ML20133B987

Insp Rept 50-382/85-16 on 850401-0531.Violation Noted: Failure to Identify Affected Document,Presence of Contamination Outside Radiation Control Areas & No Evidence of Completed Fire Appliance Impairment Form

ML20133B987
Person / Time
Site:	Waterford
Issue date:	07/18/1985
From:	Constable G, Flippo T, Andrea Johnson, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20133B972	List: IR 05000401/2005031 ML20133B969 ML20133B987
References
50-382-85-16, NUDOCS 8508060312
Download: ML20133B987 (8)

See also: IR 05000382/1985016

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APPENDIX B

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REGION IV

U. S. NUCLEAR REGULATORY COMbiISSION

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'NRC Inspection Report: 50-382/85-16

License: NPF-38

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Docket: 50-382

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

,'142 Delaronde Street

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New Orleans, Louisiana- 70174

Facility-Name: Waterford Steam Electric Station, Unit 3

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Inspection At

Taft, Louisiana

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Inspection Conducted: April'l through May 31, 1985

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Inspectors:

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Senior Resident Inspector

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T. A. Flippo', Resident Inspector

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W. B. Jones',' Reactor Inspector

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A. R. Johnson, Reactor Inspector

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Approved:

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G. L. ConstableT Chief

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Reactor Project Section B

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Inspection Summary

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. Inspection. Conducted April 1 through May 31,-1985 (Report 50-382/85-16)

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. Areas Inspected: Routine, announced inspection'of:

(1) Control of

Design-Changes; (2) Radioactive Contamination of Secondary System; (3).Startup

Test Procedure Review; (4) Phase III Test Witnessing; ,(5) Test

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Results' Evaluation;.and (6) Survey of Licensee Responses to Selected Safety

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. Issues. ~The inspection-involved 638 inspector-hours onsite by 4 NRC

inspectors.

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Res'ults: Within the areas inspected, 3 violations were identified.

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DETAILS

1.

Persons Contacted

Principal Licensee Employees

• R. S. Leddick, Senior Vice President, Nuclear Operations

• R. P. Barkhurst, Plant Manager, Nuclear

• T. F. Gerrets, Corporate QA Manager

• L. F. Storz, Assistant Plant Manager, Operations and Maintenance

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

• 0. D. Hayes, Operations Superintendent

L. M. Meyers, Assistant Operations Superintendent

• J. R. McGaha, Maintenance Superintendent

• R. G. Pittman, Operations QA Audit Supervisor

• J. N. Woods, Plant Quality Manager

• K. L. Brewster, Onsite Licensing Engineer

G. E. Wuller, Onsite Licensing Coordinator

• T. C. Payne, I&C Supervisor

• G. F. Koehler, Operations QA Engineer

• Present at exit interviews.

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

with various operations, engineering, technical support, and

administrative members of the licensee's staff.

2.

Plant Status

The Waterford 3 site is presently in the startup testing phase.

The 80%

testing plateau has been completed and the plant is in an approximate 3

week outage to perform preventive maintenance associated with a chemical

buildup in the main electrical generator.

3.

Control of Design Changes

On March 3, 1985, the NRC inspector observed that the licensee had

implemented the modifications described in Station Modification Package

(SMP) 760.

This SMP changed the automatic isolation of component

cooling water (CCW) to the reactor coolant pump (RCP) integral seal

cooler from high CCW discharge seal cooler pressure to high CCW

discharge seal cooler temperature.

In addition, SMP 760 added control

switches on CP-2 which allow the operator to reopen the CCW isolation

valves.

While reviewing LP&L's off-normal operating procedures on April 10,

1985, the NRC inspector noted that OP-901-022, Revision 1, "High

Activity in Component Cooling Water System," had not been revised to

reflect the above modifications.

The procedure indicated that CCW to

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the RCP integral seal cooler isolated on a 2/3 high pressure indication;

however, the station modification isolated the CCW line on a 1/1 high

temperature indication.

In addition, the procedure did not address the

installed switches on CP-2 for reopening an isolated CCW line and the

100 second interlock that will reclose the valve if the high temperature

indication is still present.

The NRC inspector reviewed Plant Engineering Procedure PE-2-006,

Revision 5, " Station Modifications," Attachment 6.14. " Document Update

Record," to determine if OP-901-022 had been identified as needing

revision during the SMP 760 review process.

The review revealed that

OP-901-022 had not been identified as an affected document; however,

Off-Normal Operating Procedure OP-901-010, Revision 1, " Reactor Coolant

Pump Malfunction," and Operating Procedure OP-01-002, Revision 4,

" Reactor Coolant Pump Operation," were identified as affected documents

on Attachment 6.14 and were subsequently revised to reflect the changes

made in SMP 760.

This is a violation. (382/8516-01)

4.

Radioactive Contamination of Secondary System

On April 3, 1985, the licensee d tected small amounts of radioactive

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contamination, approximately 10

uCi/ml, at the discharge of the

condenser vacuum pumps.

A walkdown of the secondary system by the

licensee revealed that an auxiliary operator had failed to close Boron

Management Valve BM-435 A(B) as required by LP&L Operating Procedure

OP-07-001, Revision 4, " Boron Management System," after the last

operation of the boric acid concentrator (BAC).

Boron Management Valve

BM-435 A(B) normally isolates the boric acid concentrator steam chest

from the radioactive system relief header.

The relief header is fed by

various chemical volume control systems (CVCS) and BM relief valve,

including CVCS letdown relief valve CVC-115, and discharges to one of

the four holdup tanks.

During operation of the BMC, steam is fed to the BAC steam chest through

the auxiliary steam header.

BM valve BM-435 A(B) is then opened to

allow venting of noncondensables to the relief header. When the

auxiliary operator failed to close BM-435 A(B) after closing down the

BAC, a direct flow path from the radioactive relief header to the

auxiliary steam header and gland seal leak-off tank was provided.

Earlier, on April 3, 1985, the CVCS letdown line experienced a flow

transient which caused a pressure spike downstream of the letdown

orifice valves.

This resulted in CVCS letdown relief valve CVC-115

opening to reduce the pressure spike; however, the valve failed to

properly reseat after the pressure was reduced producing approximately

1 gpm RCS leakage to the relief header.

The relief header was properly

aligned to a holdup tank, but backpressure in the tank bypassed the RCS

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water to the BMC steam chest.

From the steam chest, the RCS water

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contaminating both.

In addition, the gland seal leak off tank.

discharged.to the condenser.

From the condenser the contaminated

condensate made its way into the polishers and steam generators via the

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' normal secondary system flow path.

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' Samples.of the condenser exhaust tank by the licensee revealed the

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following:

Xe - 133

6.8 x 10[ uCi/ml

Xe - 135

3.9 x 10

uCi/ml

Kr - 85u

1.1 x 10

uCi/ml

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In addition the licensee sampled the turbine building industrial waste

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sump.

The samples taken revealed the follow ng:

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Xe - 135

1.2 x 10[ uCf/mi

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Co - 58

7.'6 x 10,

uCi/ml

'Na - 24

7.5 x.10,

uCi/ml

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I - 133

1.4 x 10

uCi/ml

Zr - 97

8.5 x 10

uCi/ml

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Upon verifying the presence of contamination outside the radiation

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control areas, the licensee isolated the affected areas and established

controls in accordance with their approved procedures to prevent the

spread of contamination.

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.This is a violation..(382/8516-02)

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Startup Test Procedure Review

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The NRC inspectors reviewed the startup test procedures for power

ascension testing of the plant.

The procedures were reviewed for

technical content, compliance with Final Safety Analysis Report (FSAR),

and compliance with licensee's administrative procedures. ~The startup

test procedures reviewed are listed below.

SIT-TP-707

Atmospheric Steam Dump and Turbine Bypass Valve Capacity

Checks

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SIT-TP-726

Remote Reactor Trip with Subsequent Remote Plant Cooldown

No violations or deviations were identified.

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6.

Phase III Test Procedure Witnessing

The NRC inspectors observed the performance of portions of the following

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Phase III test procedures:

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SIT-TP-716

Core Performance Record

SIT-TP-718

Variable Tavg Test

SIT-TP-707

Atmospheric Steam Dump and Turbine Bypass Valve Capacity

SIT-TP-724

Temperature Decalibration Verification

SIT-TP-728

Loss of Offsite Power Trip Test

SIT-TP-726

Remote Reactor Trip with Subsequent Remote Plant Cooldown

SIT-TP-755

Natural Circulation Demonstration Testing

During the performance of the test, the NRC inspectors verified the

following:

a.

The personnel conducting the test were cognizant of the test

acceptance criteria, precautions, and prerequisites prior to

beginning the test.

b.

The test was conducted in accordance with an approved procedure and

the test procedure was used and signed off by personnel conducting

the test.

c.

Data was collected and recorded as required by the test procedure

instructions.

No violations or deviations were identified.

7.

Test Results Evaluation

The NRC inspectors reviewed Phase III test results to verify that: (1) all

changes, including deletions to the test program, had been reviewed for

conformance to the requirements established in the FSAR and Regulatory

Guide 1.68; (2) deficiencies had been adequately addressed and corrective

action completed; (3) the licensee correctly analyzed the test data and

verified that it met the established acceptance criteria; and (4) the

startup organization as well as the plant operating review committee

(PORC) had reviewed and accepted the test results.

The following test

packages were reviewed:

SIT-TP-650

Low Power Physics Testing

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SIT-1P-716

Core Performance Record

SIT-TP-717

CPC/COLSS Verification

The NRC inspectors detennined that each of the above test packages was

properly reviewed by the licensee and met the applicable acceptance

criteria.

No violations or deviations were identified.

8.

Survey of Licensee Responses to Selected Safety Issues

During this inspection period, the NRC inspectors reviewed the actions

taken by the licensee of safety issues identified in I. E. Information

Notices 83-75 and 84-06. The selected review included possible steam

binding of emergency feed water pumps and power operations with

mispositioned control rods.

The NRC inspectors reviewed each of the above safety issues to determine

whether the licensee had implemented procedures and provided training

programs to prevent, detect, and recover from a steam bound EFW pump and

mispositioned ccntrol rods.

In addition, the NRC inspectors compared the

Waterford 3 EFW system design with the designs at H. B. Robinson 2, Farley

1, and Surry 2 which have all experienced steam binding of the auxiliary

feedwater pumps.

The review revealed that the EFW system design at Waterford 3 incorporates

additional isolation valves between the EFW pump and main feedwater line

that are not used in the other auxiliary feedwater designs. Specifically,

the discharge of the EFW pump is isolated from the main feedwater line by

two check valves and two isolation valves, all in series.

The isolation

valves do not receive a signal to open until after the EFW pumps are

running. The licensee performs an operational test of the EFW pumps every

31 days in accordance with Surveillance Procedure OP-903-019, Revision 2

" Emergency Feedwater Flow Verification," to verify the operability of the

check valves.

In addition, the licensee performs Surveillance Procedure

OP-903-045, Revision 1. " Emergency Feedwater Flow Path Verification" every

31 days to ensure the isolation valves are properly aligned and free from

external leakage.

The NRC inspectors determined that the licensee's procedures for recovery

of a mispositioned control rod and verification of rod position, when one

form of indication is lost, have been adequately addressed in Off-Normal

Operating Procedure OP-901-009 Revision 3, "CEA or CEDMS Malfunction."

The licensee has also provided the operators with training in the proper

movement of control rods, and the consequences of improper movement and

operating with a mispositioned control rod.

No violations or deviations were identified.

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9.

Site Tour

At various times during the course of this inspection period the NRC

inspectors conducted general tours of the reactor auxiliary building,

turbine building, and reactor building to observe ongoing maintenance and

testing.

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On May 15, 1985, while on tour in the reactor auxiliary building, the NRC

inspector found Fire Door 0-221 to the Boric Acid Concentrator Room B had

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been removed from service. After further investigation, it was found that

LP&L Fire Protection Procedure FP-1-015, Revision 1, " Fire Protection

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System Impairments," paragraph 4.2, required in part that the shift

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supervisor / control room supervisor (SS/CRS) evaluate the fire impairment

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impact of Technical Specifications and insure that applicable ACTION

statement are complied with.

In addition, the SS/CRS is required to

complete a fire appliance impairment form and note the same in index log.

Contrary to the above statement, the NRC inspector could find no evidence

of a completed fire appliance impairnent form or that operability of the

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fire detectors on at least one side of the inoperable assembly had been

verified. Thisisaviolation(382/8516-03).

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10. Cpen Items

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Three violations have been identified in this report.

8516-01

Control of design changes

Paragraph 3

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8516-02

Failure to close boron

Paragraph 4

managenent valve during BAC

operation

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8516-03

Failure to follow procedures

Paragraph 9

for evaluation of removal of

fire doors

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11. Exit Interviews

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The NRC inspectors met with the licensee representatives at various times

during the course of the inspection.

The scope and findings of the

inspection were discussed.

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