ML20153G718

Insp Repts 50-324/88-14 & 50-325/88-14 on 880301-31. Violations Noted.Major Areas Inspected:Followup on Enforcement Matters,Maint Observation,Surveillance Observation & Operational Safety Verification

ML20153G718
Person / Time
Site:	Brunswick
Issue date:	05/05/1988
From:	Fredrickson P, Ruland W, Tobin W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20153G699	List: ML20153G696 ML20153G710 ML20153G718
References
50-324-88-14, 50-325-88-14, NUDOCS 8805110322
Download: ML20153G718 (14)

See also: IR 05000324/1988014

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' '4farg# UNITED STATES

'o NUCLEAR REGULATORY COMMISSION

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. TA REGION 11

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101 MARIETTA STRE ET, N.W.

ATL ANT A, G EORGI A 30323

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Report No. 50-325/88-14 and 50-324/88-14

Licensee: Carolina Power and Light Company

P. O. Box 1551

Raleigh, NC 27602

Docket No. 50-325 and 50-324 License No. DPR-71 and DPR-62

Facility Name: Brunswick 1 and 2

Inspection Conducted: March 1 - 31, 1988

Inspector: W ! f on s ,.* : / / )

W. H. Ruland g["-

k f/$88

Date' Signed

Accompanying Personnel: S. M. Shaeffer

R. B. Latta

E. H. Girard

Inspector:

(Paragraph 10 . .

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Date' Signed

Approved By: J .

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P. E. Fredrickson, Section Chief

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Bate' Signed

Reactor Projects Section 1A

Division of Reactor Projects

SUMMARY

Scope: This routine safety inspection by the resident inspector and others

involved the areas of followup on previous enforcement matters, maintenance

observation, surveillance observation, operational safety verification,

in-office Licensee Event Report (LER) review, followup on inspector identified

and unresolved items, plant modifications, allegation followup, and unusual

events.

Results: Three Unit i violations were identified: Failure to determine reactor

vessel pressure and shell temperature every 30 minutes during plant heatup;

failure to verify position of an unlocked manual CAD valve in its correct

position every 31 days; and failure to perfonn a 10 CFR 50.59 evaluation.

A potentially significant safety issue was identified by the licensee regarding

operator inattention during shutdown conditions. A licensed operator

unknowingly allowed reactor coolant system temperature increase 90 degrees.

This item was identified as a licensee-identified violation.

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

W. Biggs, Principal Engineer

• E. Bishop, Manager - Operations

• J. Brown, Res. Engineer - Engineering

• S.

Callis, Onsite

T. Cantebury, Licensing

Mechanical Engineer -Supervisor

Maintenance Licensing Unit

(& Nuclear

1) Fuel

G. Cheatham, Manager - Environmental & ladiation Control

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

• C, Dietz, General Manager - Brunswick Nuclear Project

W. Dorman, Supervisor - QA

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• R. Eckstein, Manager - Technical Support

• K. Enzor, Director - Regulatory Compliance

R. Groover, Manager - Project Construction

W. Hatcher, Supervisor - Security

A. Hegler, Superintendent - Operations

R. Helme, Director - Onsite Nuclear Safety - BSEP

J. Holder, Manager - Outages

• P. Howe, Vice President - Brunswick Nuclear Project

• L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)

• M. Kesmodel, Supervisor - Document Control

R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)

J. Moyer, Manager - Training

G. Oliver, Manager - Site Planning and Control

J. O'Sullivan, Manager - Maintenance

B. Parks, Engineering Supervisor

• R. Poulk, Senior NRC Regulatory Specialist

• D. Queener, Principal Engineer - Environmental & Radiation Control

J. Smith, Manager - Administrative Support

• J. Titrington, Principal Engineer - Operations

V. Wagoner, Director - IPBS/Long Range Planning

• R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)

D. Warren, Acting Engineering Supervisor

B. Wilson, Engineering Supervisor

• T. Wyllie, Manager - Engineering and Construction

Other licensee employees contacted included construction craf tsmen,

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

members.

• Attended the exit interview

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2. Exit Interview (30703)

The inspection scope and findings were summarized on March 31, 1988, with

those persons indicated in paragraph 1. The inspectors described the

areas inspected and discussed in detail the inspection findings listed

below. Dissenting comments were not received from the licensee.

Proprietary information is not contained in this report.

Item Number Description / Reference Paragraph

325/88-14-01 VIOLATION - Missed Surveillance of Reactor Vessel

Temperature and Pressure During Inadvertent

Heatup (paragraph 8.a). (A licensee identified

violation was also reviewed.)

325/88-14-02 VIOLATION - Surveillance Procedure Failed to

Check a CAD Valve Position (paragraph 8.c).

325/88-14-03 VIOLATION - Failure to Perform 10 CFR 50.59

Evaluation On Service Water System (para-

graph 8.b).

325,324/88-14-04 IFl - Correction of CAD System Hardware Problems

(paragraph 8.c).

Note: Acronyms and abbreviations used in this report are listed in

paragraph 12.

3. Followup on Previous Enforcement Matters (92702)

(CLOSED) Violation 325/86-29-01; Improper Temporary Revision to OP-03

When Operating EPA Breakers, response dated January 2, 1987. The

inspectors reviewed OP-03, Revision 8 for Unit 1 and OP-03, Revision 19

for Unit 2 and detarmined that these procedures had been modified to

provide for the opening of the EPA breakers prior to opening the MG set

output breaker. These revisions adequately address the issue identified

in the violation. This item is closed.

(CLOSED) Violation 325/87-31-01 and 324/87-35-01; Inadequate DG Surveill-

ance Procedure - TCV Not in Restoration Lineup, response dated November

25, 1987. The inspectors reviewed PT-12.2A, Revision 32, PT-12.2B,

Revision 31, PT-12.2C, Revision 31, and PT-12.2D, Revision 33 and deter-

mined that these procedures had been modified to include the subject

temperature control valves in the system restoration valve lineup.

Additionally, these procedures have been revised to provide operational

information regarding normal and abnormal valve operation and required

corrective action. The revised pts adequately address the issue identi-

fied in the violation. This item is closed.

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4. MaintenanceObservation(62703)

The inspectors observed maintenance activities, interviewed personnel, 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; administrative controls were followed; tagouts

were adequate; personnel were qualified; correct replacement parts were

used; radiological controls were proper; fire protection was adequate;

quality control hold points were adequate and observed; adequate post-

maintenance testing was performed; and independent verification require-

ments were implemented. The inspectors independently verified that

selected equipment was properly returned to service.

Outstanding work requests were reviewed to ensure that the licensee gave

priority to safety-related maintenance. '4umerous maintenance items were

reviewed throughout the reporting period. The inspectors observed /

reviewed, in detail, those portions of the following maintenance activi-

ties:

86-BAFF1 1-CAC-V10 Outboard Drywell Purge Exhaust Valve

88-AIYB1 DG No. 3 Oscillations / Troubleshooting

No significant safety matters, violations, or deviations were identified.

5. SurveillanceObservation(61726)

The inspectors observed surveillance testing required by Technical Speci-

I fications. Through observation, interviews, and record review, the

l 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 inspectors independently verified selected test results and

proper return of equipment to service.

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The inspectors witnessed / reviewed portions of the following test activi-

ties:

1MST-APRM12W APRM Channel B, D, and F Functional Test

IMST-BATT11Q Batteries, 125 VDC Quarterly Operability Test

IMST-HPCl22M Steamline Low Pressure Channel Calibration

IMST-HPCI26M HPCI Suppression Pool High Level Instrument Channel

Calibration

2MST-HPCl27M HPCI and RCIC CST Low Water Level Instrument Channel

Calibration

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2MST-RPS23R RPS High Reactor Pressure Instrument Channel Calibra-

tion

2MST-SEIS21R SMA-3 Seismic Monitoring System Channel Calibration

No significant safety matters, violations,.or deviations were identifie6

6. Operational Safety Verification (71707)

The inspectors verified that Unit 1 and Unit 2 were operated in compliance

with Technical Specifications and other regulatory requirements by direct

observations of activities, facility tours, aiscussions with personnel,

reviewing of records and independent verification of safety system status.

The inspectors verified that control room manning requirements of 10 CFR

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

supervisor, clearance, STA, daily and standing instructions, 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 Conditions for Opera-

tions. Direct observations were conducted of control room panels, instru-

mentation, and recorder traces important to safety to verify operability

and that operating 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.

Operability of a selected Engineered Safety Feature division was verified

weekly by ensuring that: each accessible valve in the flow path was in

its correct position; each power supply and breaker was closed for

components that must activate upon an initiation signal; the RHR subsystem

cross-tie valve for each unit was closed with the power removed from the

valve operator; 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 /

procedures were followed. This included observation of HP practices and a

review of area surveys, radiation work permits, posting, and instrument

i calibration.

The inspectors verified that: the security organization was properly

manned and security personnel were capable of performing their assigned

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

protected area; vehicles were properly authorized, searched and escorted

within the PA; persons within the PA displayed photo identification

badges; personnel in vital areas were authorized; and effective compen-

saton measures were employed when required.

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The inspectors also observed plant housekeeping controls, verified

position of certain containment isolation valves, checked a clearance, and

verified the operability of onsite and offsite emergency power sources.

No significant safety matters, violations, or deviations were identified.

7. In Office Licensee Event Report Review (90712)

The listed LER was reviewed to verify that the information provided met

NRC reporting requirements. The verification included adequacy of event

description and corrective action taken or planned, existance of poter,tial

generic problems and the relative safety significance of the event.

(CLOSED) LER 2-88-03; Auto-Isolation of Reactor Water Cleanup System

Inlet Outboard Isolation Valve 2-G31-F004 With Reactor Defueled. The

inspector reviewed the event and had no questions. This iten is closed.

No significant safety matters, violation,, or deviations were identified.

8. Followup on Inspector Identified and Unresolved Items (92701)

a. (CLOSED) Unresolved Item 325/88-01-03; Inadvertent Heatup During

Cold Shutdown. The inspector reviewed the OER which described the

event and documented the licensee's root cause determination and '

corrective actions. The licensee found two causes for the event:

(1) operator inattentiveness; (2) the SR0 directed the control  ;

operator to throttle the heat exchanger outlet valve, 1-E11-F003B, to

control temperature. This valve has a key lock switch on the main *

control board that only has open and closed positions. The operator

had to place the switch in a "dead" position to throttle the valve.

This action, as well as throttling the F003B valve, was not

authorized by the system operating procedure. <

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The actions initiated by the licensee to prevent recurrence included

disciplinary action for the operators involved, on-shift training,  ;

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evaluation of switch replacement for the E11-F003B valve, development  !

of a routine pre-outage training package, proceduralization of key

l parameter monitoring frequency, and evaluation of the need for

l appropriate alarms during shutdown conditions.

The inspector reviewed real time training package 88-1-8 prepared for

j the event and found it acceptable. The licensee identified the

failure to follow the RHR operating procedure and met the require- .

ments of 10 CFR 2 Appendix C, for licensee identified violations. '

Therefore, no violation will be issued for the failure to follow

procedure. However, during the heatup, the licensee failed to

determine that the reactor vessel pressure and shell temperature were

. within limits every 30 minutes per TS 4.4.6.1.1. Since the licensee

I failed to identify the missed surveillance, a Violation is being

l Issued on this particular item: Missed Surveillance of Reactor

Vessel Temperature and Pressure During Inadvertent Heatup

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(325/88-14-01).

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b. (OPEN) Unresolved Item 325/88-05-01 and 324/88-05-01; Service Water

System Operating Mode Concerns. Further inspector icview revealed

that no 10 CFR 50.59 review was performed as a result of the problem

with the V106 valve prior to Unit I start-up on February 20, 1988.

FSAR table 9.2.1-1, Service Water Flow Distribution - One Reactor

Plant, lists Reactor Building Ct'J Heat Exchanger flow rate from the

nuclear service water header during the first 10 minutes following a

Loss of Coolant Accident as zero gallons per minute. Based on the

information that the licensee had received regarding how the service

water system could operate during the above time, it was possible

with V106 failed open, that flow through the RB CCW HX would be

greater than zero gallons per minute. Although no mechanical or

electrical components were modified, this information did constitute

a change to the Service Water System in that failure of the V106

valve would cause the system to not perform as described in the FSAR.

This change necessitated a 10 CFR 50.59 safety evaluation, prior to

restart of the unit, to determine whether an unreviewed safety

question existed. The licensee first became aware of this change in

the facility ir> February 1988. A 10 CFR 50.59 evaluation was not

performed until March 22, 1988, after being identified by the inspec-

tion in February 1988. Failure to perform a written 10 CFR 50.59

safety evaluation is a Violation: Failure to Perform 10 CFR 50.59

Evaluation On Service Water System (325/88-14-03).

The licensee had reviewed the issue at PNSC meeting 88-018 held on

February 12, 1988, and concluded that an operability concern did not

exist. Actior4 items were established to resolve the issue perman- l

ently. After questions by the inspector, the licensee completed a

JC0 with a 10 CFR 50.59 evaluation (EER-88-167) for the SW V106 issue

on March 23, 1988. The EER took credit for the licensee's interim

action of limiting RB CCW SW flow to 5,000 GPM. Based on the SW pump

curve, with SW flow assumed at 8,600 GPM, the total dynamic head of

the pump would be about 108 feet, equivalent to 48 PSIG. This is

above the 40 PSIG low header pressure autostart signal for another

NSW pump. This portion of the item remains unresolved pending review

by the licensee and inspector of the SW design data and a permanent

fix, if required.

c. (CLOSED) Unresolved Item 325/88-05-05 and 324/88-05-05; CAD System

Discrepancies. The associated issues were identified during a CAD

system walkdown, and are addressed below.

The inspector found valve CAC-V168 unlocked in the open (required)

position. All unlocked valves in the system's flow path shall be ,

demonstrated to be operable at least once per 31 days by verifying

valve position. Failure to observe the valve position is a violation

of TS 4.6.6.2.a.2. The missed surveillance was due to an inadequate

procedure, in that valve CAC-V168 does not appear on PT-16.1,

Rev.12, Section 7.1.3, System Operability Verification Checklist.

This valve is a flow path valve in that closure of the valve could

prevent the system from injecting nitrogen into the drywell or torus

after a single valve failure. The inspector reviewed the history of

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all pericdic test procedures associated with the CAD system for

accuracy and timeliness. No other surveillance problems were noted.

This is a Violation: Surveillance Procedure r ailed to Check a CAD  ;

VelvePosition,(325/88-14-02).

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On-February 24, 1988, the inspector observed that the loop A and 0

CAD vessel pressure indicators, PI-2703 and PI-2704, exceeded the

maximum allowable working pressure of the CAD vessel which is 100  ;

PS!J. The pressure indicators were above their maximum gauge  ;

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graduations-(greater than 100 PSIG). The process requirement for the

CAD system is between 80 and 90 PSIG. The CAD vessel had exceeded i

these values due to the evaporation of liquid nitrogen within the

vessel. The design of the system incorporates three separate

pressure relief mechanisms: (1) two pressure control valves.

(PCV-2706 and 2707) with build-up coils designed to maintain 85 psig

in the CAS tank; (2) PCV-2705 relieves pressure to at-osphere at 90

psig; and (3) pressure safety valve PS-V3 relieves to atmosphere at

100 psig. There was no evidence that any #f these functions were

operating within their setpoints. The licensee's engineering group

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has raceemer.ded that the relief setpoints be calibrated, and if not

within tolerance, the valves should be rebui't or exchanged.

The licensee showed the inspector the vendor's hydrostatic test

results. The test was performed at 176 PSIG. Thus, no safety

concern exists regarding the CAD +ank integrity. Since the licensee

had work requests written on the above problems and the safety

significanca is small, no violation occurred. The CAD system, -

(lthough addressed in Technical Specifications, is not necessary to  :

prevent any explcsive mixture in a pest-accident containment environ-

This was shown by generic analysis done by BWROGs for the

ment.

licensee's hydrogen recombiner rule response. The CAD system would ,

only be needed post-LOCA if the containment '.ad not been inerted.

This occurs only during reactor startups and shutdowns. l

Minor support problems were identified on the liquid nitrogen lines

surrounding the CAD vessel loop 8 end. Apparently the icis.g of the

nitrogen lines has caused two wood block supports to deteriorate &nd

i become ncn-functional. The inspector observed tha' another support

was left disconnected near valve BYCV-2716. The licensee has issued

work requests to correct the deficiencies. The licensee has also

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been informed of varimus housekeeping items such as ground cables

draped over nitrogen lines and debris in the control cabinets.

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Due to ., axtent >' the deficiences, the inspector will followup on  ;

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the -ns e's cr' '"e actions. This is an Inspector Followup

Itr l e- H on ,

' System Hardware Problems (325/88-14-04 and

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d. (CLO5ED) Inspector Followup Item 325/86-24-04 and 324/86-25-04;

_ OP-19 Revisions for HPCI CST Recirculation Mode. The licensee .

performed SP-66-080, HPCI System Operability Test, on October 9,

1986. The test checked the operation of the HPCI turbine control

system during three conditions: operating in CST full flow test mode;

only on the minimum flow valve; and a dual path mode. Problems with

the governor ccntrol were noted during certain step changes in the

flow demand signal made by the operator. The turbine RPM was hunting

with 1,000 RPM swings with a reduced flow demand signal of 3,000 GPM.

A portion of the governor, designated EGM, was retalibrated and

SP-86-080 re-performed with no anomalies noted. On November 8, 1986,

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the licensee replaced, under DR-86-0140, the EGM with a newer design,

further improving the control system.

The inspector reviewed SP-86-080, interviewed plant personnel, and

reviewed the current revision (13) of 1-0P-19, HPCI System Operations

Procedure. The licensee has added sections to-describe the reactor

pressure control mode of the HPCI system. Also, while switching

modes, the procedure requires the operator to take the HPCI flow

controller to manual. Based on the inspector's review of the l

licensee's actions, this item is closed. [

e. (CLOSED) Inspector Followup Item 325/87-42-08 and 324/87-43-08; ,

Submission of TS Amendment Request for RWCU Isolation Response Time.

The licensee submitted the request for TS amendment on February 29,

1988. This item is closed.

f. (CLOSED) Inspector Followup Item 325/88-05-03 and 324/88-05-03; Sand

Introduced Into Tcrus/ Vacuum Breakers From Failed Inerting Line. The

inspectors reviewed the circumstances of the line break and the

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adequacy of the corrective actions taken by the licensee. The review

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was conducted through visual observation of the vacuum breakers and

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the replacement inerting line, discussions with the Mechanical  :

Discipline Group Project Engineer, and review of the following Work

Request and LLRT records:

Unit 1 Work Requests

WR/JO 88-AELT1

WR/JO 88-AEIW1

WR/JO 88-AEIU1

jnit 2 Work Requests g

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WR/JO 88-AENW1

WR/JO 88-AFAC1  !

WR/JO 88-AAUH1

WR/JO 88-AAVII

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Unit 1 LLRTs

PT-20.3.71 performed 1/10/88 on V17

PT-20.?.70 performed 1/10/88 on V16

PT-20.3.71 performec' :','10/88 on V17

PT-20.3.70 performed 2/08/88 on V16

Unit 2 LLRTs

PT-20.3.71 performed 2/11/88 on V17

PT-20.3.70 performed 2/11/88 on V16

Relevant information and findings obtained by the inspectors:

(1) The inerting line failure was apparently the result of excessive

cooling and resultant stress in the line. The excessive cooling

was due to the introduction of insufficiently heated nitrogen

frem a malfunctioning vaporizer.

(2) S.nd

- did not reach tra Unit 2 vacuum breakers, but rather was

found on the inboard side of closed butterfly valves (V16 and

V17) located inboard of the vacuum breakers. Also, the sand

probably did not contribute to the LLRT failures. The butterfly

valves are normally closed and sand could not easily get onto

the seals. The project engineer stated that the butterfly

valves had a past history of excessive leakage and that they had

actually leaked less than in past tests. As of March 4, mainten-

ance on one valve had resulted in acceptable leakage and the

other showed leakage just in excess of the specified limit. The

identical Unit 1 valves passed the LLRTs and required no repair.

(3) The inerting line failure did not occur on January 4, as

initially thought. In order fer the sand to have reached Unit 2

valves V16 and V17, the failura would have had to occur on or

before the Unit 2 shutdown (about January 1,1988).

(4) An attempt to repair the uderground inerting line proved

unsuccessful. When an identical break area was repaired the

! line continued to exhibit leakage; indicating an additional

j break or breaks. Most of the underground line has now been

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bypassed. A temporary replacement line installed above ground

is now being used. (This is a non-safety-related line.)

(5) The sand that entered the Containment Atmosphere Control System

through the break in the inerting line was found at the

following locations:

Almost 2 quarts in the Unit i drywell near the 18-inch

inerting line opening (underground line was 8-inch size).

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• 810wn into torus of Unit 1;(from 20-inch line).

None found in drywell or torus of Unit 2.

~' - Approximately 2 cups of-sand / water slurry was found in the

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Unit 2 pipe between V16 and V17.

Approximately 2 gallons of slurry was found between Unit i

valves V16 and V17.

About a handful of dry sand near V5 (downstream side) on

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The inspectors had no further questions. This' item is closed.

g. (OPEN) Inspector Followup Item 325/88-05-04 and 324/88-05-04,

Hydrogen Leak in Turbine Building Pipe Tunnel. The inspectors "

reviewed Operating Experience Report No.88-024 and conducted an

inspection of the rad-waste pipe tunnel and the electrical tunnel

located above the pipe tunnel. The inspectors determined that the

pre-installation pressure test requirements had been modified to

preclude pressure testing of valves on their back seat, that

temporary hydrogen monitors will be installeo above the subject

hydrogen valves per SP-88-01 (not yet accomplished as of March 29),

and that routine system checks for hydrogen leaks were being

conducted.

The potential for hydrogen intrusion into safety-related equipment

I areas does not represent a hazard in the pipe tunnel as currently

configured. This item remains open pending the inspector's review of

other possible hydrogen interactions with safety-related equipment.

One potentially significant safety matter regarding operator inattention

during plant shutdown conditions was identified. Three violations, and no

deviations were identified.

9. Plant Modifications - Unit 2(37700)

The inspector reviewed PM-87-128, Weld Overlay Repair, Field Revisions 30

and 31, Repair of CRD Line Leaks, to verify compliance with ENP-03, Rev.

35, Plant Modification Procedure. The inspector observed the pipe replace-

ment and repair, which used cryogenic couplings to connect the new pipe.

The inspector also reviewed purchase ordars and specifications related to

the cryogenic fittings as well as ALARA ;we-job reviews conducted for the

job. Personnel interviews of selected eng'neering ALARA personnel were

also conducted.

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The licensee replaced the CRD lines after leaks were found on insert and

withdraw lines inside containment during the reactor coolint system

hydrostatic test. Small weeping leaks were found on 5 lines on March 19,

1988. Additional CRD line leaks were found on subsequent inspections;

all leaks were repaired. A-brown discoloration on the lines was found, by i

CP&L's laboratory, to contain chlorides and sulfur. Two regional

inspectors reviewed the licensee's metallurgical investigation at the lab

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(see report 325,324/88-17). The inspector will continue to follow the

licensee's resolution as cart of the LER closecut. .

No violations or deviations were identified.

10. Allegation No. RII-88-A-8011

This allegation was received by the Senior Resident Inspector on

February 24, 1988, and later, on February 29, was repeated by the alleger

to the Region II Allegation Coordinator in Atlanta, Georgie. On March 23,

an inspector conducted an announced inspection of the allegation which .

involved a Shift Operations Supervisor allegedly not being fit for duty.

The inspector reviewed several pertinent documents to include the Quality ,

, Check Report No. B-88-01-10/QCR No.14221, and met with the following

individuals.

C. Dietz - Plant General Manager

A. Bishop - Plant Operations Manager

S. Choplin - Director of Personnel Relations

E. Enzor - Director of Regulatory Compliance

W. Ruland - Senior USNRC Resident Inspector

There is no USNRC regulation governing Fitness for Duty. However, the

licensae's policy, Drug and Alcohol Abuse Statement of Practice, states...

"The use of alcoholic beverages by an employee on or away from Company 1

property that may adversely affect the employee's job performance, or that

may n eflect unfavorably upon public or governmental confidence in the

manner in which the Company carries out its responsibilities may result in

disciplinary action, including possible termination".

• ha inspector learned that a similiar allegation had been anonymously

telephoned to the licensee's Quality Check Program during the evening 7

hours of January 25. Within a few hours after receipt, the Plant General i

Manager instructed his operational staff to conduct an immediate investi-

gation which included interviews of operation supervisors who had close

contact with the Shift Operations Supervisor during the shift in question.

The licensee concluded the allegation was without merit.

.

Based upon interviews and a review of documentation the inspector was

( unable to substantiate the allegation and, given the absence of an NRC ,

l regulation governing Fitness for Duty, found the licensee to have been

responsive to this allegation.

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i No significant safety matters, violations, or deviations were identified.

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11. Unusual Event / Fire (93702)

The licensee declared an Unusual Event for two minutes on March 9, 1988,

at 2:40 p.m., after a fire could not be confirmed out within 10 minutes.

The drywell chiller power cables (non-safety) on the Radwaste Building

roof were reported on fire at 2:27 p.m. The cable insulation had chafed

on the roof edge. The fire was announced at 2:28 p.m., and the fire

brigade was dispatched. The cables were de-energized at 2:31 p.m., and

the fire reported out at 2:36 p.m., by an I&C technician. Since the shift

fire brigade commander did not report the fire out vithin 10 minutes, the

licensee declared an Unusual Event per Emergency Frxecure PEP-2.1,

Revision 21, Initial Actions, which requires the declaration of an Unusual

Event with a fire within the protected area lasting more than ten minutes.

The shif t fire coannander reported the fire out at 2:42 p.m. No safety-

related equipment was affected by the fire.

Based on interviews with plant personnel, record and log review, the

inspector concluded that the licensee complied with their emergency

procedures.

No significant safety matters, violations, or deviations were identified.

12. List of Abbreviations for Unit 1 and 2

ALARA As Low As Reasonably Achievable

A0 Auxiliary Operator

APRM Average Power Range Monitor

BWROGS Boiling Water Reactor Owners Group

BSEP Brunswick Steam Electric Plant

CAC Containment Atmospheric Control

CAD Containment Atmospheric Dilution

CCW Closed Cooling Water

CRD Control Rod Drive

CST Condensate Storage Tank

DC Direct Current

DG Diesel Generator

DR Direct Replacement

EER Engineering Evaluation Report

ENP Engineering Procedure

EPA Electrical Protection Assembly

ESF Engineered Safety Feature

F Degrees Fahrenheit

FSAR Final Safety Analysis Report

GPM Gallons Per Minute

HP Health Physics

HPCI High Pressure Coolant Injection

HX Heat Exchanger

I&C Instrumentation and Control

IE NRC Office of Inspection and Enforcement

IFI Inspector Followup Item

JC0 Justification for Continued Operation

LER Licensee Event Report

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LLRT Local Leak Rate Test {

LOCA Loss of Coolant Accident  ;

MG Motor Generator

MST Maintenance Surveillance Test

NRC Nuclear Regulatory Comission

NSW Nuclear Service Water

OER Operating Experience Report

OP Operating Procedure

PA Protected Area

PI Pressure Indicator l

PM Plant Modification  ;

PNSC Plant Nuclear Safety Committee i

PSIG Pounds per Square Inch Gauge

PT Procedure Test

QA Quality Assurance  !

QC Quality Control  !

RB Reactor Building

RCIC Reactor Core Isolation Cooling

RHR Residual Heat Removal '

RPM Revolutions Per Minute

l RPS Reactor Protection System

RWCU Reactor Water Cleanup

i SP Special Procedure

l SRO Senior Reactor Operator

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STA Shif t Technical Advisor

SW Service Water

TC/ Thermostatic Control Valve

TS Technical Specification

URI Unresolved Item

USNRC United States Nuclear Regulatory Comission

V Volt

VID Violation

WR/JO Work Request / Job Order

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