ML20140E298

Insp Rept 50-458/85-81 on 851201-31.Violations Noted:Failure of Document Control Program & Inadequate Retest Following Mods or Repair.Deviation Noted:Failure to Verify Proper Diesel Generator Load Sequencing

ML20140E298
Person / Time
Site:	River Bend
Issue date:	01/22/1986
From:	Chamberlain D, Jaudon J, William Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20140E254	List: ML20140E249 ML20140E270 ML20140E285 ML20140E298
References
50-458-85-81, NUDOCS 8602030221
Download: ML20140E298 (9)

See also: IR 05000458/1985081

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APPENDIX C

U. S. NUCLEAR REGULATORY COMMISSION

REGION IV

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NRC Inspection Report: 50-458/85-81 License: NPF-47

Docket: 50-458

Licensee: Culf States Utilities Company (GSU)

P. O. Box 2951

Beaumont, Texas 77704

Facility Name: River Bend Station (RBS)

Inspection At: River Bend Station, St. Francisville, Louisiana

Inspection Conducted: December 1 through December 31, 1985

Inspectors: R LD , A W /~b'bb

D. D. Chamberlain, Senior' Resident Inspector Date

(pars. 1, 2, 3, 4, 5, 6, 7, and 8)

bb wn

W. B. Jones, Resi,Jent Inspector

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Date '

(pars. 2, 3, 4, 5, 6, 7, and 8,M

Approved: _

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J .' . J udon, Chief, Project Section A Da'te

P2act r Prt ects Branch

0602030221 060129

PDR ADOCK 05000450

0 PDR

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Inspection Summary

Inspection Conducted December 1 through December 31, 1985 (Report 50-458/85-81)

Areas Inspected: Routine, unannounced inspection of licensee event report (LER)

review, startup test procedure review, startup test witnessing, startup test

program quality assurance review, operational safety verification, and site tours.

The inspection involved 208 inspection-hours onsite by two NRC inspectors.

Results: Within the areas inspected, two violations were issued in the areas of

startup test witnessing and operational safety verification (failure of document

control program and inadequate retest following modifications or repair,

paragraphs 4.A and 6, respectively). In addition one deviation was issued in

the area of startup test procedure review (failure to require verification of

proper diesel generator load sequencing, paragraph 3).

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DETAILS

1. Persons Contacted

Principal Licensee Employees

• R. E. Bailey, Supervisor, Quality Concern

• C L. Ballard, Projects Supervisor

• C, Banks, Security

• W. H. Cahill, Jr., Senior Vice President, River Bend Nuclear Group

E. M. Cargill, Superintendent, Radiological Programs

• R. P. Carter, Security

• T. C. Crouse, Manager, Quality Assurance (QA)

• D. L. Davenport, Supervisor, Plant Security

• J. C. Deddens, Vice President River Bend Nuclear Group

• Jan Evans, Stenographer

• C. E. Foster, Assistant Plant Security Supervisor

P. E. Freehill, Superintendent', Startup and Test

A. D. Fredieu, Assistant Operations Supervisor

D. R. Gipson, Assistant Plant Manager, Operations

• P. D. Graham, Assistant Plant Manager, Services

• E. R. Grant, Supervisor, Nuclear Licensing

R. W. Helmick, Director, Projects

B. D. lley, Licensing Engineer

• K. C. Hodges Supervisor, Quality Systems

R. Jackson, Shif t Supervisor, Operations

D. Jernigan, Engineer, Startup and Test

G. R. Kimell, Supervisor, Operations QA

• R. King, Engineer, Licensing

A. D. Kowalczuk, Assistant Plant Manager, Maintenance

T. Lacy, Shif t Supervisor, Operations

• W. H. Odell, Manager, Administrative

• T. L. Plunkett, Plant Manager

W. J. Reed, Director Nuclear Licensug

D. Reynerson, Director, Nuclear Plant Engineering

• F. L. Richter, Operations, QA

• C, G. Sprangers, Engineer, QA

R. B. Stafford, Director, Quality Services

• K. E. Suhrke, Manager, Projects

• P. F. Tomlinson, Director Operation QA

C. Warren, Shif t Supervisor, Operations

Stone and Webster

• B. R. Hall, Assistant Superintendent, Field Quality Control

R. L. Spence, Superintendent, Field Quality Control

The NRC senior resident inspector (SRI) and resident inspector (RI) also

interviewed additional licensee, Stone and Webster (S&W), and other

contractor personnel during the inspection period.

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• Denotes those persons that attended the exit interview conducted on

January 10, 1985. NRC resident inspector, W. B. Jones, and NRC security

inspector, R. A. Caldwell, also attended the exit interview.

2. Licensee Event Report Review

(Closed) License Event Report (LER) 458/85-08: Reactor Pressure Vessel

(RPV) Level Transient.

The NRC inspector reviewed the corrective actions taken by the licensee to

avoid future recurrence of the RPV level transient caused by opening RHR

'.'A" suppression pool valve (1E12*M0VF004A) before RHR "A" shutdown cooling

suction valve (1E12*liOVF006A) was completely closed.

The following corrective actions are complete:

. Station Operating Procedures (SOPS) contain the necessary cautions for

the operator.

. A caution statement has been added to Surveillance Test Procedure

(STP) STP-309-601 instructing the operator to ensure that the shutdown

cooling suction valve F006A is fully shut prior to opening suppression

pool suction valve F004A.

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. All plant operators have been informed of the incident as well as the

above procedural change.

. A yellow caution sign having the same warning as that added to

STP-309-601 was mounted on the control panel near the RHR system

suction valve switches as an aid to the operators.

. STP-309-601 has been revised to restore isolation valve (1E12*liOVF008

and 1E12*t10VF009) motor breakers to the closed position prior to

restoration of valves F006A and F004A.

. Operations has completed their review of STPs involving the RHR system

having the potential to cause a level transient. This review revealed

tnat appropriate precautions or detailed instructions exist that would

preclude the simultaneous opening of the F004A and F006A valves.

In addition, the licensee is presently performing an engineering evaluation

to determine if an interlock between the F004A and F006A valves is feasible

to prevent the F004A valve from opening before the F006A valve has closed.

The R1 will monitor this engineering evaluation as an open item

(458/8581-04).

This LER is closed.

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3. Startup Test Procedure Review

The SRI reviewed STP 1-ST-31, " Loss of Offsite Power," Revision 0, and a

draft Revision 1. Several comments were generated during this review and

the corments were discussed with startup test personnel. One comment

revealed that the procedure failed to implement a FSAR Chapter 14 commit-

, ment to verify proper sequencing of diesel generator loads during a loss of .

offsite power. This failure to implement a FSAR commitment was identified I

by the SRI as a deviation (458/8581-03). The licensee took inmediate

corrective action to add the verification of proper load sequencing in

Revision 1 to 1-ST-31 and to initiate a 100% review of other STPs to assure

implementation of regulatory requirements and commitments. This review has

been completed by plant staff compliance with an overview by operations QA.

All questions generated from the review are being addressed by startup and

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no acceptance criteria changes have resulted nor are any anticipated. All

questions will be resolved prior to performance of the respective startup

test. All required actions are being taken to respond to the identified

deviation and this deviation is closed with no further response from the

licensee required.

Except for the one deviation noted, procedure 1-ST-31 appeared to meet all

applicable regulatory requirements and commitments.

4. Startup Test Witness

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During this inspection period the SRI and RI witnessed startup testing

activities conducted under the startup testing program. The NRC inspectors

observed that: (a) personnel conducting the test were cognizant of the

test acceptance criteria, precautions and prerequisites prior to beginning

the test; (b) the test was being conducted in accordance with an approved

procedure and the test procedure was being used and signed off by the

personnel conducting the test; and (c) data was being collected and

recorded as required. The NRC inspectors witnessed the following startup

tests:

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1-ST-14 Reactor Core Isolation Cooling (RCIC)  ;

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1-ST-26 Safety Relief Valves (SRV)

1-ST-27 Turbine Trip and Generator Load

Rejection

1-ST-31 Loss of Offsite Power

The following observations were made during the performances of the above

startup tests.

a. 1-ST-14 Reactor Core Isolation Cooling:

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The NRC inspectors witnessed the performance of the RCIC system run in

the condensate storage tank (CST) to reactor vessel mode and RCIC pump

cold start at rated pressure on December 2 and 5, 1985, respectively.

Prior to beginning the RCIC system run on December 2, 1985, the NRC

inspectors reviewed S0P-0035, " Reactor Core Isolation Cooling," Revi-

sion 0, located in the main control roca, which was being used in

conjunction with ST-14. The review revealed that "laters" located in

the procedure had been removed by Temporary Change Notice (TCN)

85-1287, however, this TCN was not posted at the beginning of SOP-0035

as required by Administrative Procedure ADM-003, " Development, Control

and Use of Procedures," Revision 7. Further review of S0P-0035

revealed that TCN-1287 also carrected a mistake in the electrical

line-up and added eight valves to the valve line-up sheet. This

matter was brought to the attention of the nuclear control operator

(NCO) stationed at the 601 control panel. On Decerbtr 3, 1985, the

NRC inspectors again reviewed 50P-0035 and noted thu'. TCN-1287 was not

posted in front of the procedure. This document control problem was

identified by the NRC inspectors as an apparent violation (458/8581-01).

The NRC inspectors then brought this condition to the attention

of the shift supervisor who immediately initiated actions to review

all the procedures in the control room for similar problems.

b. 1-ST-26 Safety Relief Valves:

The SRI and RI witnessed the performance of 1-ST-26, " Safety Relief

Valves," during this inspection period. This startup test has been

completed and the established acceptance criteria appears to have been

met. The completed test package will be reviewed during a future NRC

inspection.

No violations or deviations were identified in this area of

inspection,

c. 1-ST-27 Turbine Trip and Generator Load Rejection:

The SRI observed a turbine trip performed on December 23, 1985, for

startup test 1-ST-27 " Turbine Trip and Generator Load Rejection."

Reactor power was at approximately 9% when the trip was initiated and

when the turbine stop valves closed, both turbine bypass valves opened

to control reactor pressure. No reactor scram occurred and reactor

pressure was controlled as required.

No violations or deviations were identified in this area of

inspection.

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d. 1-ST-31 Loss of Offsite Power:

The SRI observed the performance of startup test 1-ST-31, " Loss of

Offsite Power," on December 6,1985. Reactor power was above 10% for

the test and when the loss of offsite power was initiated the

Division I, II and III diesel generators started and carried

safety-related electrical loads as required. The loss of offsite

power event was terminated as required by the procedure and no safety

relief valves lifted and reactor water level did not drop to the

point of initiating the high pressure core spray pump during the

event. After termination of the event, the reactor core isolation

cooling system was used to restore reactor water level to normal. An

initial review of the test data revealed that all control room

chillers did not start as required by the procedure. Licensee

investigation of the failure of all chillers to start revealed that

one standby diesel loaded approximately 3 seconds earlier than the

other diesel which allowed chilled water flow to be established on

one division and the other division chillers did not start due to

chilled water flow already being established. It appears that the

chillers performed as' designed and the licensee initiated a review to

determine if all chillers are needed during a loss of offsite power

event. The preliminary indication, based on heat load calculations

performed by S&W engineering, is that only one chiller is required

during the first 20 minutes of a loss of offsite power event and the

operator can start other chillers after 20 minutes if they are

needed. The licensee will evaluate the present design of the chiller

start logic and initiate changes if required. The SRI will evaluate

the final disposition during the required review of the final test

package.

No violations or deviations were identified in this area of

inspection.

5. Startup Test Program Quality Assurance (QA) Review

During this inspection period the RI began a review of the 1kensee's QA

audit and surveillance program for operational activities conducted under

the startup and power ascension test program. This initial review revealed

no problems and the program review will be completed during a future NRC

inspection.

No violations or deviations were identified in this area of ir.spection.

6. Operational Safety Verification

The SRI and RI observed operational activities throughout the inspection

period and closely monitored operational events. Significant operational

activities observed included several attempted turbine rolls which were

stopped due to high vibration, a successful completion of turbine roll and

initial synchronization of the main generator on December 3,1985. The

main generator was on-line for a short time (approximately 30 minutes) and

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a reverse power trip was received. The problem was traced to a wiring

problem, which was corrected, and the generator was again synchronized on

December 4,1985, and maintained on-line with a generator load of

approximately 60 megawatts. The highest power level reached during this

inspection period was approximately 200 megawatts electrical. Also, on

December 31, 1985, a turbine trip occurred with reactor power at

approximately 20%. Subsequent analysis of the trip revealed that a power

to load unbalance signal was received which caused the turbine control

valves to close and resulted in a turbine trip and subsequent reactor

scram on high pressure. The power to load unbalance signal apparently

occurred due to the combination of a failed pressure transmitter and an

electrical grid upset from loss of a 500 kilovolt line. During this event

and other operational activities, operational staff actions were observed

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to be well coordinated and efficient and the plant responded as expected.

In addition to observing operational activities, the SRI reviewed licensee

generated condition reports during the inspection period. A review of two

condition reports (CRs 85-0559 and 85-0561) and the subsequent followup by

the SRI revealed that inadequate retest was performed following modifica-

tions/ repairs made on two separate occasions. These two occasions were the

change out of wiring to the back-up scram valves with no documented

continuity checks or functional test performed and a wiring change to the

shutdown cooling suction valve (E12*F008) with no valve stroking performed

upon completion of the wiring modifications. In both instances, no docu-

mented engineering-approved alternative testing was provided, and the func-

tion of the components was compromised by the modifications / repairs.

Subsequent review of this situation by the licensee revealed that both

conditions were complicated by either a design error or by unclear design

instructions. This design control problem is being documented by the

licensee QA organization.

A review by the SRI of the significance of the compromise of the function

of the components revealed that the back-up scram valves are mentioned in

the River Bend FSAR but they apparently are not taken credit for in any

accident analysis and the shutdown cooling suction valve was locked closed

per a commitment to the NRC and shutdown cooling operation was not compro-

mised while the wiring error was in place. However, the failure to perform

adequate retest on a safety-related/important to safety component repre-

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sents a serious breakdown in the QA program and this inadequate retest

following a modification or repair was identified by the SRI as an apparent

violation (458/8581-02).

7. Site Tours

The SRI and RI toured areas of the site during the inspection period to

observe general work practices and gain knowledge of the facility.

No violations or deviations were identified in this area of inspection.

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8. Exit Interview

An exit interview was conducted on January 10, 1986, with licensee

representatives (identified in paragraph 1). During this interview, the

SRI reviewed the scope and findings of the inspection.