ML20137X384

Insp Repts 50-324/85-27 & 50-325/85-27 on 850801-31. Violation Noted:Bolts Replaced on Hydraulic Control Units W/Type Other than Specified on Drawings

ML20137X384
Person / Time
Site:	Brunswick
Issue date:	09/20/1985
From:	Fredrickson P, Garner L, Hicks T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20137X354	List: ML20137X349 ML20137X358 ML20137X384
References
50-324-85-27, 50-325-85-27, NUDOCS 8510040510
Download: ML20137X384 (11)

See also: IR 05000324/1985027

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pn Rico UNITED STATES

f 'o NUCLEAR REGULATORY COMMISSION

/ ,^S REGION 11

Eet'I *$ k I

101 MARIETTA STREET. N.W.

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f ATLANTA, GEORGI A 30323

% .' ,',#,. * SEP 2 31985

Report Nos. 50-325/85-27 and 50-324/85-27

Licensee: Carolina Power and Light Company

P. O. Box 1551

Raleigh, NC 27602

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Docket Nos.: 50-325 and 50-324 License Nos. DPR-71 and DPR-62

Facility Name: Brunswick 1 and 2

Inspection Co ed: ugust 1 ; 31,1985

Inspectors: -

\ W [

h L. arner _ Vate figned

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h T. E. 'cks D6te Sfgned

Approved By: - / 7 [fd'

P. E. Fredrickson, Section Chief Dite Signed

Division of Reactor Projects

SUMMARY

Scope: This routine safety inspection involved 177 inspector-hours on site in

the areas of maintenance observation, surveillance observation, operational

safety verification, onsite review committee, ESF System walkdown, Licensee Event

Reports review, follewup on inspector identified items, refueling activities and

plant modifications.

Results: One violation was identified: Bolts Replaced on Hydraulic Control

Units with Type Other Than That Specified on Drawings. One unresolved item was

identified: Seismic Qualification of Hydraulic Control Unit Frame.

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G ADOCK 05000324

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REPORT DETAILS

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1. Licensee Employees

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Persons Contacted

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P. Howe, Vice President - Brunswick Nuclear Project

C. Dietz, General Manager - Brunswick Nuclear Project

, T. Wyllie, Manager - Engineering and Construction

G. Oliver, Manager - Site Planning and Control

! J. Holder, Manager - Outages

l E. Bishop, Assistant to General Manager I

i L. Jones, Director - QA/QC r

M. Shealy, Acting Director - Training ,

M. Jones, Acting Director - Onsite Nuclear Safety - BSEP

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J. Chase, Manager - Operations

J. O'Sullivan, Manager - Maintenance

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G. Cheatham, Manager - Environmental & Radiation Control

4 K. Enzor, Director - Regulatory Compliance

B. Hinkley, Manager - Technical Support

L. Boyer, Director - Administrative Support

V. Wagoner, Director - IPBS/Long Range Planning

L C. Blackmon, Superintendent - Operations

j J. Wilcox, Principle Engineer - Operations

W. Hogle, Engineering Supervisor

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W. Tucker, Engineering Supervisor

B. Wilson, Engineering Supervisor

R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)

J. Moyer, I&C/ Electrical Maintenance Supervisor (Unit 1)

R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

R. Poulk, Senior.NRC Regulatory Specialist

D. Novotny, Senior Regulatory Specialist

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W. Dorman, QA - Supervisor

W. Hatcher, Security Supervisor

. W. Murray, Senior Engineer - Nuclear Licensing Unit

Other licensee employees contacted included construction craftsmen,

engineers, technicians, operators, office personnel, and security force

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members.

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2. Exit Interview (30703) i

The inspection scope and findings were summarized on September 5, 1985 with

the general manager. The licensee acknowledged the findings without

i exception. The licensee did not identify as proprietary any of the

j materials provided to or reviewed by the inspectors during the inspection.

3. Followup on' Previous Enforcement Matters (92702)

Not inspected.

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4. Followup on Inspector Followup Items

a. (Closed) Inspector Followup Item 325, 324/84-35-02; Post Trip Reviews

01-22. An open item was generated after a reactor scram in December

1984 because of a questionable post trip review. The item was to

follow the licensee's progress toward enhancing the scram review

process. Revisions 8-10 to the Post Trip Review Procedure 01-22 were

the result of this effort. These revisions clarify the responsibili-

ties of the operations engineer and other associated requirements.

This item is considered closed.

b. (Closed) Inspector Followup Item 324/84-31-03; Standby Air

Compressors. This item was opened because of a concern over the

inoperability of an automatic start pressure switch for a standby air

compressor. The licensee has failed to find a suitable replacement,

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but is presently undertaking a plant modification on Unit 1 (and is

planned on Unit 2) which will alleviate the need for these air

compressors during post accident conditions. The modification will

install a nitrogen backup system which will supply the necessary

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pneumatic pressure during accident conditions. This item is considered

! closed.

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c. (Closed) Inspector Followup Item 325, 324/85-03-01; Radwas*e Shipping.

This item was generated to track improvements in the radwaste shipping ,

quality control program and the auditing of this program. This item is

considered closed because a notice of violation 325, 324/85-17-01 was

written in this area and will track corrective actions.

5. Maintenance Observation (62703)

The inspectors observed maintenance activities and reviewed records to

verify that work was conducted in accordance with approved procedures,

Technical Specifications, and applicable industry codes and standards. The

inspectors also verified that: redundant components were operable;

j administrative controls were followed; tagouts were adequate; personnel were

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qualified; correct replacement parts were used; radiological controls were

proper; fire protection was adequate; QC hold points were adequate and

observed; adequate post-maintenance testing was performed; and independent

verification requirements were implemented. The inspectors independently

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verified that selected equipment was properly returned to service.

Outstanding work requests and authorizations (WR&A) were reviewed to ensure

that the licensee gave priority to safety-related maintenance.

a. Bolting Replacement and Seismic Qualification of Hydraulic Contr^1

Units

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Inspection Report 325/85-22- issued a notice of violation for loose

and/or missing rack-support-to-foundation bolting for the control rod

! hydraulic control units (HCU). During followup of the repair and

replacement, it was observed by the inspector that all the replaced

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bolts (five) had been replaced with bolts which were not cadmium plated

as required by plant design drawing G.E. 919D615. The apparent root

, cause was that the maintenance planner verified that Q bolting was

required but failed to check the specifications for any special

requirements. Discussion with licensee personnel indicates that this

may have been a common practice when replacing bolting. The licensee

is currently evaluating the impact of this on plant equipment.

Failure to install the type of bolting specified on G.E. 9190615

drawing is a violation of 10 CFR 50, Appendix B, Criterion V

(324/85-27-01): Bolts Replaced on Hydraulic Control Units with Type

Other than that Specified on Drawings.

The licensee has attempted to establish the seismic qualification of

the as-found conditions documented in inspection report 325/85-22,

i.e., one HCU had 2 out of the 4 rack-support-to-foundation bolts

missing. Calculations were performed on the HCU's with missing bolts,

on HCU's which stand alone and on HCU's installed back-to-back with all

fasteners properly installed. The calculations show the as found HCU's

with loose or missing bolts met at least short term criteria (IEB

79-14), i.e., bolts might deform but would not break. The same

conclusion was deterrrined for the stand alone HCU's. However, in all

cases including back-to-back installation, stresses were calculated

which exceeded the allowable stress in the tubular frame. The original

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seismic qualification was performed by the vendor (G.E.) based upon

results of field tests with the units tested back-to-back. This field

test data is not available to the licensee at this time. Without field

test data, the complexity of the installed configuration requires

several conservative assumptions to be made to allow analytic modeling.

The licensee believes that their calculated results are conservative.

Therefore, since the frame was qualified by the vendor from experi-

mental data, the licensee believes the frames to be qualified and the

HCU's are operable, i.e. seismically qualified. However, the licensee

expects to resolve this apparent discrepancy between their analytic

model and the original seismic qualification based on field data.

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Resolution of this apparent discrepancy is an unresolved item

(325/85-27-01 and 324/85-27-02): Seismic Qualification of HCU Frame,

b. Post Maintenance Test Requirement Test Sheet Fails to Specify Pressure

Test /VT-2 Inspection

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During a routine inspection of post maintenance surveillance testing

for the Unit I standby liquid control injection check valve C41-F007,

the inspector noticed that no pressure test /VT-2 inspection was

specified on the Post Maintenance Test Requirement (PMTR) sheet even

though the pressure boundary for the valve had been broken during

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maintenance. The valve is Class 1. Two similar maiatenance activities

involving Class 1 valves B21-F028B and B21-F019 were also reviewed and

found to not contain the pressure test /VT-2 inspection requirement on

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the PMTR sheet. It became clear that the maintenance planners were not

using ENP-16, Inservice Inspection Requirements, properly. After

scoping the jobs and realizing that the work involved disassembly of

the valves, the planners should have realized that the pressure

boundaries of each of these valves was to be broken and that entry into

Section VI, Visual Inspection, of ENP-16 was necessary to determine

additional post maintenance test requirements. This process was not

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done.

i The problem was discussed with plant management and the following

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immediate corrective actions were taken:

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(1) Training was conducted for all mechanical maintenance planners in

the proper use of ENP-16.

(2) A review was conducted of PMTR's for Class 1 valves worked during

the Unit 1 outage. For those which required pressure test /VT-2

inspections, an additional test / inspection was included on the

PMTR sheet. Those activities already closed out, were reopened by

an additional PMTR sheet stating the required test.

At the completion of the present Unit 1 outage, a vessel

hydrostatic test and inspection including all Class 1 piping and

components is to be performed (PT-80.1, 10 year Inservice

Inspection Reactor Vessel Hydrostatic Test). This test would have

satisfied the inservice inspection requirements for the valves

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identified. Followup of long term corrective actions will be an

inspector followup item (IFI 325/85-27-02): Administrative

!. Control Changes to Ensure Pressure Test /VT-2 Inspections are

Identified as Post Maintenance Requirements.

f 6. Surveillance Observation (61726)

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) The inspectors observed surveillance testing required by Technical

Specifications. Through observation and record review, the inspectors

verified that: tests conformed to Technical Specification requirements;

administrative controls were followed; personnel were qualified;

instrumentation was calibrated; and data was accurate and complete. The

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inspectors independently verified selected test results and proper return to

service of equipment.

A special review was performed of the following Licensee finding:

On August 21, 1985, the Maintenance Surveillance Test (MST) rewrite group

discovered that Periodic Tests, PT-A22.2-1, PT-22.2-2, PT-A24.2 and

PT-45.2.4, covering Secondary Containment Isolation Response Time Testing,

did not adequately test all the relays in the associated logic circuit.

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Technical Specification Surveillance 4.3.2.3 requires that this be done

every 18 months.

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Specifically, relays K66, K67, 3AA, 3AB, 3BA, 3BB, 3BD, A-CRMX and B-CRMX

,/ were not being included in the response time test. At the time, Unit I was

shutdown for a refueling outage (no core alterations were in progress) and

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Unit 2 was at 100% power. The surveillance test problem involved both

units.

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Licensee management conducted a Plant Nuclear Safety Committee (PNSC)

meeting concerning this problem and concluded that there was no technical

reason to consider this instrumentation inoperable based on the following:

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a. Relays not presently being timed have been verified operable in logic

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system functional tests which were performed in October 1984 (Unit 2).

b. The manufacturer's expected response time for these relays is less than

85 milliseconds.

c. The allowed response time for the instrumentation is less than or equal

to 13 seconds. Adding the relay's expected response time to the

existing instrumentation response time still results in a response time

of less than or equal to 1 second.

A special test procedure was generated to test the relays (SP-85-086) and

'was satisfactorily performed for Units i and 2 on August 25, 1985.

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These inadequate procedures constitute a violation of Technical Specifi-

i. cation Surveillance 4.3.2.3,.in that they failed to adequately response time

1, test all the necessary relays. However,10 CFR i Appendix C,Section V,

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paragraph A, states that a notice of violation will generally not be issued

if a violation meets 5 stated criteria. This violation meets these criteria

and no notice of violation will be issued.

, A permanent procedure to conduct the testing will also be written and

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implemented by the MST rewrite group prior to the end of the next

surveillance interval.

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No violations or deviations were identified.

7. Operational Safety Verification (71707) (71710)

The inspectors verified conformance with regulatory requirements by direct

observations of activities, facility tours, discussions with personnel,

reviewing of records and independent verification of safety system status.

The inspectors verified that control room manning requirements of 10 CFR

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50.54 and the Technical Specifications were met. Control room, shift

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supervisor, clearance and jumper / bypass logs were reviewed to obtain

information concerning operating trends and out of service safety systems to

ensure that there were no conflicts with Technical Specifications Limiting

j Conditions for Operations. Direct observations wera conducted of control

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room panels, instrumentation and recorder traces important to safety to

verify operability and that parameters were within Technical Specification

limits. The inspectors observed shift turnovers to verify that continuity

of system status was maintained. The inspectors verified the status of

selected control room annunciators.

J Operability of a selected ESF train was verified by insuring that: each

accessible valve in the flow path was in its correct position; each power

supply and breaker, including control room fuses, were aligned for

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components that must activate upon initiation signal; removal of power from

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those ESF motor-operated valves, so identified by Technical Specifications,

was completed; there was no leakage of major components; there was proper

lubrication and cooling water available; and a condition did not exist which

might prevent fulfillment of the system's functional requirements.

Instrumentation essential to system actuation or performance was verified

operable by observing on-scale indication and proper instrument valve

lineup, if accessible.

The inspectors verified that the licensee's health physics policies / pro-

cedures were followed. This included a review of area surveys, radiation

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work permits, posting, and instrument calibration.

The inspectors verified that: the security organization was properly manned

and that security personnel were cepable of performing their assigned

functions; persons and packages were checked prior to entry into the

protected area (PA); vehicles were properly authorized, searched and

escorted within the PA; persons within the PA displayed photo identification

badges; personnel in vital areas were authorized; effective compensatory

measures were employed when required; and security's response to threats or

alarms was adequate.

The inspectors also observed plant housekeeping controls, verified position

of certain containment isolation valves, checked clearances, and verified

the operability of onsite and offsite emergency power sources.

No violations or deviations were identified.

8. Onsite Review Committee (40700)

The inspectors attended selected Plant Nuclear Safety Committee meetings '

conducted during the period. The inspectors verified that the meetings were

conducted in accordance with Technical Specification requirements regarding

quorum membership, review process, frequency and personnel qualifications.

No violations or deviations were identified.

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9. Onsite Review of Licensee Event Reports (92700)

The listed Licensee Event Reports (LER's) were reviewed to verify that the

information provided met NRC reporting requirements. The verification

included adequacy of event description and corrective action taken or

planned, existence of potential generic problems and the relative safety

significance of the event. Onsite inspections were performed and concluded

that necessary corrective actions have been taken in accordance with

existing requirements, licensee conditions and commitments. The following

reports are considered closed:

(Closed) LER 1-80-20; Containment monitoring system isolated due -to

personnel error.

(Closed) LER 1-81-34; Four supports were found damaged due to water hammer.

(Closed) LER 1-83-10; Fire barrier / secondary containment seal degradation

allows water to leak into reactor building.

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(Closed) LER 1-83-23; Fire in 4160/480 volt E-6 transformer.

(Closed) LER 1-83-26; Diesel generator trips due to operator failing to

follow procedure.

(Closed) LER 1-83-32; Control rods have no position indication.

(Closed) LER 1-83-36; Control power fuse to motor operator blew due to

ground in circuit.

(Closed) LER 1-83-40; Inadequate surveillance procedure and personnel error

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cause HPCI to isolate.

(Closed) LER 1-83-62; Reactor building exhaust ventilation radiation monitor

actuated outside technical specification limit.

(Closed) LER 1-84-01; Air entrapped in suction header caused residual heat

removal service water pumps to trip.

! (Closed) LER 1-84-29; Spurious actuation of control building emergency air

i filtration system.

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(Closed) LER 1-84-30; Spurious actuation of control building emergency air

filtration system.

(Closed) LER 1-84-31; Spurious actuation of control building emergency air i

filtration system.

(Closed) LER 1-85-02; Automatic isolation of the control building heating,  ;

ventilation and air conditioning system due to spurious chlorine signal. l

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(Closed) LER 1-85-03; Inadequate logic system functional testing of degraded

and under-voltage relays of emergency buses.

(Closed) LER 1-85-05; Inadequate functional testing of rod block monitor.

(Closed) LER 1-85-07; Spurious actuation of control building emergency air

filtration system.

(Closed) LER 1-85-14; Steam leak causes reactor core isolation cooling

system isolation.

(Closed) LER 1-85-17; Loss of emergency bus E-1 normal feed.

(Closed) LER 1-35-19; Control building emergency air filtration system

actuation due to accidental shorting of detector circuit.

(Closed) LER 1-85-22; Standby gas treatment train IA relay over heats and

fails.

(Closed) LER 2-83-41; A " rod drift" annunciation was received due to the rod

position indication probe.

(Closed) LER 2-83-46; Wires to terminals in the terminal box were reversed.

(Closed) LER 2-83-52; Transmitter shorted to ground when alligator clip test

leads were accidentally bumped.

(Closed) LER 2-83-62; Prima ry containment temperature exceeds technical

specification limit as result of seasonal ambient temperatures.

(Closed) LER 2-83-65; Condensate storage tank level switches for High

Pressure Coolant Injection (HPCI) system improperly installed.

(Closed) LER 2-83-71; Suppression pool temperature exceeds limit as result

of HPCI run.

(Closed) LER 2-83-82; Suppression pool level exceeds limit due to personnel

error.

(Closed) LER 2-83-84. This item was voided by the Licensee.

(Closed) LER 2-83-94; Reed switch problems cause incorrect position

indication.

(Closed) LER 2-84-04; Reactor scram initiated by HFA relay replacement and

surveillance testing.

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(Closed) LER 2-84-06; Procedure failed to identify the need for jumping the

low low level signal.

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(Closed) LER 2-85-01; Surveillance test not performed within allowable time

due to inadvertent deletion from scheduling system.

(Closed) LER 2-85-03; Inadequate surveillance procedure results in an

unexpected group 1 isolation.

No violations or deviations were identified.

10. Refueling Activities (60710)

During the licensee's refueling operations, the inspectors verified that

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selected surveillance testing required by Technical Specifications was

! current and that the licensee's fuel- handling procedure was implemented.

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The following additional items were verified:

a. Selected fuel bundle movements,

b. Core monitcring during refuel operations was in accordance with

Technical Specifications.

c. Vessel water level was maintained in accordance with Technical

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Specification. ,

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d. Reactor mode switch position was as required by Technical Specifi-

cation.

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e. Continuous communications were maintained between the refueling ,

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platform and the control room and that . control room operators were

cognizant of the applicable procedure steps.

f. Health-Physics personnel maintained constant coverage of all fuel

j moving activities, ensuring area dose rates, contamination levels and

airborne samples were within required tolerances.

No violations or deviations were identified.

11. Modification Process and Masonry Walls (37700) i

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During review of design activities, the inspector reviewed the licensee's

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method for ensuring that safety related equipment would not be installed on

j- or in close proximity to masonry walls whose failure could affect the

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equipment (Bulletin 80-11 concern). The method currently employed relies

upon -notes on the wall drawings to inform the user of the wall status

(analyzed or unanalyzed) and their general engineering practice of. not

installing new supports onto masonry walls. These appear to adequately

address installation onto walls but do not provide a positive means to

j ensure that safety related equipment is not placed close to an unanalyzed

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wall (walls which did not have safety related equipment around them were not

analyzed per Bulletin 80-11). This is an inspector followup item

(324/85-27-03 and 325/85-27-03): Enhancements of controls to preclude

installation of safety related equipment in proximity to unanalyzed masonry

walls.

No violations or deviations were identified.