ML20198H771

Insp Rept 50-458/85-77 on 851031-1130.Violation Noted: Failure to Provide Clear Design Instructions on Approved Change to safety-related Design Mod

ML20198H771
Person / Time
Site:	River Bend
Issue date:	01/10/1986
From:	Bennett W, Chamberlain D, Chanberlain D, Farrell R, Jaudan J, Jaudon J, William Jones, Mullikin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20198H734	List: ML20198H728 ML20198H754 ML20198H771
References
50-458-85-77, NUDOCS 8601310093
Download: ML20198H771 (13)

See also: IR 05000458/1985077

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APPENDIX B

U. S. NUCLEAR REGULATORY C0!DtISSION

REGION IV

NRC Inspection Report: 50-458/85-77 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: October 31 through November 30, 1985..

Inspectors: l ,

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D. D.VChamberlain, Senior Resident Inspector Date

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

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W. B. Jones,-Residdnt Inspector

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Date

(pars. 3,.4, 7, 8, 9, 10, and 11)

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DA l N

. ( Mu(liki , Reactor Inspector" ' Date

(par. 3)

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R. Farfell, enior Resident InTp6ctor Date

( ars. Q_and 10)

8601310093 860127

PDR ADOCK 05000458

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W. R. Bennett, Project Engineer Date

(pars. 2 and 10)

Approved: 62 //4 dM

, ~J/. P Jauff on, Chief,' Project Section A, Date

Me ob Projects Branch

Inspection Summary

Inspection Conducted October 31 through November 30, 1985 (Report 50-458/85-77)

Areas Inspected: Routine, unannounced inspection of licensee action on previous

ir.spection findings, status of facility operating license conditions, site tours,

control and use of field change notices, licensee quality concern program

review, initial criticality witness, pipe support and restraint systems testing,

human engineering discrepancies (HEDs) review, startup test procedures review

and startup test witness. The inspection involved 483 inspector-hours onsite by

five NRC inspectors.

Results: Within the areas inspected, one violation was issued in the area of

control and use of field change notices (failure to provide clear and concise

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design instructions on an approved change to a safety-related design

modification, paragraph 5),

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DETAILS

1. Persons Contacted

Principal Licensee Employees

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

• E. M. Cargill, Superintendent, Radiological Programs

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

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

• D. R. Derbonne, Supervisor, Startup and Test

• Jan Evans, Stenographer

S. Finnegan, Control Operating Foreman

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

• A. D. Fredieu, Assistant Operations Supervisor

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D. R. Gipson, Assistant Plant Manager, Operations

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

• E. R. Grant, Supervisor, Nuclear Licensing

D. Hartz, Shift Supervisor, Operations

• R. W. Heinick, Director, Projects

B. D. Hey, Licensing Engineer

K. C. Hodges, Supervisor, Quality Systems

R. Jacksor , Shift Supervisor, Operations

D. Jernigan, Engineer, Startup and Test

• G. R. Kimmall, Supervisor, Operations QA

R. King, Engineer, Licensing

A. D. Kowalczuk, Assistant Plant Manager, Maintenance

D. J. Krueger, Supervisor, Engineering Administration

T. Lacy, Shift Supervisor, Operations

• H. M. McClellan, Senior Compliance Analyst

• I. M. Malik, Supervisor Quality Engineering

A. Middlebrooks, Shift Supervisor, Operations

E. R. Oswood, Engineer, QA

G. A. Patrissi, Engineer, QA

• T. L. Plunkett, Plant Manager

S. R. Radebaugh, Assistant Superintendent, Startup and Test

• W. J. Reed, Director Nuclear Licensing

• D. Reynerson, Director, Nuclear Plant Engineering

• J. E. Spivey, Engineer, QA

• R. B. Stafford, Director, Quality Services

• P. F. Tomlinson, Director, Operation QA

C. Warren, Shift Supervisor, Operations

Stone and Webster

F. W. Finger, III, Project Manager, PTO

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

R. L. Spence, Superintendent, Field Quality Control

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The NRC inspectors also interviewed additional licensee, Stone and Webster

(S&W), and other contractor personnel during the inspection period.

• Denotes those persons that attended the exit interview conducted on

December 12, 1985. The NRC resident inspector (RI), W. B. Jones, also

attended the exit interview.

2. Licensee Action on Previous Inspection Findings

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a. (Closed) Open Item (8425-02): Wetting of neutron monitoring system

cables below the reactor pressure vessel.

The NRC inspector reviewed the licensee's records of megger testing

and continuity testing of neutron monitoring system cables located

below the reactor pressure vessel. All of the testing was witnessed

by licensee quality assurance (QA) engineers.

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This item is closed.

'b. (Closed) Open Item (8558-04): Procedures identified for performance

of surveillance testing of Emergency Core Cooling Systems (ECCS) were

not approved for use.

The NRC inspector reviewed the licensee's surveillance test

procedures (STPs) for ECCS systems and determined that they were

adequate to meet Technical Specification (TS) requirements.

This item is' closed.

i (Closed) Open Item (8558-05): Procedures that were similarly

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performed contained inconsistencies and errors existed in procedures

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that had not been walked down.

The NRC inspector reviewed the STPs for ECCS systems and determined

that they were consistent and adequate to meet requirements.

This item is closed.

d. (Closed) Open Item (8558-06): Procedures contained "LATERS" and did

not meet requirements of TS 4.6.1.2.a.

The NRC inspector reviewed STP 000-0702 and determined that it meets

the requirements of TS 4.6.1.2.a.

This item is closed.

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e. (Closed) Open Item'(8558-08): Procedures identified to meet the

requirements of TS 4.6.4.2 were not approved for use.

The NRC inspector reviewed the STPs for TS 4.6.4.2 and determined

that they are adequate to meet TS requirements.

" This item is closed.

f. (Closed) Open Item (8558-26): Changes to TS have not been

implemented into STPs.

The NRC inspector reviewed the licensee's reference tracking system

~for implementing changes to TS into procedures and determined that

the program is adequate to ensure that procedures meet TS

requirements.

This item is closed. ,

g. =(Closed) Open Item (8558-28): Procedure STP 055-0101 (dated

July 22, 1985) is not responsive to TS 4.9.12.4 in that it'does not

verify access. interlocks and palm switches in certain locations are

operable as required.

The NRC inspector reviewed STP 055-0101 (dated September 23, 1985)-

and determined that it is adequate'to meet.the. requirements of

TS 4.9.12.4.

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This item is closed.

h. (Closed) Open Item (8527-04) (License Condition 3.a. Part 4): This

item involved the completion of procedure A0P-0031, " Shutdown From

Outside the Main Control Room," and the training of required

personnel in the procedure.

The NRC inspector reviewed a copy of approved procedure AOP-0031 and

performed a walkdown of this procedure. The procedure appeared

adequate to provide the necessary instructions to personnel to

remotely shutdown the plant due to a fire in the main control room.

All required actions were verified as being able to be performed in-a

timely manner. In addition, the required training of operators has

been performed.

lhis open item and part 4 of license condition 3.a. are closed,

i. (Closed) Open Item (8527-06) (License Condition 3.a. Part 3): This

item involved the completion of all modifications, including diesel

generator electrical circuits, to isolate the control room circuitry

from the safe shutdown panel.

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The NRC inspector reviewed the results'of the following

preoperational tests: (1) Special Situation Test Procedure 1-SST-50,

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" Remote Shutdown," Revision 0 and (2) Special Situation Test

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Procedure 1-SST-59, " Division I Remote Shutdown Tests," Revision 0.

These tests were found to be complete and acceptable.

This open item and part 3 of license condition 3.a. are closed.

j. (Closed) Open Item (8527-08) (License Condition 3.a. Part 4): This

. Item concerned the possible need for radio communications between

operators during the implementation of A0P-0031.

A review of the procedure and a walkdown by the NRC inspector.

revealed no areas where remote communications were required.

This open item and all parts of license condition 3.a. are closed.

3. Status of Facility Operating License Conditions

Facility Operating License NPF-40 for River Bend Station was issued on

August 29, 1985, and pending Commission approval, operation is restricted

to power levels not to exceed 5% of rated power. Attachment 1 to this

license contains items to be completed to the satisfaction of NRC Region IV

prior to achieving certain operational conditions. The following status

is provided for the Attachment I license cond; Lions:

a. (Closed) License Condition 3.a.:

Complete the fire protection and prevention items prior to exceeding

5% rated power.

Parts 1 and 2 of this license condition were closed in NRC Inspection

Report 50-458/85-69. Parts 3 and 4 of this license condition were

closed in paragraph 2 of this report along with open items 8527-04,

8527-06,and 8527-08.

All four parts of license condition 3.a. are closed.

b. (Closed) License Condition 3.b.:

Complete testing of liquid, gaseous, and solid radwaste systems and

place these systems in service prior to exceeding 5% rated power.

The liquid radwaste system was addressed in paragraph 3 of NRC

Inspection Report 50-458/85-70. The gaseous radwaste system was

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addressed in paragraph 3 of NRC Inspection Report 50-458/85-69.

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The solid radwaste system was addressed in paragraph 2 of NRC

Inspection Report 50-458/85-53. All of these systems have been tested

and are operational.

This license condition is closed.

c. (Closed) License Condition 3.d.:

Complete installation and testing of post-accident sampling system

(PASS) and place system in service prior to exceeding 5% rated power.

On November'10, 1985, the RI witnessed a demonstration of the

licensee's ability to provide representative liquid samples from

within the primary containment using the manual grab post-accident

sampling system. The demonstration consisted of taking two liquid

samples from the B jet pump header using the liquid sample station

located in the auxiliary building. The licensee was able to draw two

samples and analyze the first sample for pH, dissolved oxygen,

hydrogen, boron, conductivity and radionuclides within the required 3

hours from the time the decision was made to take the samples. The

RI verified that approved procedures.were used for the PASS demonstra-

tion and sample analysis.' In addition, the RI verified that work on

PASS has been completed and the system turned over to operations.

This license condition is closed.

d. ,(Closed) License Condition 3.e.:

Load reduction modifications to reduce maximum emergency service load

to 2884 kw for the Division I diesel generator and to 2780 kw for the

Division II diesel generator prior to exceeding 5% of rated power.

The licensee initiated Modification Request (MR) 850291 to reduce the

loads on the Division I and II emergency diesel generators. As a

result of this request, S&W engineering company initiated Engineering

and Design Coordination Requests (E&DCR) S-10017 and S-10018. E&DCR

S-10017 adjusted the fan blade settings on Division I and II diesel

generator exhaust fans 1HVP*FN2A and B, respectively. E&DCR S-10018

reduced the heating capacity of the standby cooling tower remote air

intake room duct heaters 1HVY*CH6A and B from 30 kw to 12 kw egh.

These modifications were performed under Maintenance Work Requests

(MWR) 003497, 008678, and 14680. MWRs 003497 and 008678 adjusted the

fan blade pitches from 37 to 20 for Division I and II diesel

generators respectively, while MWR 14680 disconnected 9 of the 15

heater elements from HVY*CH6A and 68. These modifications resulted

in final diesel generator loading of 2882.84 kw for Division I and

2766.01 kw for Division II. The licensee has also submitted a Final

Safety Analysis Report (FSAR) change notice to the NRC reflecting the

above modifications.

This license condition is clor,ed.

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4. Site Tours

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

gain knowledge of the plant and to observe general job practices.

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

5. Control and Use of Field Change Notices (FCNS)

During a routine followup from a previous NRC inspection, the SRI reviewed

the completed Field Change Notice (FCN) 19 to MR 85-0397. FCN-19 had

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become a 96-page document and the SRI found that the design instructions

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were often misleading and difficult to follow. For example, page 36 of

FCN-19 added terminal board 43B contacts (contacts 7/8 and 3/4) on an

electrical schematic for remote shutdown operation of the diesel generator

output breaker, and page 57 of FCN-19 removed the same contacts. Also,

l several pages of FCN-19 were found to be illegible due to poor reproduction

quality. The failure to provide clear and concise design instructions on

. an approved change to a safety-related design modification was identified

by the SRI as an apparent violation (8577-01). The SRI immediately

informed the licensee of this finding and the licensee took immediate

action to review other safety-related design modifications for similar

problems. Although similar documentation problems were found, hardware

inspections revealed no problems with the installed hardware.

6. Licensee Quality Concern Program Review

The licensee's quality concern prouram was described in NRC Inspection

Report 50-458/84-25. The SRI continued to review licensee actions on

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individual cases and completed the review of the current status of all

open quality concerns,

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

7. Initial Criticality Witness

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l This area of inspection was conducted to review licensee preparations for

initial critical and to witness initial criticality in order to ascertain l

l confomance to license and procedural requirements and to observe

l operating staff performance. The licensee completed preparations for

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initial criticality on Octooer 30, 1985, and began pulling the first

control rod at about 9:30 p.m. The reactor was declared critical at

12:38 a.m. on October 31, 1985, on step 27 of the rod withdrawal sequence

with 2032 notches withdrawn. Nuclear instrumentation was monitored closely

during approach to criticality and appeared to respond as expected.

Also, during this inspection period, the RI completed the review of the

results of 13 STPs performed prior to achieving initial criticality. The

RI verified that the STP had been properly signed by the personnel

performing the test, that the data was collected and recorded as required,

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that all deficiencies had been addressed and corrective action completed,

and that the surveillance tests had been performed within the required

frequency.

The. SRI concluded that the licensee preparations for initial criticality

were cautious and the operating staff performance was controlled and

cautious. License and procedural requirements appeared to have been

implemented for a safe and cautious initial criticality at River Bend.

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

8. Pipe Support and Restraint S_vstems Testing

During the initial nuclear heat up cycle, the R1 visually examined system

pipe supports to ascertain whether system thermal expansion vrould result

in any apparent failures to the dynamic pipe supports, fixed pipe supports,

or their associated component support structures. These visual examirationr.

were performed at ambient plant temperature and again at a plant

temperature of 350 F. Pipe supports and structures were examined to

verify where applicable, that; (a) deterioration, corrosion, physical

damage or deformation were not noticeable; (b) bolts, nuts, washers, were

tight and secure; (c) the equipment was not in a " lock up" or " frozen"

position; and (d) pipes, supports or other associated equipment or

compcnents were not in contact or cause rubbing due to thermal expansion.

The following pipe supports and associated support structures were examined:

Type System Identification No.

Spring Residual Heat Removal RHS-PSSH-30775

Spring Recirculation Pump Housing H-306A

Spring det Pump Header H354A

Spring Low Pressure Core Spray CSL-PSSH-3001

Spring Low Pressure Coolant Injection RHS-PSSH-3089

Snubber Feed Water FWS-PSSP-3021

Snubber Residual Heat Removal RHS-PSSP-3078

Snubber det Pump Header S-354A

Snubber det Pump Header S-3598

Snubber det Pump Header S-304B

Snubber Residual Heat Removal Rl15-PSSP-3011

Snubber Recirculation Pump Motor S-371B

Snubber Recirculation Pump Housing S-306A

Snubber Feedwater FWS-PSSP-3015

Hanger Feedwater FWPRR-804

11 anger Feedwater FWS-PRR-823

Hanger Feedwater FWS-PRR-808

Hanger Residual Heat Removal RHS-PRR-831

Bracket Main Steam MSS-814

Bracket Main Steam MSS-822

Bracket Main Steam MSS-823

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! The RI did not identify any conditions which would appear to adversely

affect the operation of the examined system pipe supports. The system

expansion data for Startup Test Procedure (ST) 1-ST-17, " System

Expansion," will be reviewed to ensure that expansion problems do not -

exist at operating temperature and pressure for the above pipe supports i

i and associated structures. The results of this review will be documented

in a subsequent inspection report.

No violations or deviations were identified in this inspection area.

9. Human Engineering Discapancies (HEDs) Review

During this inspection period the NRC RI reviewed Human Engineering  ?

Discrepancy (HED) records to verify that identified improvements in the

control room / operator interface, for selected HEDs, had been completed.

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Design Review Summary Report, issued in October 1984.

The following HEDs were verified complete:

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847 Label "SRV Low-Low Set / Reset" for A and B train.

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833 Label recorders 1CliS-PR2A and 2B to indicate containment

pressure.

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8 Mimic of RHR system. l

! 25 Establish lines of demarcation for RHR Loop B and C.

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. 830 Label switch for manual initiation of containment as

l " Isolation."

248 Label as ADS safety relief valve.

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747 Add escutcheons for manual steam isolation manual initiation

! push bottoms.

] 431 Modify total feedwater flow recorder to read in % flow.

l 328 Annunciator windows consistent in type and style.

l 422 Identify operational limits or warnings on meters.

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q 575 Accurately identify valves (up to root valves) by tagging.

307 Mcdify "HPCS NOT READY FOR AUTO START" annunciator circuit to

actuate for any of the seven conditions which will render HPCS

incapable of auto start.

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

10. Startup Test Procedures Review

The purpose of this area of the inspection was to review selected ST

procedures for compliance with regulatory requirements, FSAR commitments

and Technical Specifications. The NRC inspectors reviewed ST Procedure

1-ST-19. " Core Performance," Revision 0, ST Procedure 1-ST-26, " Safety

Relief Valves," Revision 0, and a draft major change request to 1-ST-26,

and ST Procedure 1-ST-28, " Shutdown From Outside the Control Room,"

Revision 0.

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

11. Startup Test Witness

During this inspection period, the NRC inspectors witnessed startup testing

and operational activities conducted under the low power testing program.

Testing and operational activities witnessed included a reactor startup on

November 17,~1985, a reactor core isolation cooling (RCIC) turbine run on

November 17, 1985, and initial turbine roll on November 26, 1985.

River Bend received a full power operating license (NPF-47) on November

20, 1985, and reactor power was raised above 5% on November 25, 1985.

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

12. Apparent Power Level Exceeding Licensed Level of 5%

On November 18, 1983, the apparent power level exceeded the licensed

level of 5%. The SRI, assisted by a Region IV supervisor, conducted an

inspection to determine the facts of this occurrence,

a. Background

On November 15, 1985, the NRC commissioners voted to allow the NRC

staff to issue a full power license to the River Bend facility. The

full power license was subsequently issued on November 20, 1985.

b. _0_ccurrence

During the night shift of November 17-18, 1985, the on-duty shift

supervisor directed power be raised to approximately 6.5% to

facilitate turbine steam chest warm up. Although other on-watch

personnel questioned this order, it was carried out. The shift

supervisor on watch stated to NRC personnel that he understood th6t

the Commission vote authorized operation above 5% power. The

indicated power of approximately 6.5% power was maintained for about

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2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The event was tcrminated when the oncoming shift supervisor

questioned why power had been raised above 5%.

c. Determination of Actual Power

The NRC inspectors reviewed the licensee's two methods of

quantifying actual power during the event. The first method by a

secondary calorimetric calculation. This calculation indicated that

actual power was about 4.5%; however, the accuracy of this

calculation was not considered good because feed flow was very low.

The measurement of feed flow was not considered to be accurate at

the very low end of the indicating range of the measuring device.

The second calculation was based on the valve position of the

turbine by-pass valves. There are two turbine by-pass valves at

River Bend. Each of these valves is rated at 5% of total power.

One of the valves was shut and the other valve was 50% open. Data

from General Electric taken on identical valves at other plants

indicated that these bypass have linear response or flow

characteristics (e.g., a valve that is 50% open would pass one-half

of its total capacity). The licensee' calculated that the one bypass

valve which was 50% open accounted for 2.5% of reactor power.  !

Allowing for other heat losses such as steam traps and the reactor

water clean up system,' the licensee calculated that actual power was

less than 3.5%

d. Inspection Findings

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The NRC inspectors noted that nuclear instruments could not be set

accurately until power was at a higher level so that an accurate

heat balance'could be made. Nuclear instrument gains were set high,

which is the conservative approach. Indicated power was thus

greater than actual power, and actual power had probably not

exceeded the licensed limit. The NRC inspectors concluded that

there had been a serious breakdown in training and communications.

As the result of interviews with operators and supervisors, it

appeared that this communication breakdown was limited to the one

shift supervisor who had directed the power increase. The NRC

inspectors noted that the licensee had removed this shift supervisor

from licensed duties for retraining and evaluation. It was also

found that the licensee had taken measures to improve communications.

These included the presence of an operations management individual

at all shift turnover briefings, written instructions to operators on

power limit and the resetting of rod blocks to 5%.

It was concluded that continued close monitoring of shift performance

communications was warranted to assess the effectiveness of the

corrective actions taken.

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13. Exit Interview. ,

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j An exit interview was conducted on December.12,.l'985, with licensee

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

. SRI-reviewed the scope and findings of.the' inspection.

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