ML20209J337

Insp Rept 50-382/87-07 on 870301-0415.Violations Noted: Failure to Follow Written Procedures in Maint & Surveillance Programs

ML20209J337
Person / Time
Site:	Waterford
Issue date:	04/27/1987
From:	Jaudon J, Luehman J, Staker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20209J271	List: ML20209J269 ML20209J287 ML20209J337
References
50-382-87-07, 50-382-87-7, NUDOCS 8705040265
Download: ML20209J337 (9)

See also: IR 05000382/1987007

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APPENDIX B

,s_ U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

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NRC Inspection Report: 50-382/87-07 License: NPF-38

Docket: 50-382

Licensee: Louisiana Power &' Light Company (LP&L)

142 Delaronde Street -

New Orleans, Louisiana 70174 _

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Facility Name: Waterford Steam Electric Station, Unit 3 ,

Inspection At: Taft, Louisiana

Inspection Conducted: March 1 through April 15, 1987

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Inspectors: w '//2y/s 7

f r J. G. Luehman, Senior Resident Inspector Date

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T. R~. Staker, Resident Inspector _

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Approved: r . A, M Y 27 / ,

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Jofins P. Jaudory/, Chi , Reactor $ Project '

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Inspection Summary

Inspection Conducted March 1 through April 15, 1987-(Report.50-382/87-07)

Areas Inspected: Routine, unannounced inspkction of: (1) Plant Status,

(2) Licensee Event Report (LER) Follo'wup, (3) Followup on Previously Identified

Items, (4) Monthly Maintenance, (5) Monthly Surveillance, (6) ESF System

Walkdown, (7) Routine Operational Safety Inspection, and (8) Potential Generic

Problems. *

Results: Within the areas inspected, one violation was identified (failure to

follow written procedures in the maintenance and surveillance programs,

paragraphs 6 and 7).

8705040265 870429

PDR ADOCK 05000382

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DETAILS

1. Persons Contacted

Principal Licensee ~ Employees

J. G. Dewease, Senior Vice President, Nuclear Operations

• R. P. Barkhurst, Vice President, Nuclear Operations

• N. S. Carns, Plant Manager, Nuclear

T. F. Gerrets, Corporate QA Manager

S. - A. Alleman, Assistant ~ Plant Manager, Plant Technical Staff

J. R. McGaha, Assistant Plant Manager, Operations and Maintenance

J. N. Woods, QC Manager

A. S. Lockhart, Site Quality Manager

R. F. Burski, Engineering and Nuclear Safety Manager

K. L. Brewster, Onsite . Licensing Engineer

• G. E. Wuller, Onsite Licensing Coordinator

T. H. Smith, Maintenance Superintendent, Nuclear

• Present-at exit interviews.

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

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

administrative members of the licensee's staff.

2. Plant Status

The inspection period began with the plant at full power. On the evening

of March 6-7,:1987, a power reduction was initiated at the request of the

system operator. At 9:01 a.m. (CST) on March 7th, the reactor tripped on

high pressurizer pressure from 56 percent power. In preparation for main

turbine valve testing, an operator was valving in electro-hydraulic

fluid (EH) to the No. 4 governor valve when the turbine intercept valves i

closed. The steam bypass system responded by modulating the bypass

valves. The modulation was not enough to prevent the RCS from reaching

the pressure trip setpoint. RCS temperature was below the temperature

setpoint (566*F) at which the quick opening feature of the bypass valves

is enabled. The licensee investigated and found that the EH fluid supply

and the emergency trip lines to the No. 4 governor valve were cross

connected (apparently during reassembly following the refueling outage).

This caused a loss of EH fluid when the operator opened the EH fluid

isolation valve to the No. 4 governor valve. After correcting this

problem, the reactor was returned to criticality at 10:00 p.m. on March 8,

1987.

On March 15,1987, at 3:44 p.m., a reactor trip occurred from 100 percent

power due to high level in the number one steam generator. The high level

condition occurred because of a malfunction in the No. I feedwater control

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system. After repairs to the feedwater control system, the reactor was

returned to criticality at 5:30 a.m. on March 16, 1987.

After the unit was again on the electrical grid, output was limited as

requested by the system operator. Full power was reached by the reactor

on the morning. of March 20, 1987. At 12:48 p.m. (CDT) on April 13, 1987,

the reactor tripped from 100 percent power. Prior to the reactor trip,

Control Element Assembly Calculator (CEAC) No. I was already in 'an

inoperable status because of a problem with indication on CEA No. 4. At

the time of the trip, Core Protection Calculator (CPC) Channel C was in

bypass undergoing a functional test. While performing this test, the

instrument and control technician placed the memory protect switch in

"0FF" with the selector switch in the "CEAC" position; this caused a

penalty factor to be generated. This penalty factor, in addition to the

penalty factor already present because of the inoperable CEAC, generated

LO-DNBR trips on CPC Channels A and B.

The reactor was again taken critical at 6:30 p.m. on April 14, 1987, and

the main turbine was synchronized with the grid at 9:30 p.m. At 7:01 a.m.

the following day, with the plant at 50 percent power, an automatic

insertion of control element assembly (CEA) Subgroup No. 5 by the Reactor

Power Cutback System (RPCBS) took place in response to an apparent loss of

a main feedwater pump. A primary to secondary mismatch resulted, and at

7:04 a.m. a reactor trip on high steam generator level occurred. This

event was actually initiated by a failure to follow the procedural

sequence of OP-10-001, Revision 8, " General Plant Operations".

Step 8.5.47 of OP-10-001 places the second main feedwater pump in

operation while Step 8.5.48 places the RPCBS in auto. In this case the

second feedwater pump was not latched when the RPCBS was enabled, and the

system, recognizing the apparent loss of the feedwater pump, inserted the

manudlly selected subgroup. Because of the plant power level, there was

no RPCBS initiated runback of the main turbine, and before the plant

operators could correct the induced primary to secondary mismatch, the

reactor tripped.

The reactor was subsequently taken critical again at 11:39 a.m. on

April 15th, synchronized with the grid that afternoon, and the inspection

period ended with the plant at full power.

No violations or deviations were identified.

3. Licensee Event Report (LER) Followup

The following LERs were reviewed and closed. The NRC inspectors verified

that reporting requirements had been met, that causes had been identified,

that corrective actions appeared appropriate, that generic applicability

had been considered, and that the LER forms were complete. Additionally,

the NRC inspectors confirmed that no unreviewed safety questions were

involved, and that violations of regulations or Technical

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Specification (TS) conditions had been identified.

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(Closed) LER 382/85-38, " Fire Wrap Installation Deficiency." The NRC

inspector verified that Procedure ME-3-007 was completed to verify fire

wrap operability.

(Closed) LER 382/85-42, " Reactor Trip Due to Operator Distraction." The

NRC inspector verified that Station Modification 1258, which changes the

power supplies to the four emergency diesel generator air dryers, was

completed.

(Closed) LER 382/86-05, " Surveillance Procedure Error Resulted-in the Hot

Leg Temperature Accident Monitoring Instrument Operability (Not Accuracy)

to be Suspect." The NRC inspector verified that PORC members and

alternates attended training on the requirements of 10 CFR 50.59.

Additionally, the NRC inspector verified that several plant operating

manual procedures were reviewed to ensure that there were no general

implications of inadequate safety reviews. The NRC inspector also

reviewed the changes made to Procedure UNT-1-003 as specified in this LER.

(Closed) LER 382/86-17, " Deficiency in Communications Results in Failure

to Sample the Proper Gas Decay Tank." The NRC inspector verified that

changes were made to the gaseous waste management and sampling procedures

as specified in this LER.

(Closed) LER 382/87-01, " Containment Isolation Valve Inoperable Due to

Improper Reassembly During Construction."

(Closed) LER 382/87-02, "Small Size Snubber Test Failures Due to Improper

Installation." The NRC inspector verified that the snubber installation

procedure was revised as specified in this LER.

(Closed) LER 382/87-03, " Technician Wore Contaminated Clothing Offsite Due

to Failure to Follow Procedures." The NRC inspector verified that

procedure changes were made as specified in this LER. The NRC inspector

observed that loss of power indicators were installed on the PAP portal

radiation monitors.

(Closed) LER 382/87-04, " Fire Barrier Requirement Not Met Due to Lack of

Awareness of Fire Protection Requirements." -

No violations or deviations were identified.

4. Followup of Previously Identified Items

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(Closed) Violation 382/8629-02, " Failure to Verify the Positions of Tw'o

Fire Protection Valves Every 31 Days." The NRC inspector has reviewed.the

licensee's response to this violation which is contained in a letter dated

March 9, 1987. The NRC inspector verified that Procedure OP-903-054,.

Revision 6, " Fire Protection Valve Lineups Check" has been changed to

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incorporate Valves FP-601A and FP-601B.

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(Closed) Violation 382/8629-03, " Failure to Follow a New Fuel Receipt

Procedure." The NRC inspector has reviewed Procedure NE-1-001,

Revision 4, "New Fuel Shipping Container Operations," and verified the

changes discussed in the licensee's March 9,1987, letter had been made.

(Closed) Open Item 382/8704-01, " Revision of MM-12-001." The NRC

inspector verified that this procedure was revised as specified in

LER 382/87-02.

No violations or deviations were identified.

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5. Monthly Maintenance

Station maintenance activities affecting safety-related systems and

components were observed and reviewed to ascertain that the activities

were conducted in accordance with approved procedures, regulatory guides,

industry codes or standards, and in conformance with TS.

Portior.s of the following condition identification work authorizations (CIWAs)

and maintenance procedures were observed by the NRC inspectors:

CIWA 032213 - Replacement of an electro-hy&aulic fluid system

"0"-ring in the dump valve block on turbine (;overnor valve number

two.

CIWA 032325 - Core protection calculator Channel "B" spurious

spikes.

CIWA 020838 "A" component cooling water pump "A" mechanical seal

repair and outboard bearings replacement.

CIWA 027540 - Replacement of position indicator limit switches on

control ventilation area system isolation valve 302 to satisfy

equipment qualification requirements.

Maintenance Procedure ME-4-131, Revision 4, "4.16KV GE Magne-Blast

Breaker" as performed on essential services Chilled Water Chiller "B"

breaker. The NRC inspector observed that Step 8.4.48 of this

procedure uses " maximum clearance" when referring to the minimum

clearance. Also, Step 8.4.47 states that the changing mechanism

ratchet wheel-driving and latching pawl clearances should be

"approximately equal," but there was no criteria established as

to how equal.

Maintenance Procedure MM-4-002, Revision 3, " Vibration Measurements

and Limits for Rotating Equipment," as performed on Charging Pump "A/B."

The NRC inspector observed that many PSA snubbers were installed with lock

wire in the flange bolts /capscrews while several other PSA snubbers were

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installed without lock wire in the flange bolts /capscrews. The NRC

inspector found this to be inconsistent with Procedure MM-12-001,

Revision 1, " Pipe Hanger Support Installation Fabrication, Removal," which

requires lock wire to be installed in the flange bolt /capscrews during

snubber installation per Step 9.5.10. This requirement for lock wires was

found to be consistent with the snubber vendors manual, " Pacific

Scientific Installation and Maintenance Manual." The NRC inspector

questioned the licensee about the lock wire installation inconsistences on

PSA snubbers. The licensee's response was that special high torque

bolts /capscrews were used in lieu of the PSA bolts /capscrews in order to

eliminate the lock wire. Therefore, lock wire was not required, although

it had been inadvertently included in the installation Procedure MM-12-001.

The licensee is also writing a Quality Notice on the failure to follow

procedure (not installing lock wire when installing snuboers since the

installation procedure required it). This failure to follow procedure is

an apparent violation (382/8707-01).

No other violations or deviations were identified.

6. Monthly Surveillance

The NRC inspectors observed and reviewed TS required testing and verified

that testing was performed in accordance with adequate procedures, that

test instrumentation was calibrated, that limiting conditions for

operation (LCOs) were met, and that any deficiencies identified were

properly reviewed and resolved.

The NRC inspector observed portiuns of ME-3-220, " Station Battery Bank and

Charger," specifically, the charger capacity performance test for the

3AB-1 battery charger. This test is performed to verify that the 3AB-1

battery charger meets the requirements of Technical Specification 4.8.2.1.c.4.

During the performance of the test, the NRC inspector noted that the

charger's 125 Vdc output circuit breaker was open but not danger-tagged as

required by Step 8.3.15 of ME-3-220. The NRC inspector discussed this

failure to follow procedures with licensee personnel. Apparently, the

danger-tagging requirements of this procedure are routinely ignored.

Though the failure to use danger tags will not alter the test results or

damage the equipment, the procedure requires them, and failure to use them

could result in inadvertent operation of the breakers during the test.

This failure to follow procedures is a second example of apparent

violation (382/8707-01).

While reviewing Potentially Reportable Event 86-97, the NRC inspector

questioned changes made to the surveillance procedures for measuring the

emergency safety features response times. These changes allowed the

measurement of the pump start time portion for the "A/B" High Pressure

Safety Injection (HPSI) pump with the pump lined up to either the "A" or

"B" train but not requiring measurements for the "A/B" pump lined up to

each train. The licensee stated that the response time measured for the

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"A/B" HPSI pump'would not change when lining the pump up to theI "A" or "B"

train, because the only difference would be a closed contact'between the.

pump and the "A" or "B" pump test power supply. Since the ' licensee's "-

procedures for measurinp the total ESF response time each measure portions..

of the response time with individual acceptance' criteria'for each portion

that prevent the total time from exceeding the ESF limits as specified in

Technical Specification (TS) 3.3.2, the total response time would meet the

TS requirement if all components meet their individual criteria. The NRC

inspector questioned the effect of aligning the "A/B" HPSI pump.to the

different piping systems associated with the "A" or "B" train as performed

when replacing the "A" and "B" pump, and how this change in hydraulic

circuit would affect pump start time. The licensee was already processing

procedure revision and has subsequently changed the HPSI pump start time

Procedure OP-903-029 to require that the "A/B" pump start time be measured

with the pump lined up to each loop.

After the reactor was stabilized at full power on March 20, 1987, the NRC

inspector observed the performance of OP-903-001, " Technical Specification

Logs," Attachment 10.13, " Adjustment of CPC and Excore Instrumentation."

This surveillance was performed to verify the Core Protection

Calculator (CPC) power inputs were consistent with secondary calorimetric

calculations and to change a number of addressable CPC constants to their

full power valves.

No other violations or deviations were identified.

7. ESF System Walkdown

The Essential Services Chilled Water (ESCW) system was verified operable

by performing a walkdown of the accessible and essential portions of the

system on March 10 and 16, 1987.

The NRC inspector used the ESCW valve lineup specified on Attachment 10.1

of Procedure OP-903-062, Revision 3, in conjunction with the referenced

drawings.

While performing the walkdown of the ESCW system, the NRC inspector

observed oil on the floor between the "A" and "B" chillers, which appeared

to be leaking from the "A" chiller. This oil was later cleaned up by

plant personnel. The NRC inspector noted that fire wrap installed on

electrical conduit in this area was in contact with the oil and appeared

to be discolored by the oil. The NRC inspector brought this to the

attention of the plant loss control engineer, who investigated and later

stated that he would initiate a condition identification work

authorization to remove, replace, and study this fire wrap in order to

determine the effect of oil on it.

No violations or deviations were identified.

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8. Routine Operational Safety Inspection

By observation during the inspection period, the NRC inspectors verified

that the control room manning requirements were being met. In addition,

the NRC inspectors observed shift turnover to verify that continuity of

system status was maintained. The NRC inspectors periodically questioned

shift personnel relative to their awareness of the plant conditions.

Through log review and plant tours, the MC inspectors verified compliance

with selected TS and limiting conditions for operations.

During the course of the inspection, observations relative.to protected

and vital area security were made including access controls, boundary. ,

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integrity, search, escort, and badging. ,  ;

On a regular basis, radiation work permits (RWPs) were reviewed an'd'the

specific work activity was monitored to assure the activities were being

conducted per the RWPs. Selected radiation protection instruments were -

periodically checked, and equipment operability and' calibration frequency; .~

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were verified.

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The NRC inspectors kept themselves informed on a daily basis of.overall,

status of plant and of any significant safety matter related to plant

operations. Discussions were held with plant management and various ""

members of the operations staff on a regular basis. Selected portions of_ f

operating logs and data sheets were reviewed daily.

The NRC inspectors conducted various plant tours and made frequent visits

of the control room. Observations included: witnessing work activities

in progress; verifying the status of operating and standby safety systems

and equipment; confirming valve positions, instrument and recorder

readings, annunciator alarms; and housekeeping.

During this inspection period, the NRC inspector reviewed the licensee's

routine use of the Containment Atmosphere Release System (CARS) for

containment pressure control. Paragraph 6.2.1.1.2 of the Waterford Steam

Electric Station Unit No. 3 Final Safety Analysis Report states that the

Containment Cooling System maintains containment pressure during normal

plant operation. While paragraph 6.2.5.2.3 states that the CARS is a

backup to the Hydrogen Recombiner System and is employed only after a loss

of coolant accident (LOCA).

Because of a large amount of noncondensible leakage in the containment

(mostly from control air systems), the Containment Cooling System.has not

been an effective method of pressure control. The CARS has been routinely

used to transfer the containment noncondensibles to the Reactor Building

annulus where the Shield Building Ventilation System processes them to

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release via tha plant stack. The NRC inspector has reviewed the

licensee's present use of CARS, including Section 6.4, " Containment

Pressure Control Using the CAR System," of OP-8-002, Revision 4,

" Containment Atmosphere Release." The procedure is adequate, and the

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routine use-of the CARS during normal plant operations appears to conform

with all regulatory requirements. The NRC-inspector concluded that the

FSAR needs to be revised to reflect this routine use of the CARS.

Licensee management agreed that such a change was'necessary,.and that

additional reviews would be pursued to see that over two years of

operating experience, additional employment of systems, which were not

. fully described in the FSAR, had developed.

While performing a routine plant tour, the NRC inspector noted that the

nuts securing a number of pipe whip restraints in the area of the main ~

steam isolation valves appeared to be improperly installed. The nuts on

some of the restraints, though fully engaged on the stud, were not fully

tightened against the washer and support. Other nuts that appeared to be

up against-the washers were less than hand tight.

The NRC. inspector brought these observations to the attention of the plant

technical support staff. Review by the licensee of installation and~ test

records indicated that the critical tolerance for these u-bolt restraints

is the clearance between the restraint and the pipe circumference.

Therefore, the nuts securing these restraints have to be far enough down

the stud to remove sufficient play to ensure that, should the restraint be

called on to prevent pipe movement, the maximum clearance between the pipe

and restraint would not be exceeded. There did not appear to be a

requirement to torque the nuts; however, the licensee agreed that leaving

them on at least hand tight would prevent vibration-or thermal expansion

and contraction, which could cause the nuts to back down the stud,

allowing the maximum clearance between the pipe and restraint to be

exceeded. The licensee plans to mark the studs on pipe whip restraints to

indicate the maximum nut engagement position to ensure proper restraint

clearance. Followup of the marking of whip restraints is an open

item (382/8707-02).

No violations or deviations were identified.

, 9. Potential Generic Problems .

The NRC inspector provided the licensee with the fol' lowing 10 CFR Part 21

reports:

Defective coaxial cable in post-LOCA High Range Radiation Monitors

(GA Technologies) '

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No violations or deviations were identified.

10. Exit Interview

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The inspection scope and findings were summarized on April 16, 1987, wi,th

those persons indicated in paragraph 1 above. The licensee acknowledged

the NRC inspectors findings. The licensee did not identify as proprietary

any of the material provided to or reviewed by the NRC inspectors during

this inspection.

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