Text

Insp Rept 50-458/86-20 on 860501-0615.Violation Noted: Failure to Follow Administrative Procedures for Issue of Temporary Change Notices & Failure to Implement Procedures to Maintain Safety Sys Drawing Configuration
	ML20203H770
Person / Time
Site:	River Bend
Issue date:	07/24/1986
From:	Chamberlain D, Jaudon J, William Jones; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20203H757	List: ML20203H752; ML20203H760; ML20203H770;
References
	TASK-1.G.1, TASK-TM 50-458-86-20, NUDOCS 8608050087
	Download: ML20203H770 (15)
	v • d • e

. _ __ _ . _ .

.

,

APPENDIX B

, U. S. NUCLEAR REGULATORY COMMISSION

<

REGION IV

NRC Inspection Report: 50-458/86-20 License: NPF-47

'

Docket: 50-458

Licensee: Gulf 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: May 1 through June 15, 1986

Inspectors: {

D. D. diamberlain, Senior Resident Inspector

'

Date

(pars. 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, 11 and 12)

% , ~

W. B.FJones, Resident inspector l

b'h$b

Date

(pars. 1,2,3,4,5 6,7,8 9, and 11)

/

'

Approved: _ _ /w MYl 72 Y/[

J./.Jfudon,Ghiel,ProjectSectionA Dafe /

Eeactor Proj& cts Branch

8608050087 860730

, PDR ADOCK 05000458

G PDR

't

. . , - - ,- r__- ,_-.r,_m..-,.m-m_ _,.,_..,m , _y _, ,. . - - , , _ _ _ , , . , . ,, . . - _

.

-2-

Inspection Summary

Inspection Conducted Iby 1 through June 15, 1986 (Report 50-458/86-20)

Areas Inspected: Routine, unannouncad inspection of licensee action on

previous inspection findings, status of operating license conditions,

Nuclear Review Board activities, startup test witness, safety system

walkdown, operational safety verification, maintenance witness, surveillance

witness, licensee plans for coping with strikes and status of IMI action

item.

Results: Within the ten areas inspected, two violations were idantified

(failure to follow administrative procedures for issue of temporary

change notices, paragraph 9, and failure to implement procedures to

maintain safety system drawing configuration, paragraph 6).

_- - __ - ____-_

-

o

-3-

DETAILS

1. P_ersons Contacted

Principal Licensee Employees

M. Arant, Technician, Ir.strumentation and Control (I&C)

R. J. Backen, Supervisor (Acting), Operations Quality

' Assurance (QA)

W. J. Beck, Supervisor, Reactor Engineering

W.E H. Cahill, Jr. , Senior Vice President, River Bend

Nuclear Group

E. M. Cargill, Supervisor, Radiation Programs

T. C. Crouse, Manager, QA

J. R. Cummings, Procedure Coordinator,

P. E. Freehill, Superintendent, Startup and Test

A. O. Fredieu, Assistant Operations Superviser

P. F. Gillespie, Senior Compliance Analyst

D. R. Gipson, Assistant Plant Manager, Operations

E. R. Grant, Supervisor, Nuclear Licensing

B. R. Hall, Supervisor, Plant Services,

R. W. Helmick, Director, Projects,

G. K. Henry, Supervisor, Electrical Engineering i

K. C. Hodges, Supervisor, Quality Systems

R. J. King, Licensing Engineer

A. D. Kowalczuk, Assistant Plant Manager, Maintenance

W. H. Odell. Manager, Administration

T. F. Plunkett, Plant Manager

S. R. Radebaugh, Assistant Plant Manager, Services

W. J. Reed, Director, Nuclear Licensing

D. Reynerson, Director, Nuclear Plant Engineering (NUPE)

N. Simpson, Technician I&C

M. H. Small, Acting Supervisor, Operations Quality Control (QC)

R. B. Stafford, Director, Operations QA

K. E. Suhrke, Manager, Projects

^ P. c. Tomlinson, Director, Quality Services

D. Williamson, Operations Supervisor

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

interviewed additional licensee personnel during the inspection period.

Denotes those persons that attended the exit interview conducted on

June 19, 1986. NRC Region IV Section Chief, J. P. Jaudon, NRC resident

inspector (RI), W. B. Jones and Nuclear Reactor Regulation (NRR)

licensing Project Manager, S. Stern also attended the exit interview.

.

-4-

2. Licensee Action on Previous Inspection Findings

a. (Closed) Violation (458/8569-01): Failure of design document control

program.

i s

This violation was a failure to post approved design changes against

the effected design documents and a failure to distribute design

change documents to document control stations. NUPE revised

procedures NUPE-AA-54 and 59 to provide more control for the posting

and routing of design change documents. A 100 percent audit of

s design change files was conducted, and all noted discrepancies were

-

?' corrected. Training was conducted on the revised procedural

requirements, and-subsequent quality assurance surveillances revealed

no recurrence of.the problem. The SRI reviewed the revised

-

. ,

procedu'res and,the"other corrective actions.

This, viol'ation is_ closed.

~

_ b. (Closed) Violation (458/8569-02): Improper use of a field change

, , notice.

'

NUPE issued procedure NUPE-AA-64, " Control and Approval of Field

Change Notices (FCN's)" which provides detailed instructions and

'

restrictions for the use of FCNs. The licensee had conducted an

audit of the design change files and the discrepancies noted had been

corrected. Training of NUPE personnel responsible for completing

FCN's had been completed.

This violation is closed.

i

c. (Closed) Violation (458/8604-01): Failure to control temporary

circuit alterations administratively. .

The licensee actions in response to this violation included: a

complete inspection of control room panels for unauthorized lifted

leads or jumpers; QC hold points included in electrical maintenance

work requests to inspect for proper restoration; implementation of a

main control room cabinet access and work monitoring program; the

change of control room panel locks; a maintenance procedure revision

serialized tagging of any lifted lead or jumper for accountability

and the temporary alterations program was suspended and replaced by

design modification request procedures. The SRI has monitored

licensee actions relative to temporary alterations, and the

additional controls appear to be effective.

! Ihis violation is closed.

d. (Closed) Violation (458/8581-01): Failure to maintain a controlled

copy of a temporary change notice (TCN) in front of the affected

. controlled procedure in the control room Station Operating

Manual (50M)

_ . _ _ _ -- _-

_

- _ _ . . . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ . __

.

,

-5-

'

<The licensee :took immediate corrective actions by performing a

~ departmental review of all station operating procedures (SOPS),

, , abnormal' operating procedures (A0Ps), and emergency operating

-

procedures (EOPs). . During the review,.the licensee identified

several SOP's with duplicate copies of the same TCN and a few TCN's

filed with the wrong S0P. These conditions were immediately

corrected.'

,

m Responsibility for maintaining and ensuring that updates to the Main

' Control- Room procedure manual are properly filed has been reassigned

to Station Document Control (SDC). Periodic reviews of the Main

Control Room SOMs are being conducted by the SDC in accordance with

administration procedure ADM-005, " Station Document Control,"

Section 6.6.

'

_

This violation is closed.

3. Status of Operating License Conditions

Facility Operating License NPF-47 for River Bend Station was issued on

November 20, 1985, and Attachment 1 to this license contains items which

must be completed to the satisfaction of NRC Region IV. The following

status is provided for the Attachment 1 license conditions:

a. (Closed) License Condition 1.a.: Verify the station electric

distribution voltage analyses are in accordance with the guidelines

of Branch Technical Position PBS-1, Position 4, prior to completion

of the initial test program.

GSU has completed special situation test 1-SST-6, ." Bus Load Test,"

and the results were provided to Stone and Webster (S&W) for

comparison to analytical model results. Memorandum S-CRB-9031 dated

June 4, 1986, summarizes the results of that comparison and indicates

that the test versus analytical results are acceptable with no test

voltage drops more than 3 percent lower than the analytical values as

recommended by Branch Technical Position PSB-1.

This license condition is closed.

b. (Closed) License Condition 1.b: Evaluate and complete modifications

on battery powered lighting systems used in areas of the plant

outside the main control room required for safe shutdown and

personnel evacuation prior to completion of the initial test program.

The RI reviewed the licensee's emergency lighting plan as detailed on

Stone and Webster Lighting Plan drawings 12210-EE-65 thru 79.

Emergency lighting stations were selected from the drawings and

verified to be installed and operational for areas identified in

Final Safety Analysis Report (FSAR) Table 9.5-2, " Illumination Level

and Type of Fixtures used in Plant Areas Necessary for Safe Shutdown

and Evacuation of Personnel." In addition, areas previously

__

,

, -

.

,

. . . .

-

.. , -

. .

,

-6-

,

identified as being deficient in illumination were selectively

observed to meet the requirements of Table 9.5-2. As a result of

this walkdown, one area was identified which did not meet the minimum

illumination requirements. The area identified was an egress

stairwell located on the east side of the turbine building between

the elevation 95'0" and 123'6". The licensee initiated modification

request (MR)86-162 and maintenance work request (MWR) 41124 to

install the emergency light. This work was completed on June 15,

1986.

This license condition is closed.

4. Nuclear Review Board Activities

The RI reviewed the Nuclear Review Board (NRB) minutes for the period

February 1985 thru December 1985, to assess the overall effectiveness of

the licensee's implementation of the off-site review committee. These

minutes were evaluated against the NRB responsibilities outlined in

Section 6.5.3 of the Technical Specifications (TS) and the NRB Manual.

The RI noted during the above review, that the NRB has chartered four

subcommittees to assist the NRB in fulfilling their responsibilities.

These subcommittees are:

o "Unreviewed Safety Questions Committee" (USQC);

o " Quality Assurance Program Audit Committee" (QAPAC);

o " Quality Concern Subcommittee;" and

o "NRB/FRC Committee"

The USQC was established to assist the NRB in meeting its responsibilities

for reviewing proposed changes to the plant and its documentation to

ensure that changes are not made which constitute an unreviewed safety

question. Specifically, the USQC will review:

o all safety evaluations for changes to procedures, equipment, systems

or experiments which were determined not to involve unreviewed safety

questions;

o selected procedures, equipment, systems, tests and experiments which

did not receive an evaluation to verify that they were properly

classified and did not require a safety evaluation;

o changes which were determined to be unreviewed safety questions and

the associated changes to licensing documents;

o. proposed changes to the Operating License or TS; and

o violations of codes, regulations, orders, TS license requirements,

procedures and instructions having nuclear safety significance.

-

.

-7-

The QAPAC was chartered to advise the NRB on the effectiveness of the

'

Quality Assurance Program. This is accomplished thru the QAPAC's

participation in and review of audits performed by the QA audit group for

the areas required by 6.5.3.8 of the TS. The requirement of the QAPAC to

audit the Operational Quality Assurance Program every 24 months for

compliance with 10 CFR part 50, Appendix B, is being fulfilled thru the

Joint Utility Audit Group (JUAG).

The Quality Concern Subcommittee receives and coordinates quality concern

inquiries dealing with the operation of the quality assurance department

and management of the River Bend Nuclear Group. This committee insures

that proper action is taken by all departments to allegations or inquiries

which are received.

The NRB established the NRB/FRC Committee to monitor the FRC's activities.

This Committee assesses the FRC's fulfillment of its responsibilities by:

o reading all FRC minutes and other reports issued by the FRC;

o occasional attendance at FRC meetings;

o occasional verification of an FRC review item; and

o semi-annual meetings to discuss FRC activities.

The above subcommittees chartered by the NRB appear to be adequate to meet

the function and responsibilities established in the Technical

Specifications for the licensee's off-site review committee. The NRB

meeting' minutes reviewed by the RI for the period February 1985 thru

December 1985 demonstrate that the NRB members are cognizant of their

responsibilities and have established programs to fulfill these

responsibilities. The NRC inspectors will evaluate the effectiveness of

the NRB during future inspections.

No violations or deviations were identified in this area of the

inspection.

5. Startup Test Witness

During this inspection period, the SRI and RI witnessed startup testing

conducted under the startup testing program. The NRC inspectors observed

that: personnel conducting the test were cognizant of the test acceptance

criteria, precautions and prerequisites prior to beginning the test; the

test was conducted in accordance with an approved procedure; the test

procedure was being used and signed off by the personnel conducting the

test; and data were being collected and recorded as required. The NRC

inspectors witnessed the following startup tests:

'

.

-8-

o 1-ST-27 Turbine Trip and Generator Load Reject

o 1-ST-28 Shutdown from Outside the Control Room

o 1-ST-25B Main Steam Isolation Valve (MSIV) Full

Closure

o 1-ST-19 Core Performance

The following observations were made during the performance of the above

startup tests:

o Test 1-ST-27 Turbine Trip and Generator Load Reject:

The SRI and RI observed the performance of Section 6.3, "High Power

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

Generator Load Rejection" on May 29, 1986. The reactor was at

approximately 96 percent thermal power when a generator load

rejection was initiated by tripping a generator differential relay.

This caused both generator output breakers to open and a turbine

control valve fast closure (TCVFC) to occur. The reactor scrammed,

as expected, when the TCVFC signal was initiated. The peak reactor

pressure reached during the transient was 1106 psig. The NRC

inspectors noted that the bypass valves opened along with nine safety

relief valves (SRV) to reduce and control reactor pressure.

Following the initial SRV blowdown, only 1 SRV was observed to

reopen, which is consistent with the test acceptance criteria. The

licensee is presently evaluating SRVs B21*F041D and B21*F041F which

apparently opened briefly in their safety mode. The results of this

evaluatior, will be reviewed by the NRC inspectors during their review

of 1-ST-27 test results. The licensee has collected test data to

evaluate for conformance to acceptance criteria.

No violations or deviations were identified in this area of

inspection.

o 1-ST-28 Shutdown from Outside the Control Room:

The SRI and RI observed the performance of Section 6.3, " Cold

Shutdown from Outside the Control Roem", on May 30, 1986. Following

the reactor scram initiated during 1-ST-27, reactor pressure was

reduced to 120 psig from the main control room (MCR). The licensee

had previously demonstrated the ability to scram and maintain the

reactor in hot shutdown from outside the control room on February 15,

1986. With reactor pressure at 120 psig, control of the Division I

residual heat removal (RHR) system was transferred from the MCR to

the remote shutdown panel (RSP). When the nucleer control operators

(NCO) began to realign the RHR system from the low pressure coolant

injection (LPCI) mode to the RHR mode, the suppression pool suction

valve 1E12*F004A indication failed in the intermediate position.

- _

. . _ . ._. . - - __

. _ _ _ . _ . . _ _ . . . __ _ _ . _ _ _ _

.

-9-

l

After verifying locally that 1E12*F004A was closed, the NCO attempted

to open 1E12*F006A valve as required to establish shut down cooling.

This attempt failed however because of the interlock that prevents

4 -the F006A valve from opening if the F004A is not in the closed

. (indicated) position. Control of the division II RHR system was then

transferred to the division II RSP and the system aligned to the shut -r

down cooling mode without incident. The NCO established a cooldown

rate of approximately 80 F per hour with the heat transfer path to

the' standby, service water system. The licensee has collected the

test data to evaluate for conformance to acceptance criteria.

,

'

% violation or deviations were observed in this area of inspection.

'

,

.

,

-

,o' '

.T,est 1-ST-25B Main Steam Isolation Valve (MSIV) Full Closure:

.

2

,

Th'e RI witnessed the pefformance of startup test 1-ST-25B, "MSIV Full

Closure", on-June 8,1986 with reactor power at 100 percent and rated

., . core f, low at 96 percent. The licensee initiated the test at

1733 hours0.0201 days 0.481 hours 0.00287 weeks 6.594065e-4 months .by simulating a loss of condenser vacuum which results in

the MSIVs closing when in the run mode. Upon initiation of the MSIV

full closure, the reactor tripped and the SRVs opened momentarily to

control pressure. Following the initial opening of the SRVs, only

'

, 1 SRV was observed to reopen. The high pressure core spray (HPCS)

and reactor core isolation cooling (RCIC) systems initiated on

reactor vessel water level reaching the level 2 setpoint. The

reactor vessel water level recovered quickly because of the

subsequent swell of vessel water and the continuous feed from the

reactor feed water pumps. The NC0 secured the HPCS injection valve

prior to the system injecting based on the rising vessel water level. ,

Approximately 3 minutes into the transmit, the reactor feed pumps

tripped on vessel high level. The reactor feed water B pump was

subsequently restarted and vessel level maintained within the normal

band. Control room conduct during the test was observed to be well

coordinated and efficient. The licensee has collected the test data

to evaluate for conformance to acceptance criteria.

"

No violations or deviations were identified in this area of

inspection,

i

! o Test 1-ST-19 Core Performance

The SRI witnessed the running of traversing incore probe (TIP) traces

in preparation for computer calculation of reactor power / thermal

limits. The required data was extracted and inserted into startup

test procedure 1-ST-19, " Core Performance" by the licensee for test

condition (TC) 6 verification of core thermal limits and core thermal

i power. A preliminary review of the test data by the SRI revealed

i-

that core thermal limits were well within the TS limits and core

thermal power was approximately 99.3 percent of rated thermal power.

1

.

i

. _ _ _ _ , . _ . , , , , _ , m . _ , _ _ _ _ . _ . _ . . .,.-,,...,_,_...,,_,_,,,.,,m._,,,_,,_._,_, -

,m.__~_. .m-- - _-

l

-

. ,.

-10-

L

~

,

No violations or deviations were identified in this area of

. inspection.

,

6., Safety System Walkdown.

l'Duringthisinspebtionperiod,theSRIandRIperformedawalkdownofthe

'

"C",RHR system to verify proper system alignment for operability as

required by TS for Operational Conditions 1, 2, 3, 4, and 5. It was

observed that:

,

o sy' stem valves were properly aligned;

o abnormal control-room instrumentation readings or alarms were

-

present;

o no leakage from major components was present;

o the "C" RHR pump upper and lower bearing oil reservoirs were properly

filled; and

o accessible hangers and supports were intact.

No anomalies were noted that would have affected "C" RHR system

operability for low pressure coolant injection (LPCI). However, certain

discrepancies were noted when comparing the system condition / lineup with

the engineering piping and instrument diagrams (PIDs). It was noted that

pipe caps were missing from five vent and drain locations where PID-27-7C

indicated caps were installed; valve E12*MOVF064C which was open showed

closed on PID-27-7C; and valve E12*VF063C was not locked as shown on

PID-4-3C. Also, several valves were locked closed although the drawings

did not indicate locks installed. The failure to implement procedures to

maintain a safety system condition / lineup as shown on design output

drawings or to modify the design output drawings to reflect the required

system condition was identified by the SRI as an apparent violation

(458/8620-02). Subsequent discussions with licensee management revealed

that minimum flow valve E12*MOVF064C is open for standby operation and

condensate fill and flush valve E12*VF063C (not a major flow path valve)

had been identified as unlocked during an operations review of all valves

on the locked valve list. It had been noted that the valve was unlocked

and inaccessible without a scaffold or ladder. No action had been taken

to gain access to the valve to install a lock. The licensee took

immediate action to obtain ladders, and the valve was locked as shown on

PID-4-3C when the SRI identified that the valve was not locked. It was

also noted that the system operating procedure valve lineup did not

indicate that valve E12*VF063C was to be locked closed as show on the PID.

In response to the identified violation, the licensee should address how

they will assure that PID drawing, Stone and Webster flow diagrams, system

operating procedure valve lineups, locked valve lists and actual system

configuration are consistent for an identified system operational

condition and they should identify procedural controls which allow

deviation from drawing requirements. For example, procedures may allow

-

,

s a--

..

,

.

-11-

the locking of more valves than shown on the drawings at the discretion of

the operations staff. This issue should also be addressed for other

drawings / documents used routinely by operations and/or maintenance for

performance of work activities.

7. Operational Safety Verification

The SRI and RI observed operational activities throughout the inspection

period and closely monitored operational events. Control room activities

and conduct were observed to be well controlled and efficient. Proper

control room staffing was maintained and access to the control room

operational area was controlled. The licensee was adhering to limiting

conditions for operation (LCO) as they occurred. Operators were

questioned regarding lit annunciators and they understood why the

annunciators were lit in all cases. Selected shift turnover meetings were

observed, and all necessary information concerning plant status was

apparently being covered in these meetings. A walkdown of the "C" RHR

system was conducted, and the valves were observed to be in the proper

position for standby operation. Several plant tours were conducted and

overall plant cleanliness was good. During these plant tours, radiation

protection area postings were observed to be accurate.

During this inspection period the licensee completed a 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months

verification run at full power operation, and all planned startup testing

was completed.

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

8. Maintenance Witness

During this inspection period, the RI observed portions of selected

corrective maintenance activities to verify that maintenance activities

are being conducted in accordance with approved procedures, TS and

appropriate industrial codes and standards. The RI verified through

direct observation and review of records that:

o maintenance activities did not violate LCOs;

o redundant components were available;

o required administrative approvals and tagouts were obtained before

initiating work;

o procedures were adequate to control the work;

o radiological controls were properly implemented where applicable;

o QC hold points were established and observed; and

o replacement parts and materials used were properly certified.

, o

-12-

The_following two corrective maintenance activities were observed:

o Control Rod Drive (CRO) Cooling Water Check Valve:

On May 7, 1986, the licensee experienced a failure of control

rod 24-33 to insert or withdraw during control rod manipulations.

Trouble shooting of the hydraulic control unit (HCU) revealed that

CRD cooling water check valve C11-V138 was not reseating when drive

water was applied to HCU 24-33. This condition allowed the drive

water to flow back thru the cooling water line and thus the necessary

lift was not being provided to the CRD drive piston to insert the rod

or retract the collet finger to allow withdraw of the control rod.

This condition would not have prevented the control rod from

inserting during a reactor scram.

~The. licensee initiated prompt maintenance work request (MWR) 39099 to

clean', inspect and replace if necessary the ball checks to CRD

cooling water check valve C11-V138 on HCU 24-33. The RI verified

prior to initiating work that the MWR had been properly initiated; QC

notification points were established, the job plan was appropriate to

control the work; and a job briefing had been performed as evidenced

by maintenance personnel signatures on the job briefing sheet. In

addition, the RI verified the requirements of T.S. 3.1.3.1 for an

inoperable control = rod were being complied with.

Prior to in'itiating work, the nuclear equipment operator (NE0)

obtained the shift supervisors permission to isolate HCU 24-33. The

, ' HCU was then isolated and tagged out using clearance number 86-192,

and.the ball was removed from the check valve. Inspection of the

ball revealed surface scratches and abrasions. Subsequent flushings

of the cooling water line produced a 1/2" x 1/16" round metal sliver.

'

A new ball was then installed in the check valve and the system

. verified operable using surveillance test procedure STP-052-0101,

" Control Rod Movement Operability Check." LC0 86-394, which was

initiated to track the action requirements of T.S. 3.1.3.1 was closed

based on the acceptable performance of STP-052-0101.

No violations or deviations were identified in this area of

inspection,

o Division II Emergency Diesel Generator:

During the performance of surveillance test STP-309-0202," Diesel

Generator Division II Operability Test," on June 5,1986, the " ready

to load" light did not illuminate after the diesel generator achieved

rated frequency and voltage. The licensee noted this condition and

immediately shut down the diesel. The failure of the " ready to load

light" to illuminate indicates that the standby generator breaker

1 ENS *ACB27 would not have closed onto standby buss IENS*SW618.

Prompt MWR 41549 was then initiated to trouble shoot and restore the

ready to load circuit to operable status. The RI reviewed the MWR

. *

.

-13-

prior to the initiation of work and verified that a deficiency tag

had been placed and the applicable LC0 initiated. The appropriate

hold points were placed in the procedure and a quality control

representative was present during the performance of this MWR. The

RI verified that the lifted leads were identified in the Lifted Lead

and Jumper Log as tag, numbers 86-3101-001 thru 006 and that the

, restored leads were independently verified.

The licensee identified-that relays EGS*UVRA-UVRB needed the pick up

voltages adjusted to within the setpoint limits. This work was

subsequently performed under MWR 41549, and the diesel restored to

operable statuu at 2041 hrs after the successful completion of

surveillance STP-309-0202. LC0 86-472 was then cancelled at

2045 hours0.0237 days 0.568 hours 0.00338 weeks 7.781225e-4 months .

No violations'or deviations were identified in this area of

inspection.

9. Surveillance Witness

The SRI and RI witnessed surveillance testing conducted by the licensee

during this inspection period and the following observations were made:

o Surveillance Test STP-051-4210:

The SRI witnessed a portion of the instrumentation

surveillance STP-051-4210, "RPS/RHR Reactor Vessel Steam Dome

Pressure - High, Monthly Chfunct, 18 Month Chcal, and 18 Month LSFT

(821-N078B,B21-6798)" conducted on May 28, 1986. The portion of the

test observed was known to cause a half scram signal and the

technicians were cautious in verifying the other division was not

tripped prior to test performance. They also limited the time that

the half scram was allowed to be in by close coordination with the

operations staff. During the performance of this test it was noted

that there were more than one copy of certain pages of the procedure

in the official work copy. This created some confusion during test

performance. The SRI discussed this with instrumentation maintenance

supervision and it was discovered that there were seven open TCNs

against this procedure and the preparer of the last TCN (No. 86-0581)

had failed to use copies of previous permanent TCN pages to markup

for the new TCN as required by administrative procedure. This

resulted in the official work copy of the procedure containing three

page eights and three page elevens with a different TCN number on

each page. This failure to follow administrative procedures for

issue of TCNs was identified by the SRI as an apparent violation

(458/8620-01).

The SRI also discussed the status of incorporation of TCNs in

procedure revisions in response to a previous NRC violation and it

was found that there are a total of 38 procedures with more than

three open TCNs out of a total procedure population of

. - - .. - .-- - .

,, .

-14-

approximately 3500. The SRI requested a date from licensee

management for revision of these 38 procedures to incorporate the

'open TCNs and management stated that these procedures would be issued

by September 1, 1986.

o Surveillance Test STP-309-0202:

The RI observed the performance of surveillance STP-309-0202, " Diesel

Generator Division II Operability Test," Revision 5 on June 5, 1986,

.with the plant in operational condition 1. This surveillance is

'

. designed to demonstrate the operability of the Division II diesel

,

generator and satisfie's T.S. Sections 4.8.1.1.2.a.1 through

i 4.8.1.-l.2.a.7;and 4.8.1.1.2.c.1 and c.2.

.

-

) Prior to' initiating the test, communications were established between

'

the NE0 at the remote diesel panel and the NC0 in the Main Control

i Room. LC0 86-472 had previously been initiated in accordance with

! T.S. 3.8.1.1.b because of the preplanned preventive maintenance (PM)

'

which had been performed on this diesel generator prior to beginning

'STP-309-0202. Upon initiation of the STP, the diesel generator was

observed to attain rated frequency and voltage within the required 10

seconds, however the " ready to load" light did not illuminate at

either the remote panel or in the main control room. The diesel

generator was shutdown within 30 seconds of its starting and prompt

MWR 41549 initiated. The failure of the " ready to load" light to

illuminate indicates that the standby generator breaker IENS*ACB27

would not have closed onto standby bus IENS*SW618. After verifying

the diesel generator had failed its surveillance test, the licensee

initiated the actions required by T.S. 3.8.1.1.b for an inoperable

diesel generator, because of any cause other than the performance of

preplanned preventive maintenance.

MWR 41549 was completed at 2015 hours0.0233 days 0.56 hours 0.00333 weeks 7.667075e-4 months on June 5, 1986, and

STP-309-0202 was satisfactorily performed. LC0 86-472 was then

closed based on the diesel generator satisfying the operability test.

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

i 10. Licensee Plans for Coping With Strikes

,

The SRI reviewed licensee plans for coping with strikes during this

inspection period. It was found that the licensee had addressed such

issues as personnel / training requirements, security support requirements,

>

offsite support, etc. This area will be reviewed further during future

NRC inspections as required.

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

1

!

. . . .

-15-

11. Status of TMI Action Item

(Closed) NUREG-0737, Item I.G.1: Trair;ing during initial startup test

. phase.

The licensee has completed the initial startup test program training

requirements as described in NUREG-0737, Item I.G.1 and FSAR

Section 14.2.3 Training During Initial Startup Test Phase. The startup

test training program was established to assure that personnel from each

of the six operating shift crews:

o observed a reactor scram;

o observed a pressure regulator transient;

o observed a water level setpoint transient;

o operated the operation of RCIC system; and

o observed a turbine trip or load rejection.

The startup test program was balanced, as much as practical, between the

six shif ts to assure that each shift was exposed to the above, off-normal

events. This has resulted in each shift having at least four individuals

who have observed actual plant responses for each of these events. Based

on the shift experience that now exists for each shift crew, no further

licensee action regarding NUREG-0737, Item I.G.1 is necessary.

This TMI action item is closed.

12. Exit and Inspection Interview

An exit interview was conducted on June 19, 1986, with licensee

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

SRI reviewed the scope and findings of the inspection.

ML20203H770

Text

Navigation menu

Search