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Insp Rept 50-498/88-24 on 880405-0502.Violation Noted.Major Areas Inspected:Startup Testing Results Review,Operability of Steam Generator power-operated Relief Valves,Monthly Surveillance Observations & Security Observations
	ML20154M564
Person / Time
Site:	South Texas
Issue date:	05/24/1988
From:	Bess J, Carpenter D, Constable G, Hunnicutt D; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20154M561	List: ML20154M557; ML20154M564;
References
	50-498-88-24, NUDOCS 8806020005
	Download: ML20154M564 (17)
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A_PPENDIX .

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U.S. NUCLEAR REGULATORY COMMISSION

REGION IV :

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NRC Inspection Report: 50-498/88-24 Operating License: NPF-76 ,

Docket: 50-498 I

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

P.O. Box 1700

Houston, Texas 770,01

Facility Name: South Texas Project, Unit 1 (STP)

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Inspection At: STP, Matagorda County, Texas

4 Inspection Conducted: April 5 through May 2, 1988

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Inspectors: 4 T!2/ I

T.7. Carpenter, Senior Resident Inspector Date /

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J. E. Bess, Resident Inspector, Reactor Date /

Project Section 0, Division of Reactor

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D. M. Hunrficutt, Senior Reactor Inspector

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Reactor Project Section 0, Division of '

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8806020005 880524

PDR ADOCK 05000498

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Accompanying

Personnel: J. P. Clausner, F ench Commissariar A L'Energie

Atomique, Institute De Protection Et De Surete

Nucleaire

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Approved: w<_ f/2 /[47

'E L. Constable, Chief, Reactor Project Date'

Section D, Division of Reactor Projects

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Inspection Summary

Inspection Conducted April 5 through May 2,1988 (Report 50-498/88-24)

Areas Inspected: Routine, unannounced inspection including startup testing

results review, operability of the steam generator power operated relief

valves, No. 13 standby diesel generator oil' spill essential cooling water

leaks, monthly surveillance observations, security observations, radiological

protection observation, operational safety verification, and engineered safety

feature system.walkdown.

Results: Within the areas inspected, four-apparent violations were identified '

Tlnadequate documentation of tests, paragraph 3a; inadequate review of test

results, paragraph 3c; Operation in TS 3.0.3, paragraph 4; and failure to follow

procedure for equipment clearance orders, paragraph 5).

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DETAILS

1. Persons Contacted

.HL&P

W..P. Evans, Licensing Engineer

G. L.-Jarvela, HP Division Manager

J. E. Gaiger, General Manager, Nuclear Assurance

S. L. Rosen, General Manager, Operations Support

S. M. Head, Support Licensing Engineering

J. J. Nesrsta,' Plant Engineering Manager

T. J. Jordan, Project Quality Assurance Manager

D. N. Brown, Construction Manager,' Unit 1

W. H..Kinsey, Plant Manager-

J. N. Bailey, Manager Engineering,' Licensing

In addition to the above, the NRC inspectors also held discussions with

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various licensee, architect engineer- (AE), constructor and other

contractor personnel during this inspection.

Denotes those individuals ~ attending the exit interview r- n.cted on May 2,

1988.

-2. LicenseeActiononPreviousInspectionFindings927pt

(Closed) Violation (498/8807-01): Failure to~ Maintain Quality

Assurance (QA) Records _

This item concerned surveillance test' records in'the engineering

department record file that were found to' lack required record retention

control, a records custodian-with the responsibility for the control of

records while in the division file had 'not been designated and procedures

did not exist to control access to the division file. The licensee had

taken the following actions: revised Plant Procedure OPGP03-ZE-0004,

"Plant Surveillance. Program" to include provisions that control access to

the division file; provided a.one hour fire rated, lockable filing cabinet

for document storage; established a log to track the removal of Technical

Specifications (TS) packages and provide an index of the file contents;

and designated a records custodian.

This violation is considered closed. <

3. Startup Testing Results Review

During this period of time, the following test results were reviewed:

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1 PEP 04-ZX-0001 Test Sequence for Initial Criticality and Low

Power Testing

IPEP04-ZX-0003 Boron Endpoint Measurement

1 PEP 04-ZX-0004 Isothermal Temperature Coefficient (ITC)

Measurement

1 PEP 04-ZX-0006 N-1 Rod Worth Verification

1 PEP 04-ZX-0007 Rod Cluster Control Assembly (RCCA) Pseudo

Ejection Test

1 PEP 04-ZX-0010 Natural Circulation Verification

The purpose of this inspection was, through the review of completed

procedures, to assess the performance and the overall adequacy of the

tests performed during the' Hot Zero Power (HZP) plateau.

This program, starting with initial criticality on March 8,1038, and

continuing with the low power testing appears to have been conducted,and

completed successfully by-March 16, 1988. It was verified by reviewing

test data and evaluating test results that the acceptance criteria were

met. A detailed discussion of the review of each of these procedures is

as follows:

a. Test Sequence for Initial Criticality and Low Power Testing

(IPEP04-ZX-0001)

During the determination of the Nuclear Heat Test, Step 6.6 of

Procedure 1 PEP 04-ZX-0001, a steam generator power operated relief

valve (PORV) opened.

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The NRC inspector identified that this information was not noted in

the chronological test log, although the' test sumary mentions that a

PORV opened momentarily and that the problem was corrected quickly J

(the valve was stuck open about 20 percent for about 45 seconds).

The information in the test sumary, however, does not describe which i

valve opened, under what circumstances the event occurred, and what I

corrective actions were taken. Some available information of this i

failure was found in Maintenance Work Request (MWR) MS-48225, issued

soon after the event occurred. It appears that during

troubleshooting, signs of hydraulic leakage were found, but the

source of the leak was not determined,

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The MWR (page 4a paragraph "as found condition") also indicated that '

the valve (1) would not move from 30 percent open, and (2) later

moved from 30 percent to 90 percent open with no signal either from

the control room or locally. Reasons to explain these deficiencies

were not documented. In addition, a review of the operator log book

did not provide any new information, except documentation of entrance

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intothelimitingconditionforoperation(LC0)correspondingtothis

problem (TS 3.7.1.6)

Th'e NRC inspector considers that'the licensee should have adequately

documented this malfunction affecting-safety-related equipment.

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Therefore, this. constitutes of an apparent violation (498/8824-01) of

the licensee's procedures,

b. Boron Endpoint Measurement (1 PEP 04-ZX-0003)

The purpose of this procedure was to determine the boron endpoint

concentrations at various configurations as specified in the test

sequence for initial criticality and low power testing.

The only specific boron endpoint acceptance criterion given in the

procedure is for the .all rods out- (AR0) condition. The predicted

value was 917 50 parts per million (ppm). The actual measured value

was 953.5 ppm. For the other configurations, the boron endpoint

concentrations measures were Control Rod Banks; CD-in; 896.8 ppm;

CD+CC-in; 817 ppm; CD+CC+CB-in; 717.9 ppm; CD+CC+CB+CA; 624.4 ppm.

All these values appeared to be very close to the predicted value.

The NRC inspector verified by reviewing test data and independently

evaluating test results from the Baron Endpoint Measurement test

procedure performed on March 8-11, 1988, that the licensee had

complied with the test requirements and that the acceptance criteria

were met,

c. ITC Measurement (11 ?04-ZX-0004)

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The NRC inspector evaluated and verified that the licensee met the

requirements and conditions of the facility license by reviewing test

data and evaluating test results from the ITC measurement test

performed from March 8-11, 1988. The NRC inspector ascertained that I

the licensee was adhering to station procedures and that te:,t program I

records were adequete. l

The purpose of this test was: (1) to determine the ITC of reactivity

at three different control rod configurations:

AR0

Control Bank D in

Control Banks D and C in

(2) to derive the AR0 MTC, and (3) to perform the surveillance

requirement of TS related to the MTC limit which constitutes the

safety criterion.

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Reviewing the AR0 and Bank D configurations, the NRC inspectors noted

that.during.the beginning of the test, cooldowns had been suspended

because of a cooldown rate too fast (the procedure required

approximately 10 F/ hour). After resetting conditions, the' test was

resumed at a slower rate.

ITC was detenninated by measuring the slope of reactivity change

versus change in average RCS temperature during heatup and cooldown )

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Reviewing the AR0 configuration results, the NRC inspectors noticed a l

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discrepancy between theand

the heatup configuration slope

theofreactivity

reactivitychange

change recorded

value (deltaduring

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noted on the data sheet. After verification, the licensee agreed to the i

identified error and consequently committed to correct this error. '

The value given by Westinghouse in Procedure T6X/ TUX-SV3.3.6, step !

No. 6.2 specifies that the measured values of the ITC obtained from

heatup and cooldown cgree within +1 pcm/ degree Fahrenheit. This

value was not considered a criterTon by the licensee and,

consequently was not included in the Procedure 1 PEP 04-ZX-0004.

Howuer, in this particular case, it would have permitted the test '

director and the verifier to note that the ITC values found were l

outside of 1 pcm/ degree Fahrenheit criterion.

Although this error does not impact on the acceptance criteria of the

test procedure (the hic value become J 44 pcm/ degree Fahrenheit '

instead of -2.06 pcm/ degree Fahrenheit which is in accordance with

the TS requirement), this error was not found during the verification !

process. The NRC inspector noted that the licensee then took

appropriate action to correct this error. A Station Report j

Problem (SRP) 88-0130 was issued on April 15 that involved the l

procedure correction and required reverification of all test data i

recorded during the core physics testing on April 16, 1988.

This constitutes an apparent violation (498/8824-02) the licensee's

procedural requirements,

d. N-1 Rod Worth Verification (1 PEP 04-ZX-0006) l

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The NRC inspector evaluated and verified that the licensee met the

requirements and acceptance criteria of the procedure by reviewing l

test data and evaluating . test results for the N-1 Rod Worth l

Verification Measurement ' test performed on March 12 and 13,1988, i

The purpose of this test was to verify that the HZP insertion limits

defined under TS provide adequate shutdown margin with the most

reactive RCCA defined as being RCCA F-14 (Control Bank B) stuck in

the withdrawn position.

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The total worth of all RCCA Banks, less the most reactive RCCA, was

determined to be 8283.8-pcm. This value met the acceptance criterion

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of 7661 766 pcm. The HZP shutdown margin is given by the worth of !

all rods less the most reactive RCCA minus the worth of all rods

.above the Rod Insertion Limit (specified in TS 3.1.3.6 and obtained

fromtheNuclearDesignReport). This value (8283.8 - 1429 =

'6854.8 pcm) exceeds'the minimum value of 1750 pcm required by

TS 3.1.1.1.

Since these results met the acceptance criteria, the NRC inspector

concluded that the test was satisfactory,

e. RCCA Pseudo Ejection Test (1 PEP 04-ZX-0007)

The NRC inspector verified by reviewing test data and evaluating

test results from the RCCA Pseudo Ejection Test performed on March 13

and 14, 1988, that the licensee complied with the test requirements

and the acceptance criteria of ~ the. test procedure were met.

The purpose of this procedure was to verify the conservatism of the

assumed worth of an ejected rod at zero power and of the assumed hot

channel factors for an ejected-rod at zero power from the

configuration assumed,in the Safety Analysis Report (SAR).

The reactivity change caused by the withdrawal of the most reactive

rod (Rod D-12) was used to determine the worth of the rod. The

equilibrium boron concentration and the reactivity change caused by

the withdrawal of the rod was used to determine the boron endpoint

concentration. A flux map was performed then evaluated and the hot

channel factors were detennined.

The NRC inspector noted that two Field Change Requests (FCR) were

initiated in this test to correct inadequacies in the test procedure.

In particular, FCR 88-0530 was initiated to change measurement

uncertainty so as to match the Westinghouse-startup procedures.

The NRC inspector verified that licensee technical reviews were ,

performed and confirmed the value as correct from Westinghouse

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startup Procedure TGX/THX-SU3.1.1, Revision 1.

During this test, a Test Deficiency Record (TDR)88-028 was j

generated. The test method consists of determining the core

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reactivity to endpoint condition with the controlling bank at the

zero power insertion limit, the most reactive rod being borated out

to approximately 230 steps, then fully withdrawn. At this point, the

procedure specifies (Step 6.12.6.3), "Allow the neutron level to

increase until the flux signal is greater than 30% of full scale and

the indicated reactivity is constant." When the most reactive rod

(0-12) was pulled out it was observed that the reactivity addition

resulting from withdrawal from 228 to 259 steps was only 2.33 pcm due

to the single rod moving through the region of low relative flux in

the upper portion of the core. This very small amount of reactivity ,

addition resulted in the flux signal increasing very slowly (period l

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of approximately 4500 seconds). Because the raactivity trace had

stabilized .and the flux signal was slowly inci .asing at a rate

consistent with the very small reactivity addition, rod 0-12 was

reinserted and the reactivity change measured prior to-the flux

increase above-30 percent of full scale as called for in the

procedure.

The licensee repeated the test three time's and an average reactivity.

was used in the end point calculation. Although this was not

consistent with the guidance in the procedure, the licensee

considered the reactivity measurement valid for the following

reasons:

(1) the reactivity trace had stabilized prior to rod reinsertion.

(2) the flux trace increased at a rate consistent with a such small

reactivity addition.

(3) the reactivity change resulting from the endpoint was so small

that the measurements resulting were especially susceptible to

the introduction of an additional error due to temperature

changes. Thus, by attempting to delay reinsertion of the rod

until the flux signal reached 30 percent, the endpoint

measurement accuracy would have been affected by the negative

reactivity feedback from the ITC.

Westinghouse provided information that the basis for the 30 percent

value used in the startup procedure is to allow for margin above the

20 percent minimum value and that the additional error resulting from

reactivity measurements with flux level between 20 and 30 percent is

negligible. If the reactivity measurement (2.33 pcm) had a

100 percent error, the difference in the total measured worth of RCCA

D-12 should be only 0.57 percent and has no effect on the test

results.

Therefore, the worth of the. ejected RCCA at HZP was determined to be

407.9 pcm. The worth must be less than or equal to 860 pcm (Final

Safety Analysis Report (FSAR), 15.4.8) and the peak hot channel

factor FQ(z) resulting from the ejected rod at HZP was determined to

be 7.269. This value is to be less that or equal to 13 (FSAR,

15.4.8).

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Since these results met the acceptance criteria, the NRC inspector

concluded that the test was satisfactory.

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f. Natural Circulation Verification (1 PEP 04-7X-0010)

The NRC inspector verified by reviewing test data and evaluating test

results from the Natural Circulation Verification test procedure

performed on March 14 and 15, 1988, that the licensee complied with

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the' test requirements and the acceptance criteria of the procedure

were met.

The purpose of this procedure was to demonstrate the ability of the

RCS to. remove decay heat from.the reactor core and to obtain coolant

flow and temperature distribution data under a condition'of loss of

forced convective cooling.'

During the power ascension while delta-T indicated about 1 percent

power,.the nuclear power range channels indicated 8 percent. The

licensee performed the Procedure'1 PEP 04-ZY-0040 (Initial Adjustment

of Nuclear Instrumentation) twice to adjust the gains of the power

range instrumentation to the power shown by loop delta-T. This

adjustment was necessary tu avoid the P10 interlock actuation.

The natural circulation was established and maintained for one hour.

During this period of time, the acceptance criteria were verified as

indicated by steam pressure remaining constant, the RCS hot leg

temperature approximately equal to core exit thermocouple temperature

and remaining constant, and RCS cold leg temperature approximately

equal to the saturation temperature for the indicated steam pressure.

Thermocouple maps were obtained while natural circulation was

maintained.

The NRC inspector noted that the TSAT had to be recalculated because

of wrong values of the Loop 1 cold leg temperature. It appears that

this problem was found after the completion of the test and a MWR was

issued to troubleshooting. However, this problem was not reported on

the test summary (Section 5.0 - problems encountered'during testing).

Further, in a previous NRC inspectie report (50-498/88-12), NRC

inspectors noted that the reactor cou tnt system (RCS) flow

measurement test results appeared to lack in documented test

information and test data. The above observation is an additional

example that indicates the licensee must provide more detailed

information in the chronological -test log and test sunnary,

especially when problems are experienced during test performance.

These problems are additional examples of inadequate test

documentation apparent violation 498/8824-01, noted in paragraph 3.a.

4. Operability of Steam Generator PORVs

On April 12, 1988, the licensee received notification from

Paul-Munroe/ENERTECH, the vendor, that seal materials installed in the

hydraulic actuators and hydraulic pumps for the steam generator PORVs in

both Unit 1 and 2 were not of the correct material and were incompatible

with the hydraulic fluid (FYRQUEL-EHC). Subsequently, a letter dated

April 18, 1988, was received by the NRC Region IV Administrator from

Paul-Munroe/ENERTECH providing the details of this issue. Specifically,

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the seai kits installed in the PORV pump shaft seals and hydraulic

cylinder rod seals were specified as Viton (a fluorcarbon) and were

accompanied by a Certificate of Conformance (C0C) from the seal vendor

(ParkerCylinderDivision). However, in fact, the seals installed were

BUNA-N, a nitril, that undergoes erosion when in contact with FYRQUEL-EHC

hydraulic fluid. Erosion of the seal materials could prevent proper

operation of the PORVs.

The PORVs are identified in the TS as the atmosphere steam relief valves.

The TS Section 3.7.1.6 requires that all four PORVs be operable in

Modes 1, 2, 3, and 4. With one PORV inoperable, the plant is in a 7-day

actionstatement(TS3.7.1.6.a). With two PORVs inoperable, the plant is

in a 72-hour action statement (TS 3.7.1.6.b). Having more than two PORVs

inoperable is not defined in the action section of TS 3.7.1.6. Thus, with

more than two steam generator PORVs inoperable, the plant would be

governed by TS 3.0.3 which would require initiation of a plant shutdown

within one hour and be in cold shutdown (Mode 5) within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> or

resolve the condition which would allow resumption of operations under an

action statement or LCO.

On April 23, 1988, the licensee issued a 10 CFR 50.59 Safety Evaluation

(SE) and a justification for continued operation (JC0) which addressed the

operability status of the four PORVs with the improper BUNA-N seal

material . Tne PORVs are safety-related and are required to be

environmentally qualified (EQ) because they are located in the isolation

valve cubicles (ICVs) and as such could be subject to high temperature,

high humidity, spray, chemical and radiation effects under certain

accident conditions. The NRC inspectors did note, however, that the SE

and JC0 failed to consider that the BUNA-N seal material would undergo

degradation in a radiation field that could be expected to be present

under certain accident senarios, for example gross steam generator tube

failures. The EQ aspect of this issue will be addressed in NRC Inspection

Report 50-498/88-30. The SE and JC0 concluded that plant operation could

continue as long as at least two PORVs remained operable as defined by an

enhanced surveillance program. This program included (1) the manual

timing of the hydraulic pumps cycle frequency for the three worst PORVs

(A, B, and D) once per shift, (2) the visual examination of the PORV skids

once per shift, and (3) a stroke test of all four PORVs once per day. A

pump cycle frequency of less than 10 minutes, gross oil leakage, or a

failure of the stroke test would result in a declaration of inoperability

for the applicable PORV that would lead to the appropriate action

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1988, at 1:30 a.m. (CST), the "D" PORY was declared

OnApril23(TSactionstatement3.7.1.6.a).

inoperable The same day, at 7:30 p.m.,

the "A" PORV was declared inoperable due to excessive hydraulic pump l

cycling (TS action statement 3.7.1.6.b). On April 24, 1988, at 4:22 p.m. ,

(CST) the licensee exceeded the TS action statements in order to comply I

with the SE/JC0 surveillance requirement and isolated a third PORV ("C" l

PORV) which by default placed the plant outside the requirements of

3.7.1.6 and into TS 3.0.3. The licensee entered TS 3.0.3 for operational

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convenience with no stated intent of shuttin'g do'wn. Eight minutes later,

the "C" PORV was returned to service and the licensee exited TS 3.0.3 and

consequently returned to TS action statement 3.7.1.6.b. At 4:33 p.m.', the

licensee isolated "B" PORV for stroke testing' and, for operational

convenience, again entered.TS 3.0.3 for 10 minutes.. . These events were

recorded in the reactor operators log book,' including the statement-on'

entering and exiting TS 3.0.3. The "A" and "D" PORVs were subsequently

repaired that week, returned to service and TS 3.7.1.6. was exited.

0n April 30, 1988, at 6:20 p.m. (CST), the "A" PORV failed its stroke.

test, was declared inoperable and TS 3.7.1.6.a was entered. At 9 p.m. ,

the same day "B" PORY failed because.of a problem that was not associated

with the hydraulic seal problem. .The plant was steady state'at 30 percent

reactor power when the "B" PORV was driven full open. The control room

dispatched an operator who shut the manual block valve'after about six

minutes. The "B" PORV remained open about another nine minutes, then

closed. The "B" PORV was then declared inoperable and TS 3.7.1.6.b was

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entered.

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printout for theNRC "B" inspectors reviewed

PORV activation. The the plant

PORY computer

opened (ERFDADS)(stroke

and closed

time and rate) exactly as if a demand signal-was present. .However, no

such signal or need for that signal either manually or automatically from

the QDPS computer.was indicated on.the computer printout. The plant then

had two of the four PORVs failed under different failure modes, one of

which was undetermined and unanalyzed since it was outside of. the scope of

the SE and JC0 that was written for the BUNA-N hydraulic seal' issue. ~ :

The NRC inspectors discussed with a shift supervisor the stroke testing of l

the two remaining operable PORVs as required by the SE and JC0 without ;

exceeding the action statement of TS 3.7.1.6 and entering TS 3.0.3, for

operational convenience. The shift supervisor indicated that it appeared I

the licensee would have to enter TS 3.0.3. The NRC inspectors then asked

if it was acceptable to voluntarily enter TS 3.0.3 for operational

convienence with no intent to shutdown. The shift supervisor indicted he

was not sure.

On May 1, 1988, at 9:15 p.m. (CST), the. licensee began.a normal shutdown

of Unit 1, one day ahead of schedule. The licensee indicated that the

reason for the early shutdown was that 30 percent plant testing was ,

essentially complete and thLt the licensee would not be able to meet the l

enhanced surveillance testing requirements established by the SE and JC0

without entering TS 3.0.3.

As a result of this inspection, the NRC inspectors concluded that the SE

and JC0 was incomplete in that it ignored the effects of radiation; the

licensee violated TS 3.7.1.6 on two occasions the day of April 23, 1988,

when three_PORVs were inoperable and entered TS 3.0.3 for operational

convenience; and that the staff and management of the licensee does not

have a uniform understanding of the bases for and application of TS 3.0.3.

The two instances of the violation of the requirements of TS 3.7.1.6 is an

apparent violation of NRC-requirements. (498/8824-03).

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5. No.13 Standby Diesel Generator (SDG) Oil Spill - Unit 1 I

On April 19, 1988, an Equipment Clearance Order (ECO) was issued per Plant

Procedure OPGP03-ZO-0001, Revision 7, "Equipment Clearance," on the

safety-related No.13 SDG lubricating oil (LO) circulation pump. The

system was tagged out per the ECO. In preparation for disassembly, the

pump required draining. The control room concurred with draining the

small amount of L0 from the pump into the No. 13 SDG sump. Drain

Valve LU-0071 was opened and ralve LU-0119 was un4cked and opened. An

entry was made in the Loc'Ked Valve Deviation logbock, However, none of

these valves were identified as boundary valves requiring clearance tags

on the ECO. Subsequently, these valves were left open and a change of

plans lead to the deferral of the L0 circulation pump work till another

time and the EC0 was cleared.

The system alignment was reestablished per the ECO. The L0 circulation

pump and heater was turned on. After about 5 minutes of pump operation,

the control room noted No. 13 SDG LO was reading approximately 170 F and

2 psig pressure. The operator went to investigate and noticed smoke

coming from the LO heater. He deenergized all power to the SDG. His

investigation determined that the L0 drain valves were open and th L0

sump was about 1100 gallons below normal, i

A review of the event indicated there was an inadequate lineup review, no

independent ..rificatica of system restoration, and no review of the ;

Locked Valve Deviation log prior to energizing the SDG support system, i

Timely observation and action by the crew prevented a fire and the !

resultant damage to the SDG.

After the event, the NRC inspector observed the end of the cleanup of the

spilled oil. The area was posted "No Smoking" and there was security

guards posted to control access and work activities. The events leading

to this oil spill involving the failure to follow the appropriate ,

procedures is considered a violation of NRC requirements (498/8824-04). I

6. Essential Cooling Water (ECW) Valve Body and Pipe Fitting Leaks

On April 1,1988, the licensee found evidence of leakage on small bore

(less than two inches) valves and pipe fittings of the safety-related ECW

system. The ECW is constructed of welded aluminum bronze material. Three ,

of these small bore fittings were initially removed and sent off-site for I

analysis.

The licensee reported preliminary results of Bechtel metallurgical

laboratory ar,alysis on two 1-inch sockets removed from the essential

cooling water system. The sockets exhibited corrosion with apparent

through wall leaks. Microsections of the aluminum bronze meterial

revealed selective corrosion associated with the dealuminization of one

phase of the bi-phase alloy used in castings for small bore socket joints.

Selective corrosion has not been observed on the single phase aluminum

bronze pipe. The corrosion appears to be associated with the socket

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crevices. The actual weld and heat affected zone have not exhibited

corrosion. The bi-phase alloys were not expected to corrode in this

manner; however, improper heat treatment of these fittings could lead to--

the observed corrosicn. The bi-phase alloy normally consists of an alpha

and beta phase. -If the metal is not cooled quickly during heat treatment

then a_ gamma-2 phase alloy forms which is susceptible to corrosion.

The average deterieration-of wall thickness was up to 88 percent. The

system has- been in use for. three. years. Evidence of through wall leaks

(amounting to a maximum of 8-10 ml per day) were observed on approximate

70 sockets. The corrosion has only been observed on 2- and 1-inch

fittings. Larger pipe is not joined with socket welds. The licensee is

identifying all applications of the suspect two-phase alloy in the system

for evaluation.

The corrosion appears to progress slowly and a monitoring program is being

developed to measure the-rate of. deterioration.of the fittings. The

licensee's analysis indicates that if-70 percent of the wall thickness

were lost, the system would still retain its structural integrity.

Therefore, the licensee considers the ECW system to be operable with no

immediate safety concerns. The licensee plans to complete the evaluation

of the problem and to establish a'n action plan acceptable to the NRC prior

to exceeding the 30 percent power testing plateau. NRC Region IV and NRR

are closely monitoring the licensees evaluation and corrective action plan

development.

7. Monthly Surveillance Observations - Unit 1 (61726)

The NRC inspector observed selected portions of surveillance testing and l

reviewed completed data packages to verify that TS requirements are being

met for safety-related systems and components. The following surveillance

tests were observed:

1 PSP 03-SI-006, Revisions 5, "High Head Safety Injection Pump 1C

Inservice Test"

1 PSP 02-SP-0008B, Revision 0, "Train B Diesel Generator Slave Relay

Test"

1 PSP 03-DG-0002, Revision 5, "Standby Diesel 12 Operability Test"

The NRC inspectors verified the following items during the inspection:

Test results were reviewed by personnel other than the persons

directing the test.

The surveillance testing was completed at the required frequency per

TS requirements.

Testing was performed by qualified personnel using approved

procedures.

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Removal and restoration of the affected system and or components were

accomplished.

Test instrumentation was calibrated.

No violations or deviations were identified;

8. Security Observations (71881)

The NRC inspectors verified the physical security plan was being

implemented by selected observation of the' following items:-

Security monitors at the secondary and central alarm stations were

functioning properly for assessment of possible intrusions.

Persons and packages were properly cleared and checked before entry

into the protected area (PA) was permitted.

The security organization was properly staffed..

The PA barrier was maintained and the isolation zone kept free of

-

transient material.

Vital area barriers were maintained and not compromised by breaches

or weakness.

Illumination in the PA was adequate to observe all areas during hours

of darkness.

No violations or deviations were identified.

9. Radiological Protection Observations (71709)

The NRC inspectors verified that selected activities of the licensee's

radiological protection program were implemented in conformance with

regulatory requirements. The activities listed below were observed:

Radiation work permits contained the appropriate information to

ensure work was performed in a safe and controlled manner.

Personnel

area (RPA) properly frisked prior to exiting the radiation protected

.

Personnel in the RPA were wearing the required personnel monitoring

equipment.

Radiation and contaminated areas were properly posted based on the

amount of activity levels within the area.

No violations or deviations were identified.

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10. Operational Safety Verification (71707)

The objectives of the inspection'were to ensure that the plant is being-

operated in a safe manner and in conformance with regulatory-requirements,

the licensee's management controls are effective in discharging their

responsibilities, and TS requirements are being met.- The NRC inspectors

visited the control room on a daily basis.to verify the_following:

Operators adherence to approved procedures: and TS requirements.

Control room staffing was proper.

Management personnel toured the control room on a regular basis.

Operability of the safety parameter display system.

MWRs were written for equipment / components in need of reairs.

Conduct of reactors were professional.

No violations or deviations were identified.

11. Engineered Safety Feature (ESF) System Walkdown (71710)

The NRC inspectors conducted a walkdown of the accessible portions of

Train "A" of the ECW system to verify operability of the system. A review

was performed to confirm that the licensee's system operating procedure

matched plant drawings and the as-built configuration. Eqt. jpment

conditions, valve and breaker positions, housekeeping, labeling, permanent

instrument indications and apparent operability of support systems

essential to actuation of the ESF system were all noted as appropriate.

The NRC inspectors identified the following observations to licensee

management:

a. MWR Tag No. 02099, which was attached to a conduit junction box was-

not filled out completely. The tag did not have a date which the tag

was hung as required by Step 4.3.9 of Procedure OPGP03-ZM-0003.

_

b. Permanent identification tags were not installed on Valves EW-0259, i

I

EW-0198, EW-0018, EW-0017, and EW-0120.

c. Emergency Diesel Generator ECW Return Isolation Valve EW-0019 is 1

incorrectly identified as a test connection on the attached i

identification tag. I

d. A large diameter overhead conduit in the Train "A" Supplemental

Chiller Room was missing a coverplate.

e. An unidentified cable was found hanging from cable trays CIXMIATJAA

and CIXMIATSVA in the Train "A" Supplemental Chiller Room,

f. The tubing channel supporting instrument root valves EW-0009 and l

EW-0010 plus the associated tubing was not attached to its support.

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g. Train "C" Return loop Chlorine Analyzer Panel mounting bolt was not

installed correctly. Additionally, the panel did not have a

permanent identification, tag,

h. ECW system Drawing No. SR289F05038, Revision 13, shows Valve EW-0117

as locked shut and. Valves EW-0020 and FW-0259 as locked open. These

valves'were found in the open and throttled-positions respectively,

as required in the system operating Procedure 1P0P02-EW-001-1.

1. The vent cap downstream of the common vent valve NI-EW-PSL-6882 and

NI-EW-WPI-6882 was not installed.

j. The local control station for the Standby Diesel Generator No.11

Jacket Water Makeup Valve is missing the valve control switch,

k. Valve EW-0331 had a work traveler from 1985 attached.

The above listed discrepancies do not appear to be safety significant;

however, collectively they represent weaknesses on the part of the

licensee regarding attention to detail in safety-related activities.

No violations or deviations were identified.

12. Exit Interview

The NRC inspector met with licensee representatives (denoted in

paragraph 1) on May 2,1988, and summarized the scope and findings of the 1

inspection. Other meetings between NRC inspectors and licensee management

were held periodically during the inspection to discuss identified

concerns. The licensee did not identify as proprietary any of the ;

information provided to or reviewed by the inspectors during this i

inspection. !

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