Text

Insp Rept 50-458/85-43 on 850501-0615.Violation Noted: Failure to Obtain NRC Approval of QA Program Change
	ML20133B909
Person / Time
Site:	River Bend
Issue date:	08/01/1985
From:	Boardman J, Chamberlain D, Jaudon J; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20133B883	List: IR 05000501/2006015; ML20133B879; ML20133B909;
References
	50-458-85-43, IEB-77-06, IEB-77-6, IEB-79-15, IEB-79-23, IEB-79-27, IEB-80-08, IEB-80-16, IEB-80-8, IEB-84-03, IEB-84-3, NUDOCS 8508060287
	Download: ML20133B909 (13)
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, APPENDIX B .

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U. -S. NUCLEAR REGULATORY COMISSION

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REGION IV

NRC Inspection.Reporti 50-458/85-43 CP: CPPR-145

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, 1985

Inspect : v_/ 8

D. 7. C6am@ rTain, Senior Resident Inspector Date-

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

Y A >J 61^

hJ yB rt9 nan, Reactor Inspector, Operations

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( Secti n, Reactor Project Branch (par. 8)

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J./P. yaudo , Chief, Project Section A, Reactor

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froject anch

Approved: c , /4 /4 . /) A [ dI

J./ P. paup6n, Chief, ProIect' Section A, Date

(Reac'to(Projects Branch

Inspection Summary

l- Inspection Conducted May 1 through June 15, 1985 (Report 50-458/85-43)

Areas Inspected: Routine, unannounced inspection of licensee action on

previous inspection findings, site tours, reactor protection system

preoperational test witness, reactor pressure vessel leakage test witness,

l control rod drive system full core scram test witness, reactor coolant system

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8508060287 850002

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hydrostatic ~ test results evaluation, construction quality assurance (QA)

~ program review,-Inspection and Enforcement (IE) Bulletin followup, and

allegation followup.' The inspection involved 147 inspector-hours onsite by

three NRC' inspectors.

Results: .Within the areas' inspected, onb violation was issued in the area of

construction QA program review (failure to obtain.NRC approval of a QA

program' change, paragraph 8).

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DETAILS

1. Persons Contacted

Principal Licensee Employees

K. Arnstedt, Quality Assurance (QA) Engineer

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

T. F. Crouse, QA Manager

J. Davis, QA Engineer

P. J. Dautel, Licensing Staff Assistant

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

D. R. Derbonne, Supervisor, Startup and Test

S. Finnegan, Control Operating Foreman

P. E. Freehill, Superintendent, Startup and Test

D. R. Gipson, Assistant Plant Manager, Operations

P. D. Graham, Assistant Plant Manager, Services

R. W. Helmick, Director, Projects

K. C. Hodges, Supervisor, Quality Systems

D. Jernigan, Engineer, Startup and Test

G. R. Kimmell, Supervisor, Operations QA

G. V. King, Supervisor, Plant Services

J. L. Pawlik, Engineer, Startup and Test

T. L. Plunkett, Plant Manager

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

S. F. Sawa, Control Superintendent, Startup & Test

J. E. Spivey, QA Engineer

R. B. Stafford, Director, Quality Services

K. E. Suhrke, Manager Project Planning & Coordination

L. Sutton, QA Engineer

P. F. Tomlinson, Director, Operations QA

A. Valenzuela, Startup and Test

J. Venable, Mechanical Maintenance Supervisor

D. White, Engineer, Startup and Test

Stone and Webster

D. P. Barry, Superintendent of Engineering

W. I. Clifford, Senior Construction Manager

F. W. Finger, III, Project Manager, Preliminary Test Organization (PTO)

M. Fischete, Engineer, Startup and Test

P. H. Griffin, Site Advisory Manager

B. R. Hall, Assistant Superintendent, Field Quality Control (FQC)

Q. E. Harper, Hydro Test Engineer

D. Hill, Maintenance Engineer

R. L. Spence, Superintendent, FQC

The NRC senior resident inspector (SRI) also interviewed additional

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licensee, Stone and Webster (S&W), and other contractor personnel ~during

lthe inspection period.

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Denotes .those persons that attended the exit' interview conducted on

!. jJune21,1985. _.

[2. Licensee Action on Previous Inspection Findings

a'. (0 pen) 0'enp Item (458/8408-01): ' Review to determine if and how the

diesel; generator loading restrictions of calculation 12210-E-122 are

implemented in plant operating procedures.

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The. SRI obtained a copy.of calculation 12210-E-122, Revision 4,

" Standby Diesel generator Loading Calc. ," dated March 1,1984. This

revision.of the calculation is based on a 3500-KW loading limit on

the diesel and does not' reflect the latest load restriction of 3130

KW. However, a GSU letter RBG - 20,086 dated February 6, 1985,

.contains revisions to the FSAR "to establish a qualified load for

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each of the diesel generators." These revisions include, for diesel

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1EGS*EG1A loading, a requirement.that "LPCS or RHR A pump shall be

. manually tripped after 2.0 hr of LOCA, depending upon the available

' diesel generator sets" and for diesel 1EGS*EG1B loading, a requirement

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that "RHR C is stripped manually.by the operator after 2.0 hr of

operation after LOCA, depending upon the combination of the available

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diesel generator. sets."

Abnormal Operating Procedure A0P . .0004, Revision 1, " Loss of Offsite

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Power," dated April 10, 1985, appears to address the latest loading

' 1 restrictions (3130 KW) for the diesels,' but the stated action for

' manual tripping of an RHR or LPCS pump in Section 5.8 needs some

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clarification. For example, it is not clear under what conditions

that RHR A pump is tripped,instead of the'LPCS pump. This item will

remain open pending issue'of an approved calculation reflecting

. qualified diesel loading and pending the required clarifications in

procedure Section 5.8.

, b. (0 pen) Unresolved Item (458-8408-04): Review of licensee ' "

program,'

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. for tracking of commitments to the NRC. -

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GSU has developed and implemented a commitment tracking program at

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River Bend. Project Procedure' No. 8.2,'" Identifying and Tracking

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Project Commitments," was issued on Augu~st 9,.1984,Dto " provide

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guidelines for River Bend Nuclear Group <(R8NG) organizations for

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identifying, documenting, tracking,,and closing commitments made .to

." . regulatory agencies." Nuclear licensing is responsible for the

' :l :' ' ' tracking system' with QA responsible for verification of completion .

( of commitments on a sampling basis.' ' GSU' uses the "19AC" computer

w program and they have identified approximately ^2,152,t.cenitments to

date. The present status of the 1,191'open commitments is 717.high'

priority commitments required for fuel, load and 156 required after

fuel load, and there are 280 low priority commitments required for

fuel load and 38 required after fuel load.

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L c. This item will remain open for the SRI to evaluate the GSU method _of

w identifying commitments that require closure prior to fuel load and

> for identifying those that can be completed after fuel' load.

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c. (Closed) Open Item (458/8434-01): Implementation of preoperational

3*, test commitments ~by the control rod drive'(CRD) hydraulic

preoperational test procedure (1-PT-052).

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ir^ The' specific items of concern were impienentation of'the following ,

Final safety Analysis Report (FSAR), Chapter 14, test commitments:

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"1.e. To verify the failure mode of the CRD

L system on loss of power."

' "3.j. The CRD pumps are tripped and the time for

accumulator' inoperable alarms to occur is

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,. recorded as baseline data."

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"4.f. . All scram valves open on a loss of

g instrument air to the CRD, system."

- The above items were addressed in the. following manner:

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1. 'e . Th'e failure mode of the CRD system was

tested by verifying the scram function on a

> loss of power to the scram' pilot solenoids.

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3.j. ~A minor change request (MCR 09) was issued

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to require recording of the times for all

hydraulic control units which exhibited low

accumulator pressure alarms within 10

minutes after tripping oflthe CRD pump.

4.f. An acceptance criteria step 10.9 was added

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9 by a major change request (MRC 06) to

reference the backup scram valve test which

demonstrates that the scram valves open on a

loss of instrument air.

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This item is closed.

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d. .(Closed) Violation (458/8507-01):- Procedures were not implemented-

to maintain Class 8 cleanliness requirements in the spent fuel storage.

area'where the new fuel was to be stored'in accordance with the

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Special Nuclear Material License issued on January 15, 1985.

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GSU took immediate action to issue an unsatisfactory inspection

report and fuel' receipt was delayed 1 day to allow removal and

inspection of the spent fuel racks and clean up of the spent fuel

pool floor. Following'the cleaning, the spent fuel racks were

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reassembled in the pool and fuel receipt progressed as scheduled.

Also, to prevent recurrence of this type of item, housekeeping and

cleanliness procedures shall be implemented as a prerequisite to the

governing procedures.

This item is closed.

e. (Closed) Deviation (458/8507-02): Preoperational test procedures

are not being provided for NRC review 60 days prior to the scheduled

test performance in all cases.

Only four preoperational test procedures remained to be submitted for

NRC review at the time of this deviation. The.four remaining

procedures were expedited and all have now been submitted.

This item is closed.

f. (Closed) Open Item (458/8522-07): GSU has installed motor operated

valve'(M0V) circuit breaker trips that can be reset either manually

or automatically. GSU has not established control to verify that all

such resets are in the manual mode.

Motor control center starters for MOVs at River Bend have both

thermal overload trip and magnetic trip devices. The magnetic trip

devices trip the manual circuit breaker to remove the overload

condition.' The magnetic, trip device then resets automatically, but

the~ circuit breaker must be manually reclosed to provide power to the

' starter. The thermal overload trip device opens the circuit to the

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motor starter to: remove the overload condition. The thermal overload

device'has a hand / automatic option on reset. GSU has chosen to place

all of the thermal overload ' devices in the hand reset position.

Temporary Change Notice.No.85-131 has been issued to revise Procedure

No. CMP-1026, " Corrective Maintenance of MCC Starters," to include a

step for verifying that hand automatic reset selectors are in the

hand position for thermal overload trip devices. GSU operations was

notified of this' condition per memorandum APM-M-85-94 dated June 13,

1985. It was also noted during the review of this item that certain

loss of coolant accident initiated MOVs would have not thermal

overload trip devices installed.

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This item is closed.

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i g. (Closed)) Open Item (458/8522-12): A system had not been provided

for assuring that each piece of measuring and test equipment (M&TE)

is calibrated and adjusted on or before the date required.

Procedure ADM-0029, Revision 4, " Control of Measuring and Test

Equipment (M&TE)," has been revised via temporary Change Notice

< 85-733 to clarify the system used and the responsibilities for the

recall of M&TE for calibration. Also, in addition to the recall

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. requirements, the M&TE issue facility must verify that the M&TE

calibration due date is current prior to issue of the M&TE and users

of M&TE are required to verify that the calibration of M&TE is

current prior to use. All of these requirements are intended to

preclude the use of any M&TE for which the calibration has expired.

This item is closed.

3. Site Tours

The SRI toured areas of the site during the inspection period to gain

knowledge of the plant and to observe general job practices. The site

tours conducted included special tours on separate occasions with

Commissioner Bernthal and with a group fron NRC Nuclear Reactor Regulation

headed by Harold Denton. Both of these tours included the conduct of mock

scenarios on the River Bend plant simulator.

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

' 4. Reactor Protection System Preoperational Test Witness

The SRI witnessed portions of the reactor protection system response time

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measurements testing conducted during this inspection period. The

specific testing witnessed included reactor pressure sensor response

timing, reactor vessel level sensor response timing and drywell pressure

sensor response timing. Testing personnel experienced several problems

with the set up of the response time test equipment which caused testing

delays. They also experienced problems with interpreting the response

time curves such that the proper ramp was generated for acceptance.

criteria purposes. The vendor for the test equipment was brought to the

site and the test equipment problems were corrected. Also','a uniform

method for interpreting the response time curves was formulated. The SRI

conducted a preliminary review of several response time curves and it

appeared that the response times were within acceptance criteria limits.

The major testing remaining for the reactor protection system

preoperational test at the end of this inspection period was the

intermediate range monitor (IRM) and average power range monitor (APRM)

response time measurement testing.

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

5. Reactor Pressure Vessel Leakage Test Witness

This special reactor pressure vessel (RPV) leakage test (1.MPRV.002) was

performed in order to disposition a Nonconformance and Disposition (N&D)

Report No. 11275. This N&D resulted when the review of the N-5 data

reports on the reactor pressure vessel indicated that no hydrostatic test

was performed subsequent to the installation of reactor-internals or

rework on nozzle safe ends. This included installation of items such as

control rod drive (CRD) housings, incore housings, recirculation and

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feedwater safe end rework, jet pump.pene'tration seals, etc. The reactor ,

. pressure-: vessel system hydro procedure (1-G-ME-15) and associated

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. ; documentation did not identify the welds for these items,as being'withinJ

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.the scope of the RPV hydro-test inspections.' _Therefore, during the RPV '

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hydro' conducted in May 1984, documentary evidence of the inspection off

,t these welds was not obtained.

'TheSRIwitnessedthe-RPVleakagetestingandweldinspectionspe[ form ~

Jon May 16, 1985. The leakage testing was performed at a design pressure

E of 1250 psig. Initially, trouble was. experienced with obtaining the test-

. _ pressure _due to excessive leakage through the gaged safety relief valves.

Test personnel obtained apnroval from the relief valve vendor which

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allowed the relief valve gags to be torqued to 30 foot pounds psig at a

. .RPV pressure of 1000 psig. This was accomplished and they were then able

, , to obtain the required test pressure. The test pressure was held for a

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minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the performance'of the official inspections.

The inspections were' performed and no problems were identified.

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

L< 6. -Control Rod Drive System Full Core Scram Test Witness *

During this: inspection period, two back up scram valve full core reactor

scram tests were performed to complete the control rod drive system

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preoperational testing. Also, a special full core reactor scram test was

performed to evaluate the scram discharge volume level instrument response.

The. SRI witnessed the perfo'rmance of the special scram test on May 29, 1985.

'This special scram test was performed ~in conjunction with reactor

, ' protection system response time testing and the scram was initiated by a

reactor-vessel low water signal.

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

7. Reactor Coolant System Hydrostatic Test'Results Evaluation

The SRI conducted a review of the completed test results_for the RPV

system hydrostatic test (Procedure 1-G-ME-15) and for the subsequent

reactor pressure vessel' leakage test.(Procedure 1.MPRPV.002). The

specific areas reviewed and findings noted included the following:

a. Changes to the test procedures were documented and implemented in

'accordance with the licensee's administrative controls.

.b. The system boundary either included all piping and equipment

protected by the safety' relief valves or documentation was provided

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to show that; separate hydros were performed on equipment.or piping

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that could be isolated from the RPV. ,

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c. The water quality met all requirements.- ,

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. . d. The licensee held the maximum test pressure (1.25 times the design

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pressure) for at least 10 minutes during the RPV system hydro test.

The hydrostatic test pressure did not exceed the maximum pressure

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.f. -.The reactor coolant temperature ~was maintained above the~ nil

ductility transition temperature throughout the hydro and leak

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g. All' identified test exceptions 'have been resolved, but a concern was

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identified with Test Exception TE-13 for Procedure 1-G-ME-15. This

test exception addressed certain flexible hoses that were not

installed'at the time of the original hydro. There were 40 hoses

i identified on the test exception and they were apparently identified

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on system punch lists for~ installation and hydro at a later date.

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. , , -The SRI selected 4 flexible hoses (Nos. 114, 121, 122, and 140) for a

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review of the rework documentation to determine if the required

, hydros were performed. Of the four selected, documentation was

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- obtained to verify that the' flex hoses did receive ~a subsequent hydro

However, it was noted that the flange connections on these

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flex hoses had been blanked during the hydro'and only two of the.four

. - rework control forms required a subsequent operational . leak test

(OLT). This was discussed with test personnel and it was determined

that the performance of OLTs, on a flange connection that.is completed

after a hydro, has been normal practice at River Bend. Further

review revealed that three of the four hoses received an OLT during

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the RPV leakage test per a startup trouble ticket (STT). The fourth

hose was not specifically. mentioned on the STT, but it was identified

on. working drawings as being inspected. The SRI believes that no

problem was created by failure to note an OLT requirement on the

rework document and test personnel stated that the normal practice

will continue for performance of OLTs.

l h. The test results have been reviewed and approved by those personnel

charged with the responsibility.

The SRI also. reviewed selected vendor supplied pump and valve hydro

records and no problems were noted.

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

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8. Construction QA Program Review

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, 'A review of the licensee construction QA program revealed that on June 10,

1983, GSU forwarded fnr approval a revision of their construction QA

program as required by 10 CFR 50.55(f)(2). Discussions with licensee

personnel revealed that revised Section 17.1.2.4.A reducing the periodicity

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of GSU . review of all controlled documents from 1 year to 2 years was made

after March 11 1983. Licensee personnel further stated that this change

o .had been implemented, although it had never been approved by NRC.

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Implementation of this change is in violation of 10 CFR Part 50.55(f)(3),

which requires prior NRC approval of licensee QA program changes made

after March 11, 1983, when such changes reduce commitments. (8543-01)

9. Inspection and Enforcement (IE) Bulletin Follow Up

The purpose of this inspection was to followup on licensee action taken in

response to Inspection and Enforcement Bulletins (IEBs),

a. IEB 84-03

This bulletin concerned the consequences of a failure of the

refueling cavity seal. The NRC inspector reviewed the following

licensee correspondence to the NRC (Region IV).

Letter Serial Date Items Addressed

RBG-19487 11-29-83 . Cross Seal Failure

. Maximum Leak Rate

Because of Seal

Failure

.Make Up Water

Capacity

. Potential Effect on

Stored Fuel and

Fuel in Transfer

.0ther Consequences

RBG- 20042 02-01-84 . Emergency Operating

Procedures

RBG- 20635 04-05-85 . Time to Cladding

Damage Without

Operator Action

RBG-21023 05-15-85 . Time to Cladding

Damage Without

Operator Action

These four letters address all of the points required by IEB 84-03.

The design of the seal used at River Bend is a stainless steel

bellows assembly welded to its support structure. The maximum

credible leakage rate is within make up capacity. Total failure of

the seal without operator action could result in a problem for fuel

in transit between the reactor vessel and the containment fuel

storage pool. This is addressed in licensee Procedure A0P-0032,

which requires the fuel to be placed in either the vessel or storage

racks. All other fuel would remain covered with water, and no vital

equipment would be flooded by a complete draining caused by a bellows

failure. The bellows is also protected from direct impact by a

radiation shield and a guard ring.

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IEB 84-03 is considered closed.

b. IEB 77-06

This bulletin concerned General Electric Series 100 containment

electrical penetrations. River Bend does not use this type of

electrical penetration. Therefore, there is no action required at

River Bend for IEB 77-06.

IEB 77-06 is considered closed.

c. IEB 79-15

This IEB addressed deep draft pump deficiencies and the long term

operability of these pumps.

The NRC inspector found that the licensee had addressed operability

of pumps in the FSAR, and this was recognized in NUREG-0989, the

safety evaluation report for River Bend. Additionally, the final

draft technical specifications contained surveillance requirements

for monthly demonstration of pump operability in accordance with the

ASME code,Section XI and an 18-month system operability test. Since

the question of deep draft pump operability is being addressed by the

normal review process and implementation of the requirements to test

is under the routine inspection program, no additional tracking of

this IEB is warranted.

IEB 79-15 is closed for record purposes.

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d. IEB 79-23

This IEB concerned the potential failure of emergency diesel

generators. The failure could result if there was a large

circulating current between the exciter transformer and the

generator. Such a circuit could be set up by connection through a

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common ground. It was found that the design of all three emergency

diesel generators at RBS was such that exciter transformers had a

floating primary neutral. Subsequent testing of the emergency diesel

generators did not disclose any problems of the type discussed in

this IEB.

IEB 79-23 is considered closed.

e. IEB 79-27

This IEB concerned the loss of non-Class 1-E instrumentation and

control power. This IEB was not specifically directed to River Bend,

but it was included in the FSAR review, becoming Question 421.003 and

as Confirmatory Item 31 of NUREG-0989, the Safety Evaluation Report.

Since the action required by this IEB is being tracked as a confirma-

tory item, IEB 79-27 is closed for record purposes.

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f. IEB 80-08

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This IEB addressed radiography of flued head design penetrations of

the containment. The licensee was found to use flued head design in

both the containment and the drywell. The licensee committed to use

radiography on all flued head design penetrations of the containment

and other nondestructive tests on other penetrations.

, This IEB is closed.

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g. IEB 80-16

This IEB concerned Rosemont pressure transmitters, Models 1151 and

1152. When these transmitters were fitted with either "A" or "0"

output code cards, it was possible for the transmitters to have an

ambiguous output and the input signal was either an over pressure or

a reverse pressure signal. The licensee found that there were four ,

Rosemont 1152 transmitters with an "A" output code. These four

transmitters were modified to have an "N" output board.

IEB 80-16 is considered closed.

10. Allegation Follow Up

The NRC inspecto.a did a followup inspection of an allegation.

. Backaround. An anonymous letter was sent to both Gulf States

Utilities (GSU) and the NRC. This letter forwarded an internal piece

of Stone and Webster (S&W) correspondence. This S&W correspondence

was a letter signed by engineers in the design group for small bore

pipe. The letter complained that the group was being required to

account for on-the-job time in a log and alleged that such a time

accounting procedure was inimical to quality assurance.

. Licensee Action. GSU conducted a quality assurance review of the

small bore piping group March 4-22, 1985.

. NRC Review. The NRC inspector reviewed the report of the licensees'

review. It was noted that the licensee had concluded that the

allegation was not substantiated in that the use of a time log to

account for time spent on various charge items does not have a direct

relationship to quality assurance. It was further noted that there

were four additional concerns noted by the quality assurance review.

The NRC inspector noted that the licensee had followed up on these

four concerns and closed them.

. Conclusion. The NRC inspector concluded that the allegation was

substantiated in that the time log was kept but that it was invalid

as a safety concern.

This item is closed.

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11. Exit-Interview .

An exit interview was conducted on June 21, 1985, with licensee

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

SRI reviewed the scope and findings of the inspection.

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ML20133B909

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