.

The unit supervisor determined the reason was that the Mode 5/6=

switch was.in the Mode 5/6 position and-proceeded to

Procedure 50P-711A, " Solid. State Protection System," to restore the

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SSPS to a normal switch lineup so-the proper . system response could be

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obtained during the performance of OPT-467A. Upon placing the-mode

selector switch in the normal position (Step 5.3.6 of.

Procedure SOP-711A), the Train A motor-driven AFW pump started on-a

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low-low > steam generator level because Steam Generators 2 and 3 were

drained. The AFW pump receives a start. signal on low-low level in-

-one steam generator.

The operator performing the test did not ensure that the motor-driven:

AFW pumps were disabled even though low levels existed in.two steam

generators. The. performance of Step 5.3.5 of SOP-711A, which is'a

determination of trip signals present that would actuate alarm

relays, and the caution following Step 5.3.5'regarding engineered

safety features.(ESF) function actuations, should have alerted the

operator to the fact that an ESF actuation would occur.

The NRC Headquarters Operations Center duty officer was notified at

5:42 a.m.

A ONE form was generated (No. 90-1717) and sent to

operations for disposition.

OPT-467A stated'that the procedure may be performed in any mode and

did not address the need for either.the mode selector switch or S604

to be in any specific position prior to starting the test. With the

initial switch lineup present during this event, the surveillance

could not be performed. OPT-467A did not provide guidance as to

which procedure or section should be used in order to ensure that the

system was in the proper switch lineup.

The inadvertent ESF actuation and subsequent licensee corrective

actions will be the subject of further inspection followup pending

the issuance of LER 445/90-018.

Failure to provide sufficient guidance in OPT-467A to permit

performance of the test and to prevent emergency safeguards

actuations during procedure performance is a violation of

TS 6.8.la (445/9022-02).

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Inadvertent Actuation of a Fire Protection Deluge Value

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On June 4,1990, at 5:36 p.m. a fuel' building railroad bay fire

protection deluge valve was actuated.- Initial indication and

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licensee. investigation revealed that the deluge valve had been

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inadvertently manually initiated.

The licensee took appropriate

corrective action as a result of the findings of.the investigation.

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. 7.

Monthly Maintenance Observation (62703)

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Station maintenance activities for the safety-related and nonsafety

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systems and components listed-below were observed to_ ascertain that they

were =cenducted in accordance with approved' procedures, regulatory guides,

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industry codes or standards,'and in conformance with the TS.

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a.

The inspector witnessed several aspects of the incore detector

thimble. cleaning activity (Work Order C90-4020). Observed portions

iincluded the final disconnection and removal-of the seal table,

disconnecting the thimble- tube vacuum drying assembly'from one row of.

thimbles and reconnecting to another row of thimbles, and testing of

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two rows of thimbles with the use of a dummy detector cable inserted

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the-full length of the thimble.

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~ Radiological requirements were adhered to by all personnel involved

in the task.

The licensee maintained the seal table room as a

contaminated area for the duration of the task even:though very

little contamination was observed and was limited to the top of the-

seal table. The area was maintained very clean with personnel

entering the area to perform decontamination activities several times

eduring the : task. -Respiratory protection was required by the licensee

during the cleaning until it'was determined that no airborne

contamination was being generated.

Respiratory protection was also

required _during the. testing portion of the procedure. General

radiation protection practices were good.

Samples of the residue present in the thimble tubes were collected as

the cleaning tubes were withdrawn from the thimbles. The filters

used for processing the cleaning water effluent from the thimble

tubes were also retained. The licensee has the preliminary results

of an isotopic analysis of the residue performed onsite.

It was the

intent of the licensee to have the sample and filter residue

chemically analyzed by an offsite laboratory in an attempt to

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determine the source of the thimble tube blockages.

The procedure used for cleaning the thimbles was written by the

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contractor used for the task.

The procedure was reviewed and

approved as a nuclear engineering (NUC) procedure and incorporated

into the work order.

It appeared to be adequate.

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The inspector' observed the repair of a feedwater isolation

valve (1-HV-2136)-as a result of a body-to-bonnet leak (Revisions 7

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and 8 of. Work Order C90-3318).

]

All observed activities were in accordance with approved procedures,

including limit switch disassembly and removal, use of calibrated

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torque wrenches, determination of correct torque values, application

-of appropriate-torque values, and proper quality control (QC)

coverage on.the required-steps.

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During tightening of the packing gland fo' lower, the inspector-

observed that the orientation of the folloster ' racket was different

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from its position when. initially loosened.

The inspector asked'the

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. mechanics if the orientation was significant and they responded that

it was not.

Upon attaching the limit switch assembly, it was

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determined that one of the shut position limit switches came in

contact with one of the bolts on the follower bracket when the switch

was in'the full down position.

The mechanics obtained the proper

signature sheets to loosen, reposition the packing gland follower,

and retorque the fasteners.

Again, this was a complished in

accordance with the appropriate procedure

c.

The inspector observed the video probe inspection of the piping

upstream and downstream of the No. 3 steam generator feedwater

isolation bypass valve (Work Order C90-4171).

d.

The inspector witnessed a backflow. leak test of the 4-inch

Borg-Warner check valve used to replace Valve 1-AF-101, installation

of the valve internals into the body in the AFW system, and the-

subsequent backflow leak test with the valve installed in the system

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(Work Order C90-4153).

e.

The inspector witnessed the postwork testing for a design

modification (DM 90-183) which changed the pressurizer spray line low

temperature setpoint. The test was performed using Temporary

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Procedure PPT-PT-90A-20 (Work Order C90-4202),

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f.

The inspector witnessed the recharging of MSIV-2334 hemisphere (Work

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Order C90-3109).

g.

The inspector witnessed the performance of the postimplementation

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testing for Design Modification 90-194 (PPT-TP-90C-001) on Radiation

Monitor X-RE-5896A.

h.

A review of Work Order C90-3823 indicated that Counter Card No. 405

in the logic cabinet of the rod control system was faulty. A failure

of this card would provide the same indications that were present in

previous rsd control system failures. The licensee now believes that

the two cards that were replaced (A2 and J2 in the IBD power cabinet)

during one previous rod control system problem only caused the

previous alarm condition to clear when the cards were pulled and

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either reinserted.or replaced. The' removed cards were onsite

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awaiting disposition.

The licensee was evaluating =the possibility of

card failure as a~ result of environmental conditions in the area ofJ

the rod control cabinets. .This had been previously identified as

IFI 445/9019-01.

The. inspectors will continue to monitor licensee-

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' actions in-this area.

1.

-On June 26, 1990, the inspector observed mechanical maintenance

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personnel opening the Unit 2 reactor coolant pump seal water return'

filter. -This was being performed under Work Order.C90-3176 as a1

training evolution in preparation for changing the corresponding

- f.ilter on Unit 1.

The inspector identified that the workt order-

authorized changing the Unit 2 reactor coolant filter'and that.the

first step (verification that the component being worked on was the

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one authorized) had not been completed. When the working foreman was

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notified, he had the seal water return filter canister closed, and he

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stopped in job.- ONE Form FX 90-1798 was initiated. ' The reactor

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coolant

'ter had been tagged out under Clearance 2-90-0069,.but the

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seal retur$ filter had not.been placed under a clearance,

While performance of a training evolution on Unit 2' components-was a-

beneficial initiative, working on a component not authorized by the

work order is an apparent violation of Section 6.6.2.9 of

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-Procedure STA-606, Revision.14, " Work Requests and Work

Orders" (446/9022-01),

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The inspector witnessed the attempt-to stop a body-to-bonnet leak.on

1-HV-2492B, the motor-driven AFW pump discharge isolation valve to

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.No. 2 steam generator (Work Order C90-4373) by increasing the torque-

applied to the body-to-bonnet fasteners. This attempt was not

successful.

k.

During this reporting period, a short duration planned maintenance

outage occurred, during which several significant maintenance related

activities took place. One of these activities involved the

modification of eight Borg-Warner, 4-inch pressure seal, swing check

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valves, which are located on the discharge lines from the AFW pumps

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to the steam generators. This modification, as described in Design

Change Notice (OCN)'1196, Revision 0, involved the addition of a

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counter weight to the: stem / disc assembly in order to enhance the

operational characteristics of these check valves under low

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differential pressure conditions which would assist in the prevention

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of temperature transients between the hedwater and the AFW systems

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as previously identified in NRC

v Report 50-445/90-13;

50-446/90-13.

In addition to the installation

unterweights, these repair

activities for Check Valves 1AF

1, -083, -086, -093, -098,

and -106 included the replacemen'

swing arms with Investment

Cast A747 CB7C4-1 components and the i.glacement of the disc / stud

assemblies.

The swing arm and disc / stud assembly for Valve 1AF-101

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was'previously replaced as documented in NRC Inspection-

Report 50-445/90-19; 50-446/90_-19.

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The inspector witnessed all- aspects of. these repair activities.which

were conducted in accordance with authorizing work orders for:each of

the.eight valves and Maintenan:e Procedure MSM-CO-8801. These -

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observed activities included the replacement _ of the valve body for

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Valve-1AF-101.which was-identified as having a leak between the seat

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ring :and -the associated seal. weld to the. valve . body.

In general, the

observed maintenance activities associated with the AFW check valves

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were acceptable and the operational characteristics of these-

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components were obssrved to be improved during the subsequent plant

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startup. Hewever, a flange' misalignment condition which was

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documented by mecha..ical' maintenance personnel on ONE Form FX90-110

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during the disassembly of the flow orifice immediately upstream of

Check Valve 1AF-093, was incorrectly dispositioned by project

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engineering to use-as-1s without adequate analytical justif.ication.

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This incorrect' disposition resulted from the failure of engineering-

personnel to perform onsite inspections of the discrepant condition

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and their reliance on the reported " approximate" misalignment value

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of 1/4 . inch as a bounding upper limit in their response.

In fact,

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the flange misalignment was indetermir, ate.because the loosened flange

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fasteners left 'in the joint restrained the two sections of. pipe.

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Subsequent to the identification of this condition by the inspector,

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the flange fasteners were removed and accurate measurements of-the

flange misalignment and lack of flange face parallelism were recorded

on ONE Form FX90-1715. These' values were reevaluated by project

engineering and the resultant stresses were determined to be below

the allowables. However, the licensee's failure to initially. perform

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an adequate engineering evaluation to address this deviation from the

installation specification (2323 MS-100) and Procedure MSM-CO-0203 is

identified as a violation of Criterion XVI, Appendix B,

10-CFR 50 (445/9022-03).

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It is noted that, subsequent to the identification of this particular

issue, project engineering management acted quickly to resolve the

technical aspects of this item by performing a rigorous computer

analysis of the piping geometry; however, a potential programmatic

concern exists with the timeliness and adequacy of engineering

evaluations performed to support operational conditions as previously

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identified ir NRC Inspection Report 50-445/90-19; 50-446/90-19.

1.

The inspector evaluated the licensee's corrective maintenance

activities associated with the the feedwater (FW) preheater bypass

valves which were disassembled in order to correct indicated seat

leakage.

The observed maintenance work activities were well controlled and

executed. During the conduct of this repair work, a significant

amount of debris was identified on the upstream side of all four FW

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preheater bypass valves.

The potential impact of this debris was

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addressed in the TE associated with One' Form 90-1703; Although this:

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TE appeared to provide an adequate analysis as- to'the potential-

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impact of the' debris on the steam generators, the impact of:the-

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materialcon'the operability of the valves-was absent, which is

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regarded:as a: weakness in the TE. iAt the conclusion of this

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'aporting period, the' licensee had not completed their evaluation of-

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this-issue; therefore, the results of this evaluation will be

reviewed and. documented in a future inspection report.

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A second issue associated with the debris which was discovered in the

FW preheater bypass valves involved the licensee's reliance on 1982

system flush data toLestablish a basis for system cleanliness in the

FW system. This questionable reliance on flush data, which was

8 years old, ignored the potential introduction of foreign material'

during the retubing of the Unit 1 main condenser and the moisture

separator reheaters which occurred subsequent to the' referenced

. system flush activities. Accordingly, this matter is. considered a

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weakness in the licensee's preoperational. test program for Unit 1.

8.

Monthly Surveillance Observation -(61726)

The inspectors observed the surveillance testing of. safety-related systems

and components. list'ed below to verify that the activities were being

performed in accordance with the TS. The applicable procedures were

reviewed for adequacy, test instrumentation was verified to be in

calibration, and test data was reviewed for accuracy and completeness.

The inspectors ascertained that any deficiencies' identified were properly

reviewed and resolved.

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The. inspector witnessed portions of the following surveillance test

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activities:

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PPT-TP-90A-024, Revision 0, " Main Feedwater Isolation Valve

Accumulator Pressure Test"

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The purpose of this test was.to obtain stroke time data for the feedwater

isolation valves (FWIVs) under various nitrogen accumulator pressures in

order to determine FWIV. operability. All four FWIVs were stroked at

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nitrogen accumulator pressures of approximately 2550 psig, 2250 psig

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-(main control board annunciator setpoint), and 2040 psig (the setpoint

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which Alarm Response Procedure ALM-0081A, Revision 2, " Alarm

Procedure 1-ALB-8A," requires that the FWIVs be declared inoperable). The

FWIVs were stroked under zero flow and differential pressure conditions.

4

In each case, all four FWIVs stroked closed in less that 5 seconds, which

is the maximum specified time required by TS Survelliance

Requirement 4.7.1.6.

Power Operated Relief Valves

During a review of the activities associated with the troubleshooting of

leaking primary Power Operated Relief Valves (PORVs) 1-PVC-455A

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and 1-PVC-456, the inspector learned of procedural problems encountered by

instrumentation and control (I&C) personnel during the performance of

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analog channel operational tests (ACOTs) on the valves.

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While preparing to perform INC-7757A, Revision- 5, the ACOT for instrument

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-channels associated with PORV 456, the technicians noticed'that Step 10.9

'specified the pulling of fuses that'were associated with PORV 456..

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Additionally, while the technicians were-performing INC-7758A, Revision 5,-

tne ACOT for instrument channels associated with PORV 455A they noticed

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'that Step 11.2.1.4 called for annunciator Window'2.4 on ALB-5B to be lit.

This condition could never be' met with the procedure as' written.

For bot; cases, the-errors were corrected by using procedure change:

notices. .Although these ACOTs had never been performed before while in

Mode-3, the errors-indicate a need for improved procedure review prior to

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approval.

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Liquid Waste Effluent Monitor

The inspector witnessed the performance of.a digital channel operational

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test on the liquid waste effluent process radiation monitor-(INC-7081X).

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Step 11.1.1 of the procedure had the technician set the process flowrate-

low alarm to 0.0 gpm,

Step 41.1.2 required the technician to set the

' sample flowrate low alarm to 0.0 gpm.

Step 12.2 required the technician

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to enter the values from the digital radiation monitoring system (DRMS)

data base (353-1000) as the' final setpoints for the process and sample

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flowrate alarms. The technician entered the as-found values that were the.

setpoints without verifying that the origina1' values were the same as

those in'the DRMS data base. The inspector compared the as-found values

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used by the technician to the DRMS data base values and they were

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identical. The I&C manager was informed of this-observation. One

= procedure revision was necessary to complete the surveillance. The

observed procedure .and two similar procedures were subsequently revised-to

clearly state that the as-found setpoint values are to be used as the

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final setpoint values.

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Motor-Driven Auxiliary Feedwater Pump

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The inspector witnessed the performance of an operability test of the

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Train A motor-driven _ auxiliary feedwater (MDAFW)' pump i 'PT-206A) on

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June 22, 1990.

Step 9.1.12 (recording of MDAFW pump suction pressure)

required a value of >18 psig. The actual value was 15.75 psig.

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technical evaluation (SE 90-1883) was performed as justification for a

revision to the. procedure to change the pump suction pressure acceptance

criteria from >18 psig to 14 psig. The procedure was changed and the

surveillance was completed and determined to be satisfactory based on the

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14 psig acceptance criteria. The licensee determined on June 25, 1990,

that an error was made in the technical evaluation and the actual

acceptance criteria should have been approximately 17 psig instead of

14 psig.

Subsequently, on June 26, 1990, it was determined that an

acceptance criteria for pump suction pressure was not needed because it

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was not required for pump operability in accordance with the CPSES

in-service testing (IST) program.

Pump differential pressure is the

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appropriate acceptance criteria for IST requirements. A procedure change

was generated to revise OPT-206A to reflect the deletion of the acceptance

criteria for suction pressure. The operations manager was notified on

June 27, 1990, of the error in the original technical evaluation.

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Slave Relay

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The inspector witnessed the performance of Train A Safeguards Slave

Relay K601 actuation test (OPT-463A),

Turbine Driven Auxiliary Feedwater Pump

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The inspector observed the running of this pump in the pump room during

performance of its cold start surveillance in accordance with

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Procedure OPT-206A, Revision 4. " Auxiliary Feedwater System Operability

Test." Environmental conditions in the room were acceptable during tMs

test. This indicated that the licensee's efforts to improve the steam

line drains have been effective.

Although several problems were identified with surveillance procedures by

licensee nersonnel, actions taken were appropriate to stop the tests and

correct the procedure deficiencies. Overall performance in this area.was

considered good.

9.

Str.: tup Test Witnessing (72300. 72302,_72583)

The inspectors witnessed selected startup tests in ceder to verify

conformance to testing commitments and procedural requirements, observe

staff performance, and verify that adequate test pNgram records were

mn;.a'ned.

The following items were considered during test witnessing:

Availability of current revision of test procedure,

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Minimum crew requirementt,

Test prerequisites and initial conditions,

Calibration status of test equipment,

Technical adequacy of test procedure,

Test coordination and crew performance,

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Preliminary results satisfactory or deviations documented for further

evaluation, and

Adherence to TS during testing.

In additien, the inspectors reviewed various logs and reports and attended

meetings and crew briefings related to the test program.

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During this inspection period, the following startup test was observed:

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ISU-223A, Revision 2, " Remote Shutdown Capability Tr st"

The licensee performed a remote shutdown of Unit 1 from outside ^?e

control room on June 7, 1990. The purpose of the tests were to oemonstrate

the capability to safely shut down the plant from outside the control room

from an initial power level of 10-25 percent, demonstrate the capability

to maintain the plant in a hot standby condition from outside the control

room, and demonstrate the capability for conducting an RCS cooldown from a

. hot standby condition by at least 50'F from outside the control room.

Seven resident and region-based inspectors witnessed the test from the

control room, the remote shutdown panel (RSP), the shutdown transfer

panel (STP), and var'.ous other locations in the plant.

Prior to the test,

the inspectors attended licensee planning meetings as well as the

test briefing.that was conducted just prior to the test.

Test

Procedure 15U-223A and Abnormal Conditions Procedure ABN-905A, Revision 3,

" Loss of Control Room Habitability," were previously reviewed by the

inspectors. The results of this review were documented in NRC Inspection

Report 50-445/90-13; 50-446/90-13. The licensee adequately resolved the

weaknesses associated with these procedures prior to conducting this test.

Followup review of these procedures is documented in NRC Inspection

Report 50-445/90-19; 50-446/90-19.

The test was well executed and no significant problems were noted by the

inspectors. During the performance of ISU-223A, the inspector cSserved

the actions of an auxiliary operator while aligning the charging pump

suction path to the RWST, isolating RCS dilution paths, transferring

control of the steam generator atmospheric relief valves to the remote

shutdown panel, and verifying the operation of the motor-driven auxilia;

feedwater pumps.

Communications and procedural compliance were evaluateo

as excellent by the inspectors.

Some minor equipment problems were noted.

The test was successfully completed at approximately 1:04 p.m. after

the plant was cooled down to approximately 477*F.

Upon transferring

control of equipment back to the control room, the inspectors noted that

some control room operetors did not Apoear to be fully aware of all the

equipment that had been transferred to 1.he RSP. However, the operator at

the RSP was cognizant of the remaining plant quipment that had to be

transferred back to control roon control, ano the transfer was completed

with no problems noted.

10.

Followup of Written Reports of Nonroutine Eveits (90712)

The following LERs were reviewed and closed.

Tho inspectors verified that

reporting requirements had been met, causes had been identified,

corrective actions appeared appropriate, generic applicability had been

considered, and the LER forms were complete.

The inspectors confirmed

that unreviewed safety questions and violations of TS, license conditions,

or other regulatory requirements had been adequately described.

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(Closed) LER 445/90-007-00, " Personnel Errer and Procedural

Inadequacies Leading to Inadvertent Actuation and Subsequent

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. Disabling of Control Room Air-Conditioning System Engineering Safety

Feature."

b.

(Closed) LER 445/90-008-00, " Train A Diesel Generator Inoperable."

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c.

(Closed) LER 445/90-011-00, " Containment Isolation Root Valve Left in

Incorrect position as a Result of Procedural Deficiency."

d.

'(Closed) LER 445/90-012-00, " Time Limits of Technical Specification

Action Statement Exceeded Due to Personnel Error."

e.

(Closed) LER 445/90-015-00, " Missed Chemistry Sample Special

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Condition Surveillance Due to Procedural Deficiency."

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11.

Followup of Previously Identified Items (92701)

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(Closed) Open Item (445/9009-01):

Turbine-driven auxiliary feedwater pump

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room fills with steam when pump runs. The licensee has modified the steam

line drains in this room.

This modification appeared to be successful

during observation of a pump run observed during this inspection period.

.

This item is closed.

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12.

Followup of Bulletins. Information Notices, and Generic Letters (92701)

The purpose of this inspection was to verify, on a sampling basis, the

effectiveness of the licensee's program for handling NRC Bulletins,

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Information Notices, and Generic Letters.

In particular, records were

examined to determine whether these documents were reviewed for

applicability, received prot r distribution to the appropriate personnel,

!

and resulted in the performance of appropriate actions.

!

The inspector interviewed the principal site coordinator for NRC

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Bulletins.

It was noted that action was underway by the licensee to

revise the procedure, Texas Utilities Nuclear Licensing (TNL) 4.11,

" Evaluation of NRC Bulletins," Revision 0, to clarify the manner in which NRC

Bulletins are handled and documentation requirements.

The inspector

reviewed the site files for the completed responses to the following NRC-

Bulletins: 88-10, 89-01, and 89-03. The files reflected a comprehensive

review of each issue and the concurrence / approval of the appropriate

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disciplines. The licensee's handling of NRC Bulletins appears adequate.

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The inspector interviewed the principal site coordinator for NRC

Information Notices and reviewed the governing procedure for this

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activity, STA-507, Revision 1, " Review and Assessment of Industry

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Operating Experiences." The inspector reviewed site files perteining to

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the following Information Notices:

85-17, 86-49, 87-25, 88-40, 89-33,

and 90-08. On the basis of this inspection activity, the inspector

concluded that the licensee's program for addressing NRC Information

Notices was adequate.

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The inspector interviewed by telephone the principal licensee coordinator

(in Dallas) for NRC Generic Letters and reviewed the governing

administrative procedure, TNL-4.01-1, Revision 3 " Incoming

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Correspondence." The inspector reviewed licensee documentation in

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response-to NRC Generic Letters 89-19 and 89-21. On the basis of this

inspection effort the inspector concluded that the licensee's handling of

generic letters is adequate.

13. Unit 2 Walkdowns (70302, 71302)

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During this inspection period, routine tours of the Unit 2 facility were

conducted in order to assess equipment conditions, security, and adherence

to regulatory requirements.

In particular, plant areas were examined for

evidence sf fire hazards and installed instrumentation damage and to

determine the acceptability of system cleanliness controls and general

housekeeping. Additionally, the inspector conducted evaluations of

existing plant programs for the preservation and maintenance of installed

systems and components as well as the utility's preparations for the

resumption of construction activities for Unit 2.

In part1cular, the inspector reviewed selected aspects of the Unit 2

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implementation of lessons learned from the Unit I corrective action

program (CAP) and postconstructuo hardware validation program (PCHVP) as

delineated in the following TV Electric documents:

TV Electric '.etter TXX-88373 dated April 14, 1988;

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TV Electric Letter TXX-89271 dated May 19, 1989; and

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Procedure 2EP-1.04, " Evaluating Unit 1 Post-Construction Hardware

Validation Program Results for Applicability to Unit 2."

As a result of these inspection activities, it was determined that the

licensee is currently reviewing Unit 1 commitments to assure that these

commitments will be adequately addressed in the Unit 2 program.

,

Additionally, it was ascertained that the Unit I des gn basis documents

are being reviewed by the licensee and, where applicable, are being

revised to reflect Unit 2 specific information.

Hardware validations are currently being conducted by commodity and

programmatic corrective actions are being addressed in the licensee's

integrated construction / inspection procedures which are currer tly under

development for Unit 2.

It is also noted that a public meeting regarding

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the licensee's proposed Unit 2 project: completion plans is currently

scheduled for July 17, 1990.. in the NRC's Region IV office located in

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14.

Exit Meetina 30703)

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- The inspection scope and findings were summarized on July 3,1990, with.

those persons indicated in paragraph l'of this report.

The licensee

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acknowledged the inspectors findings.--The licensee did not identify as

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_ proprietary:any of the materials provided to, or reviewed by, the

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, inspectors during this inspection.'

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