IR 05000454/1987038

Insp Repts 50-454/87-38 & 50-455/87-35 on 870901-1001. Violation Noted.Major Areas Inspected:Licensee Action on Previous Insp Findings,Lers,Operations Summary,Training, Surveillance,Maint,Operational Safety & Allegations

ML20236B871
Person / Time
Site:	Byron
Issue date:	10/13/1987
From:	Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20236B797	List: IR 05000454/1987038 ML20236B794 ML20236B826
References
50-454-87-38, 50-455-87-35, NUDOCS 8710260385
Download: ML20236B871 (13)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

i Report Nos. 50-454/87038(DRP); 50-455/87035(DRP)

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Docket Nos. 50-454; 50-455 License Nos. NPF-37; NPF-66-l Licensee:

Commonwealth Edison Company l

Post Office Box 767 Chicago, IL 60690 Facility Name: Byron Station, Units 1 and 2 Inspection At: Byron Station, Byron, Illinois Inspection Conducted:

September 1 - October 1, 1987 Inspector:

P. G. Brochman

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Approved

J. M. Hinds, hief t o IS 87 eactor Projects Section 1A Date Inspection Summary

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l Inspection from September 1 - October 1, 1987 (Report Nos. 50-454/87038(DRP);

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I 50-455/87035(DRP))

Areas Inspected:

Routine, unannounced safety inspection by the resident inspector of licensee action on previous inspection findings; LERs; operations summary; headquarters requests; training; surveillance; maintenance; operational safety; Region III requests; allegations; and management meetings.

Results: Of the nine areas inspected, no violations or. deviations were identified in seven areas. One violation was identified in one of the remaining areas (failure to perform a written safety evaluation after a I

modification to the plant's design paragraph 9). The violation was of

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more than minor safety significance with more than minimal potential to l

affect the public's health and safety.

Two additional violations were identified in the other remaining area; however, in accordance with 10 CFR 2, Appendix C, Section V.A, a Notice of Violation was not issued (failure to notify the NRC of a change to the Unit 2 startup test program paragraph 3.a; failure to perform a surveillance test within the required time interval paragraph 3.b ).

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8710260385 871020 PDR ADOCK 05000454 G

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DETAILS-

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Persons Contacted

1 Commonwealth Edison Company

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• T. Maiman, Vice. President,..PWR j

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• K. Graesser, General Manager, PWR Operations

• E. Fitzpatrick, Braidwood Station-Manager-

• R. Querio, Station Manager

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• R. Pleniewicz, Production Superintendent-

• R. Ward, Services Superintendent

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• W. Burkamper, Quality Assurance Superintendent

• K. Ainger, Nuclear Licensing Administrator.

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• L. Sues,. Assistant Superintendent, Operating-

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• G. Schwartz, A'ssistant' Superintendent, Maintenance

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• T. Joyce, Assistant Superintendent, Technical Services '

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• F.

Lentine, PWR~ Licensing ~ Supervisor -

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• D. Elias, Engineering Superintendent.

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• D. St. Clair, Asststant Superintendent, Work Planning'

T. Higgins,. Operating Engineer, Unit 0

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J. Schrock, Operating. Engineer, Unit 1 D. Brindle, Operating Engi_neer, Unit 2 T. Didier, Operating: Engineer, Rad-Waste

• M. Snow, Regulatory Assurance Supervisor j

.F. Hornbeak, Technical. Staff Supervisor

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• R. Flahive, Radiation / Chemistry Supervisor,

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P. O'Neil, Quality Control Supervisor-i

• W. Pirnat, Regulatory Assurance Staff.

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• t. Zittle, Regulatory Assurance-Staff

• W.- Walter, Assistant Technical Staff Supervisor

• D. Robinson, Onsite Nuclear Safety' Supervisor,

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• A. Snow, Assistant Training Supervisor i

• A. Chernick-Training Supervisor-j

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• P. Barnes, Regulatory Assurance Supervisor i

• W. Bielasco, Lead Health Physicist The inspector also contacted and interviewed other licensee and j

l contractor personnel during the course of this inspection.

• Denotes those present during the management meeting and exit interview l

on October 1, 1987.

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Action on Previous Inspection Findings (92701 & 92702)

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(Closed) Violation'(454/86029-01(DRP)):

Failure to cont'rol

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nonconforming material, installation of a defective component as a

a' reactor coolant system pressure; boundary,.and the~ breakdown.'

of management controls-in the maintenance'and: quality. assurance programs. This violation was classified as severity. level;III,'

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and a $25,000 civil penalty was proposed in a letter from J._G. Keppler to J. J. O'Connor, dated November 18, 1986, for.

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installation of a defective pressurizer code safety valve. :The licensee's response, a letter from B. L.'_ Thomas _to J. M. Taylor,

dated December 20, 1986, described the corrective action taken

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and a basis for completely mitigating the= civil penalty.

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the civil penalty and issued an order imposing the proposed civil penalty, without any further mitigation, in a letter from J. M. Taylor to J..J. O'Connor, dated June 11, 1987.

The licensee paid the civil penalty on July 14, 1987, as documented it. a letter from L. D. Butterfield to the NRC.

The inspector verified that the following corrective actions have been implemented, as described in the licensee's December 30, 1986-letter.

The defective code safety valve was replaced.

Procedures-for bench testing safety valves were revised to more clearly

indicate equipment identification. Cages have.been constructed to i

store contaminated nonconforming and ready for use material. A i

conduct of maintenance program was developed by the station. This program is under review by the licensee's corporate office with a view to implementing some form of it at all Commonwealth Edison plants.

Based on the actions taken, this item is considered closed, b.

(Closed) Unresolved Item (454/86040-05(DRP); 455/86040-08(DRP)):

Potential failure of valves 1AF024 and 2AF024 due to an unanalyzed fire scenario.

The inspector had postulated a scenario wherein i

both trains of auxiliary feedwater for a unit could be lost due to a fire concurrent with an earthquake or tornado..The licensee responded to this concern in a letter from K. A. Ainger to l

J. G. Keppler, dated December 15, 1986.

In this response the licensee stated that the NRC regulations for fire protection do not require a licensee to design against a fire and concurrent tornado or seismic event. The inspector discussed this position with the NRC fire protection consultants (Brookhaven~ National

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Laboratory) and Region III fire protection inspectors and verified that this regulatory position is correct.

The licensee has installed label plates at the auxiliary feedwater controls in the main control room and at the remote shutdown panels to alert the operators to verify recirculation line flow before reducing total pump flow to less than 85 gpm.

Based on a review of'the 10 CFR 50 Appendix 12 requirements, this item is considered closed.

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(Closed) Violation (454/86047-02(DRP))r Failure to analyze grab l

samples with radioactive effluent monitoring channel 1RE-PR002 l

inoperable. This violation,was related to inadequate communication

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between the operating and radiation chemistry departments.

Grab'

samples are taken and analyzed as a compensatory measure when a radiation monitor is inoperable.

As corrective action, the licensee I

requires that the LOCAR (limiting condition for operation action i

requirement) be hand carried to the radiation-chemistry department l

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a forimplementationofthsLC0 action'redirementsl.Dperating

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personnel review the status of. all, LOCARs;each shift and verifyl

that the action; requirements are beingLimplemented.. The inspector'

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interviewed' operating and radiation-chemistry department personnel

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to veri fy their" understanding..of. these requirements.

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actions,:this' item is considered closed.

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(0 pen) Violation (454/87022-01(DRP)): Containment. spray ~ additive L

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system was. inoperable due to mispositioned valves.

The? inspector-i reviewed the licensee's response 'and verified that1the 'actio'ns -

completed to date had actually been accomplished. :The-inspector:

did not identify.'any concerns with"the corrective' actions taken.

H to date. Additional corrective actions have not'~yet been: completed, j

and the licensee is presently. revising the administrativeLprocedures l

which control'the locked valve and the abnormalTvalve positiony

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I programs. This violation will remain open pending completion of t the licensee's revision of these; programs'and the inspector's

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review of these programs.

(Closed) Unresolved Item (454/86028-01(DRP)):-. Inspector _co'ncerno e.

related to calibration of strip chart ' recorders; During' the-performance of.a Technical Specif.ication surveillance-on Power:

Ranger Nuclear Instrument N42, the inspector. observed that the strip; chart recorder in use did not have,a calibrationtsticker. Two-of

the reactor protection trip. signals' generated by the power range 1

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q nuclear instruments are positive' and negative flux rate, which l

require a minimum change'in neutron flux level over.a maximum time:

i period. To verify that this' ramp function is within Technical

'1 Specification limits,'a strip chart recorder is utilized.to' measure the time dependent portion of the ramp function.

_ ith aiknown;sp'eed.

W of the strip chart paper, measuring the distance between two points..

and dividing by tha speed will yield the length of time.

10 CFR 50, Appendix B, Criterion.XII,'as' implemented by. Commonwealth'

Edison Company's Quality Assurance Manual, Quality Requirement 12.0,-

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gauges, instruments, and other measuring and testing devices used a

in activities affecting quality are properly controlled, calibrated, D

and adjusted within spr:cific periods to maintain. accuracy.within

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necessary limits. Quality Requirement 12.0,.Section 12,2 ' requires

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that inspection, measuring and test equipment'will be adjusted-and calibrated using documented procedures at scheduled. intervals against certified standards having known: valid relationships to

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recognized standards.

Consequently, since the str.ip chart recorder-is used to measure time for an~ activity affecting quality,.-the speed of the strip chart recorder should be aligned and calibrated.

The failure to calibrate the speed of the strip chart recorder that'

q was used to perform a Technical Specification surveillance'could be; classified as a violation of 10 CFR 50, Appendix B,. Criterion'XII.

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During discussions,.the licensee's, staff stated that the motor which dri_ves the strip chart is a synchronous ' motor.

Synchronous motors-rotate at a' fixed speed, which is dependent on the Ac lin'e

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c frequency.

Since the frequency does not change ~(itLis effectively?

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constant at 60~Hz),1and the motor is_not_ adjustable, the. licensee:

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believed it was unnecessary to calibratefthe strip chart recorder R

. speed. However,'no exception is made in the quality assurance'

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program with regard to the_ calibration of strip chartJrecorders.

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Subsequently, the licensee has placed the strip chart-recordersLin the calibration _ program, with a. calibration interval of six months;

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The licensee has' verified that the strip chart recorders were within-

?l calibration tolerances. The: surveillance:have.beenfsuccessfully~

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sperformed,after the calibration of'the' strip chart recorder; The~.

'a serial. number of the' strip chart recorder is now entered on the data._

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H sheet with'all of the other test equipment;.its useican be identified should its accuracy become. suspect. This event is. considered to-be isolated in nature and had minimal.impactson the. safety of"the plant. Based on this information the NRC has decided not to. issue a violation and this' item is considered closed.

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censee Event Report (LER) Followup (92700)-

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(Closed) LERs (454/87017-LL; 454/87018-LL;. 455/87006-1L; 455/87007-1L; 455/87011-LL, 455/87012-LL; 455/87013-LL; 455/87014-LL;. 455/87015-LL;.

..L 455/87016-LL):

Through direct observation, discussions with licensee ~

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personnel, and review of records, the following LERs were reviewed to determine that the deportability requirements _were fulfilled,Limmediate j

corrective action was accomplished, and ' corrective action.to prevent-j recurrence had'been accomplished, in accordance with Technical.

Specifications,

LER No.

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l 454/87017 Two Unit 1. reactor! trips caused by-lightning strikes.

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454/87018 Reactor trip due' to feedwater pump: trip -on' failed

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C thrust bearing wear sensor.

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455/87006-1-Reactor trip from 2/4 logic coincidence on low pressurizer pressure when a pressure 1transmi.tter..

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failed.

455/87007-1 Reactor trip due to spurious 2/4 logic coincidence on i

.high neutron. flux, when an instrument inverter

failed.

M 455/87011 Reactor trip during'30%-load rejection. test ol R

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Overtemperature" Delta T, due to turbine: control system-

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.and steam dump problems.

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455/87012 2B Diesel Generator inoperable.beyond Technicali l

Specification limits due to turbocharger failure.-

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455/87013

. Violation of license. condition for. unit 2'startuo test program.

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455/87014 Surveillance intervalEfor containment airlock exceeded due.to.p'ersonnel,. error.

455/87015 Feedwater' isolation,in. Mode.4 du'e tofpersonnel:

working:on wrong component.

455/87016.

. Safety' injection in Mode 4.due~to-mispositioned switch.

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The events described in LERs'454/87017, 454/87018,.455/87011,.455/87012,

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and 455/87016 are' discussed further in inspection-reports

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and 455/87031(DRP).

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With regard to LER 455/87013, this LER discusses an event which-occurred on August'19, 1987,.with reactor power at 94%, and involves-

'I the failure to notify the NRC within one month of'a change to the'

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startup test program, as defined.in the. Byron.FSAR.

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Condition 2.C.3 of the unit 2 operating license, NPF-66, requires

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that any changes to'the startup test program, described-in chapter j

14.of the Byron FSAR, made'in accordance with the provisions of.10 CFR 50.59, shall be reported-in writing to the NRC within one month.-

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FSAR Table.14.2-67, Incore Flux Monitor System," specifies that

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the startup test of the system be performed in hot. shutdown;(mode 4)

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after core. load and before initial criticality..Startuptest'(SVT)-

l 2.45.80, "Incore Flux Mapping System Checkout," was. performed

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between December 9,'1986 and January 2,1987, while the plant' was,

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in modes 3, 4, and 5, rather than just mode 4.

l-l SVT 2.45.80 did not specify that thentest'could'only be performed in mode 4.

The discrepancy was discovered:during a post-test contractor review of the results.

The discrepancy.'was evaluated by the licensoe's offsite-engineering organization (PED), and on

July 1, 1987, a safety evaluation.was.'perfor_m.ed.in,accordance'wi.th-l 10 CFR 50.59; the evaluation determined that the: test, as performed,

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met the intent of FSAR Table 14.2-67.

However, the NRC was not I

notified of this safety evaluation and change to the test program l

until August 19, when it was discovered by Byron. station personnel.

The failure to notify the NRC within one month of-a change to the startup test program is a violation of the facility' operating

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license NPF-66 (455/87035-01(DRP)).

However, this violation meets i

i the tests of 10 CFR 2, Appendix-'C,Section V.A; consequently, no'

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Notice of Violation will be issued, and this matter.is ' considered closed.

Because all startup testing and test results reviews are completed.

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at Byron.

However, a memorandum has been sent to PED.to review the l

requirement to notify theLNRC, as the reviews of'the Braidwood l'and

2 startup test program's are still:in progress. The incorrect plantL j

conditions specified in SVT 2.45.80'will be evaluated during the.

NRC's review of the startup test program results-for Byron'2.

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With regard'to LER 455/87014, this LER. discusses'an event which occurred on August 23,.1987., with reactor power at 93%, and involves q

the failure to: perform a.surveillalnceitest on the containment-

airlock within the' required intervalc

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At 0057~on~ August 20, 1987, the~ Unit 2 containment.wasJentered via

the. normal airlock. Technical' Specification 4.6'.1.3.a requires j

that a local leak rate. test be performed ~on the volume between thel l

airlock doors'1 gaskets within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of using the airlock.

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.l Technical Surveillance 2 BVS G.I.3.a-1, " Primary Containment Type B-j Local Leak Rate. Tests of the Equipment: Hatch Airlock Door Gasket j

Interspaces,!' implements this' requirement. Technical Specification j

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4.0.2 requires that each surveillance be performed within the.

l required interval, with a maximum allowed extension of.25% of the-

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specified surveillance interval.

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To determine when the. surveillance must'be performed a. technical _

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staff engineer reviews the R-(Radiation) key ;1og-book.. Acce ss.-to --

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containment is controlled by-the radiation-chemistry department and the airlock door is secured.with a lock which requires an R key to open it.

The radiation-chemistry departmentElogs the. issuance of'

i each R key; consequently, the R. key log serves' as' a record of-j containment entries. On August'21 the engineer ~ failed'to note that there had been an entry into the Unit 2 containment on' August 20.

A review of other logs on August 24 identified that a' containment entry had been made on August 20.

After' identification of the.

containment entry. the surveillance was'.successfully completed by.-

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.J 1125 on the same day. Therefore, the 72-hour time limit plus 25%-

extension (18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />) was exceeded by 16.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

As temporary corrective action the licensee is performing leak ~ rate testing every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, regardless.of'whether any containment entries have been made. As long term corrective action.the licensee has installed an automatic airlock leakage detection system'on both units and submitted a request to change technical specifications.

The use of an' automatic system, rather than manual testing, is_

allowed by 10 CFR 50, Appendix J.

The failure to perform a local leak rate test on the Unit 2 airlock door gaskets within 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> of an entry into containment is 'a violation of Technical Specifications 4.0.2 and 4.6.1.3.a (455/87035-02(DRP)). However,.this violation meet's the tests of 10 CFR 2, Appendix C, Section V.A.; consequently, no Notice of Violation will be issued, and this matter is considered closed.

No other violations or deviations were identified.

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Summary of Operations Unit 1 operated at power levels u'p to 98%,for the entire report period.

Unit 2 remained shutdown for a forced outage until 1718 on September 4; 1987, when the unit was taken critical. 'The unit was' synchronized-

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to the grid at 0603 on September 5',

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The unit operated at power J

levels up to 95% until 1323 on October 1, 1987, when a feedwater pump f

trip and subsequent reactor trip occurred.

The unit was taken critical

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at 2111 the same day.

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Followup-on Headquarters Requests (25019)

-(Closed) Temporary Instruction (TI) (2500/19): Review of. licensee's

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actions taken to implement unresolved safety issue A-26.(systems for mitigating low-temperature overpressure transients on the reattor

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vessel).

This TI provided instructions for review of the reactor

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coolant system (RCS) cold over pressure protection' system in the

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areas of design, administrative controls and procedures, training and j

surveillance.

The design of the overpressure protection system is discussed in section 5.2.2.11 of the Byron FSAR and was reviewed by the NRC staff in the Byrcq Safety Evaluation Report (SER), NUREG-0876.

In SER paragraphs 5.2.2.2~

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and 8.47, the NRC staff concluded that the licensee design had met the requirements of.10 CFR.50, Appendix A, General Desig. Criteria 15 and 31, and Appendix G.

Additionally, the design recommendations of Three Mile Island (TMI) Action Plan Items 11.0.1, II.D.3, and II.G.I of!NUREG-0718 and NUREG-0737 have been incorporated into the design of the RCS cold overpressure protection (COPP) system.

The COPP system is manually armed by the reactor operator as the RCS is cooled down.

If the RCS is cooled

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below 350 F without arming the COPP system, an annunciator alarms, i

Additional annunciators indicate that the system has actuated, is about-j to actuate, or is not properly aligned for operation.

The inspector reviewed administrative controls and procedures ' associated with the COPP system and verified that the Safety' Injection pumps and

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one of the charging pumps are taken out of service prior to decreasing RCS temperature below 330 F.

The inspector verified that the plant cooldown procedure provides guidance for restarting stopped reactor coolant pumps and maximum temperature differentials between'the reactor vessel and the steam generators. The inspector reviewed the' licensee's training plan for licensed operators and verified that cold overpressure accidents are discussed and that the r. requirements and procedures for the use of the COPP system are included.

The inspector reviewed the surveillance associated with the COPP system and verified that the temperature, pressure, and comparison circuits are periodically aligned. Additionally, the relief valves are periodically stroke time tested.

The equipment is tested on a recurring basis, rather than prior to each use. Based on this inspection, this TI is considered closed.

No violations or deviations were identified.

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Training (41400 & 41 W +

The ef. festiveness of training programs for licensed and nonlicensed-personnel was reviewed by the inspectors during the witnessing of the licensee's performance of routine surveillance, maintenance, and operational activities and during the review of the licensee's response to events which occurred during September 1987.

Personnel appeared to be knowledgeable of the tasks being performed, and nothing was observed which indicated any ineffectiveness of training.

No violations or deviations were identified.

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Monthly Surveillance Observation (61726)

Station surveillance activities of the~ safety-related systems and components listed below were observed or reviewed to ascertain that they were conducted in accordance with approved procedures and in conformance

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with Technical Specifications.

2B diesel generator monthly operability test The following items were considered during this review:

the limiting conditions for operation were met while'affected components or systems were removed from and restored to service; approvals were obtained prior to initiating the testing; testing was accomplished in accordance with approved procedures; test instrumentation was within its calibration interval; testing was accomplished by qualified personnel; test results conformed with Technical Specifications and procedural requirements and were reviewed by personnel other than the individual directing the test; and any deficiencies identified during the testing were properly documented, reviewed, and resolved by appropriate management personnel.

No violations or deviations were identified.

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Monthly Maintenance Observation (62703)

Station maintenance activities of the safety-related. systems and components listed below were observed or reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications.

Repair of cartridge filter holder for radiation monitor ORE-PR033 Repair of cartridge filter holder for radiation monitor ORE-PR035 Following completion of maintenance on the radiation monir. ors, the inspector verified that they had been returned to service properly.

The following items were considered during this review:

the limiting conditions for operation were met while components or systems were removed from and restored to service; approvals were obtained prior to-initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or

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calibrations'were performed prior to returning. components or: systems to service; quality control-records were maintained;. activities.weref

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accomplished by qualified personnel; parts and materials ~used were properly' certified; radiological. controlswere implemented; and fire;

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. prevention controls were implemented. ' Work ~ requests were reviewed to L

determine the. status of outstanding jobs and to a'ssure'that priority is assigned to safety-_related equipment maintenance;which may.' affect system!

performance.

No violations or deviations were identified.

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Operational Safety Verification (71707,-71709, &.71881)

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L The inspector observed control room operation, reviewed applicable ~ logs and conducted discussions with control room operators during September 1987. During these discussions-and observations,-the inspector-ascertained that the operators were alert, cognizant of plant conditions, and attentive to changes in those conditions, and that they took prompt

action when appropriate.

The'. inspectors' verified the operability of.

selected emergency systems, reviewed tagout records'and verified the proper return to service of affected components.

Toursfof the auxiliary, N

fuel-handling, rad-waste, and turbine buildings were' conducted to observe -

plant equipment conditions, including potential' fire hazards,. fluid '

.h leaks, and excessive vibrations, and to_ verify that' maintenance requests had been initiated for equipment in need of maintenance.

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The inspectors verified by observation and direct interviews that the j

physical security plan was being' implemented in accordance with the i

station security plan.

The inspectors observed plant housekeeping / cleanliness conditions and

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verified implementation of radiation protection controls..The in'spectors also witnessed portions of the radioactive waste system' controls asacciated with rad-waste shipments and barreling.

During'the month of

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September 1987, the inspector walked down the accessible portions of the 1A and 2A auxiliary feedwater systems to verify operability.

l Facility operations were observed to verify that they were in accordancel with the requirements. established under Technical Specifications,10 CFR, i

and administrative procedures. During ar inspection by region III persennel, the inspector was informed of the existence of a test manifold in line IVQ02A4B of the containment purge-(VQ)'. system,'immediately downstream of containment. isolation valve IVQ002B, which was not part of.

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the permanent facility' design.

Further investigation by the inspector'

revealed that the test manifold had been installed on' November 11,- 1985,.

l to support a surveillance test of the VQ system. The surveillance test,

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1BVS VQ-4, injects a "00P" test aerosol into the ductwork upstream of, filter media, to perform in place penetration testing. When the test was terminated, the test manifold remained installed;-however, the-required'

safety evaluation and administrative procedures were not followed.

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10 CFR 50.59(a)(1) requires that the licensee.obtain commission approval,

prior to making changes in the facility.as described in the safety

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analysis report, if the proposed change involves a change in the Technical-Specifications or involves an unreviewed safety question.

10 CFR 50.59(b)(1) requires that the licensee maintain records-of'all changes to the facility made pursuant to this section, to the extent that these changes constitute changes in the facility as described in the safety analysis report.

These records must include a written safety evaluation which provides the basis for the determination that the change

does not involve an unreviewed safety question.

The VQ system is described in section 9.4.9 of the Byron FSAR. A schematic of the system is provided in FSAR figure 9.4-11.

The current piping and installation drawing of the VQ system (P&ID'M-105, Rev. AB)

i does not indicate the presence of the test manifold.

j Byron administrative procedure BAP 330-2, " Temporary Alterations,"

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Paragraph A, requires that temporary alterations (changes). to all plant

'l systems be performed in accordance with this procedure.

However, as an

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exception, temporary alterations may be made by other procedures (e.g.,

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surveillance tests) that meet certain notification and verification l

requirements, provided that control of the temporary alteration is j

transferred to BAP 330-2 if the alteration must stay installed following (

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completion of the procedure.

BAP 330-2T1, " Temporary Alteration Log Sheet," implements the requirements of 10 CFR 50.59(b)(1) and BAP 330-2.

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The failure to follow the administrative control requirements of BAP.

330-2 and the failure to complete a written safety evaluation (BAP 330-2T1) to determine that the continued installation of the test manifold in the VQ line did not constitute an unreviewed safety question are a violation of 10 CFR 50.59 (454/87038-02(DRP)).

In the response the licensee should address whether any other test equipment is " permanently" t

installed in any plant systems without the required 10 CFR 50.59 reviews.

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10.

Followup of Region III Request _s (92701)

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(Closed) Information Notice (IN) (87008): Degraded motor leads in'

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limitorque DC motor operators.

The IN describes problems with limitorque

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DC motor operators made by the Peerless-Winsmith Company, in the degradation of the motor leads and possible short circuit failure. The licensee has determined that it does not have any Peerless-Winsmith DC

motors installed at Byron.

Based on this information, this-IN is j

considered closed.

l 11. Allegation Followup (99014)

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l (Closed) Allegation (RIII-87-A-0047):

On April 15, 1987, the Region l

III office received an anonymous allegation of excessive overtime i

being worked U an individual involved in data coll'ction and analysis.

e The alleger suced that the individual worked for a subcontractor of the l

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licensee and was working six 12-hour days or.72-hours per week and that for some of the work the individual was using a cumputer for data

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reduction.

The alleger was concerned that this amount of overtime was j

excessive.

Based on the sketchy details of the allegation and the j

ongoing activities at Byron at that time, it is inferred that the-

individual was collecting data for the Unit 2 startup. testing program.

NRC Review:

The inspector obtained from the licensee a list of personnel who performed data collection, reduction, and analysis for the startup

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testing program.

The inspector reviewed the licensee's records to

determine the number of hours these individuals worked for a randomly j

selected period (April 1-15,1987).

l The records indicated that while individuals had occasionally worked

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longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in a day, there was no indication of an 1ndividual

working for greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per day for 6 days per week. The i

greatest number of hours worked by an individual in'one day was 14.8.

The number of hours worked by the test personnel were not restricted by regulation.

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The test results were reviewed by onsite and offsite personnel, so that any potential deficiencies resulting from fatigue were subject to-an d

independent review. The licensee has completed the startup test program, j

i the services of these contractors have been terminated, and these personnel have left Byron station.

Conclusion:

This allegation is not substantiated. A random sample of the working hours for individuals who were performing these activities i

did not indicate that excessive overtime (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per day, 6 days per week) was being worked. Therefore, this allegation is considered closed.

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12. Management Meetings (30702)

On October 1, 1987, Messrs. L. N. 01shan, License Project Manager, W. L. Forney, Chief, Reactor Projects Branch 1, J. M. Hinds, Jr., Chief, Reactor Projects Section 1A, and the NRC resident inspector staff met

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with the licensee management and supervisory personnel denoted in Paragraph 1 of this report. This meeting was held to discuss NRC concerns related to the licensee's recent performance in the operations, surveillance, and quality assurance and administrative controls func-tional areas and to review the status of outstanding items and pending i

modifications.

13.

Violations for which A " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation'as a standard method for formalizing the existence of a violation of a legally binding requirement.

However,

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because'the NRC wants to encourage and support licensee initiatives

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for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the tests of 10 CFR 2, Appendix C, Section V.A.

These tests are:

(1) the violation was identified by the licensee; (2) the violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if required; (4) the violation will be corrected, including l

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-measurestoprevent. recurrence,withintaireasonabletime-period;iand.(5)L it was not a violation that could reasonably,be: expected to have:been '

prevented by'the licensee's corre~tive action for a' previous violation..

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Violations lof; regulatory requirements (identified duringithe: inspection -

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.for which a. Notice;of' Violation will not be issued'are, discussed:in Paragraphs 3.a and 3.b.

14.

Exit' Interview (30703)

The inspector met with the. licensee! representatives' denoted?in paragraph:

I at the conclusion.of the. inspection on'0ctober 1, 1987. 'The inspectore

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summarized the purpose and scope lofethe inspection and1the findings.L

~ The inspector also discussed.the 'likely ' informational. content of t the-

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t inspection report, with regard to documents or processes reviewed:by the-inspector during the inspection.

The licens'eeldid not~ identify.any such documents or processes as proprietary.

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