Text

IE Insp Rept 50-298/81-14 on 810831-0904 & 0914-18. Noncompliance Noted:Posting Requirements Not Observed for Radiation Area & Overlay Weld Deposit on Mark I Torus Shell Did Not Meet Tech Specs
	ML20032C602
Person / Time
Site:	Cooper
Issue date:	10/21/1981
From:	Gilbert L, Randy Hall, Jaudon J, Johnson E; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20032C588	List: ML20032C587; ML20032C594; ML20032C602;
References
	50-298-81-14, NUDOCS 8111100635
	Download: ML20032C602 (20)
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U.S. NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

REGION IV

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- Report: 50-298/81-14

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Docketi 50-298

License:

DPR-46

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Licensee:

Nebraska'Public Power District

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Post Office Box 499

Columbus, Nebraska 68601

Facility Name:

Cooper Nuclear Station

Inspection at:

Cooper Nuclear Station Site, Nemaha County, Nebraska, and

Nebraska Public Power District Offices, Columbus, Nebraska

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Inspection conducted:

August 31 - September 4 and

September 14-18, 1981

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Inspectors:

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_ y. P./ Jandgh, Reactor Inspector, Systems and Techni.cai

Date

' Section (Paragraphs 1, 2, 6, 8, 9, 10, 12, 13,-and 14)

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E. H. Johnson, Reactor Inspector, Systems and Technical

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Section (Paragraphs 1, 3, 5, 7, 11, 13, and 14)

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L. D. Gilbert, Reactor Inspector, Engineering and

Date

Materials Section (Paragraphs 1, 4,13, and 14)

Approvea:

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/6 / 7

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R. E. Hall, Chief, Systems and Technical Section

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8111100635 811023

PDR ADOCK 05000298

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PDR

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Inspection Summary

Inspection on August 31 - September 4 and September 14-18, 1981

(Report 50-298/81-14)

Areas Inspected:

Routine, unannounced inspection of licensee action on previ-

ously identified items, radiation areas, welding, calibration, requalification

training, fire protection, special tests and experiments, QA program, audits,

SRAB audits, and organization and administration.

This inspection involved

173 inspector-hours by three NRC inspectors.

Results:

Within the eleven areas inspected, no viciauons were identified in

eight of the areas, but one violation was identified in each of the remaining

three areas (violation - failure to post radiation area; violation - reduction

of requalification progr-

requirements without prior Commission approval; and

violation - failure of Safety Review and Audit Board to review audit status).

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DETAILS

1.

Persons Contacted

Nebraska Public Power District (NPPD)

P. Borer, Engineering Supervisor

B. Brungardt, Surveillance Coordinator

T. Coates, Lead Instrument Technician

R. Creason, Acting Operations Supervisor

W. Fitzsimmons, Mechanical Foreman

W. Gilbert, Training Coordinator

C. Goebel, Administrative Supervisor

R. Janksy, Shift Supervisor (Operations Training)

H. Jantzen, I&C Supervisor

L. Lawrence, Maintenance Supervisor

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L. Lessor, Station Superintendent

D. Majores, Maintenance Planner

R. Mcdonald, Health Physicist

D. Overbeck, Administrative Assistant

J. Pilant, Director Licensing and Quality Assurance

J. Sayer, Chemistry and Health Physics Supervisor

G. Smith, QA Specialist

+P. Thomason, Assistant to the Station Superintendent

D. Whitman, Technical Assistant to Station Superintendent

F. Williams, Manager Quality Assurance

Chicago Bridge and Iron (CB&I)

R. Gorder, Project Manager

The NRC inspectors also contacted other plant personnel including

administrative, engineering, operations, and maintenance personnel.

+ Denotes presence at the exit interview conducted September 4,1981.

Denotes presence at the exit interview conducted September 17, 1981.

Denotes presence at the exit interview conducted September 18, 1981.

2.

Licencee Action on Previous Inspection Findings

(Closed) Open Item (8012-01):

This item was open because the licensee

had aade a change in off-site organization but had not forwarded a

Techlical Specification change request.

The NRC inspector found that

the 'icensee had included this change in a Technical Specification change

request, dated August 26, 1981.

This item is closed.

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(Closed)UnresolvedItem(8105-01): This item was unresolved because

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the licensee had been conducting training in industrial safety on a

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monthly basis but had not been incorporating these records into plant

t) aining records as required by Technical Specifications. The NRC

inspector found that :the licensee was now retaining records of indus-

trial safety training in permanent training records.

This itam is closed.

(Closed) Unresolved Item (8105-02): This item was unresolved pending

licensee action to assure permanent retention of . records of training

conducted in procedures. The NRC inspector found that records of-

procedure training were now being incorporated into training records.

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This item is closed.

(Closed) Unresolved Item (8105-03): This item was unresolved pending

licensee action to incorporate on-the-job training records maintained

by craft supervisors into permanent training records as required by

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Technical Specifications. The NRC inspector found that on-the-job

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training records were now being filed with permanent training records.

This item is closed.

3.

Posting Requirements Not Observed for a Radiation Area

During this inspection, the NRC inspector noted that the storage / work

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room adjacent to the NRC reactor resident inspector's office (which is

located inside the licensee's protected area) contained twenty cardboard

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boxes carrying the standard radiation symbol and marked as containing

radioactive material. A yellow and magenta rope was affixed to one wall

of this room and was tied off to a folding chair that was leaning against

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these boxes at the middle of the side opposite the wall. This rope

provided a barrier around approximately two third of the perimeter of

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the boxes, although it was draped across one of tne boxes at the corner.

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A standard radiation symbol sign attached to the rope announced

" Radioactive Mate.ial - to be manded - coveralls - 3 mr/hr at 18 inches."

A work table and sewing machine were located just outside this partial

barrier.

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The NRC inspector measured the radiation levels near the boxes with

several low range gamma detectors and determined that at the barrier

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does rates varied from 5 mr/hr to 8 mr/hr. These readings were con-

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firmed by health physics technicians from the licensee's staff. At

the work table, the dose rate was determined to be between 2.5 and

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2.8 mr/b-

-A white polyethylene bag containing what appeared to be discarded cover-

alls, zippers, paper, and other miscellaneous trash was found in the

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storage room about fifteen feet from the boxes and outside the partially

roped off area.

Radiat.on levels on contact at the centerline of the

bag were 6-7 mr/hr and at the surface directly on top of the open bag

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were 4-5 mr/hr.

The NRC inspector pointed this out to the licensee's

health physicist, who indicated that station policy was to treat all

such white plastic bags as containing ratentially radioactive material

until otherwise verified and that a fiaal check was made of each truck

leaving the site with bags of trash.

With regard to the t ,as of radioactive material in the storage room,

10 CFR 20, Section 1 J.202, paragraph (b)(2) defines a radiation area

as "any area, accessible to personnel, in which there e>ists radiation,

originating in whole or part within licensed material, et such levels

that a major portion of the body could receive in any one hour a dose in

excess of 5 millirem, or in any five consecutive days, a dose in excess of

100 millirems." Section 20.203(b) of this regulation requires that such

areas be conspicuously posted as radiation areas.

These requirements are

reflected in the licensee's Procedure 9.1.2.2, Revision 4, " Area Posting

and Access Control." Section 6.A.3.g of this procedure contains a defini-

tion of radiation area identified to that ii section 20.202 of 10 CFR 20,

and section 6.B of the foregoing procedure requires that radiation areas

be posted as " Caution Radiation Area" with clearly identified boundaries.

The failure to post the storage area containing the twenty boxes of con-

taminated coveralls as a radiation area is an apparent violation.

(8114-03)

This finding was brought to the attention of the station health physicist

who indicated that immediate corrective action would be taken to properly

dispose of this material.

The NRC inspector noted, during the second part

of the inspection, that all the .adioactive material had been removed from

the storage / work area.

4.

Welding Mark I Torus Modifications

The NRC inspector reviewed NPPD Contract 80-28 with Amendment No. 1 for

modification of the Mark I Torus and observed the welds of the installed

Torus saddles.

The NRC inspector determined that the overlay weld deposit

on the Torus shell does not comply with paragraph 6.4.2 of Section G-Part IV

to the Technical Specification for Installation of Torus Saddles in

Contract 80-28 between the NPPD and Chicago Bridge and Iron (CB&I), in that

the overlay does not extend 1/2 inch minimum beyond the toe of the saddle

attachment welds.

The NRC inspector observed areas on each of the four

saddles inspected, sadoles 8 throt.gh 11, where the overlay extends less than

1/4 inch beyond the weld toe.

The Technical Specification for installation states that a qualification

test may be performed to change from the welding techniques specified in

paragraph 6.4 of Sect b n G-Part IV.

There was no CB&I representative on

site during the inspection; therefore, it could not be determined whether

there had been a test and approval to change the welding technique to

permit this reduced overlay. This item is considered unresolved pending

review of the CB&I welding procedure qualification test report to determine

if the change in welding technique was autho hed.

(8114-04)

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5.

Calibration

The objective of this inspection effort was to examine the licensee's

program for the calibration of components specified in or closely

related to the Technical Specifications to ascertain that satisfactory

calibrations at the required frequencies are being accomplished.

To determine if the frequency for calibrations and functional checks

called out in the Technical Specifications were being met, the NRC

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inspector reviewed completed 1981 surveillance tests for approximately

fifty percent of the instrumentation'in the reactor protection system,

emergency core cooling initiation, and primary containment isolation

systems.

These instruments and surveillance requirements are spelled

out in Technical Specification Tables 4.1.1, 4.1.2, 4.2.A, and 4.2.B.

In addition, the NRC inspector selected several surveillance require-

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ments for safety-related auxiliary systems and reviewed records for

completed surveillance tests for these components.

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The following Technical Specification functional check and calibration

requirements were reviewed:

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Component / Protective Feature

Surveillance Procedure

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Manual Scram (functional)

6.1.13

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IRM Flux High (functional)

6.1.2 and 6.1.2A

APRM Flux High (15% level trip)

6.1.22

(functional)

APRM Inoperative (functional)

6.1.22, 6.1.3, 6.1.29

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APRM Flow Bias

6.1.29, 6.1.18

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IRM Flux High (calibration)

10.2

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Drywell Pressure High (functional)

6.1.10

Reactor Pressure High (calibration)

6.1.5

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Reactor Water Level Low Scram

6.1.9

(calibration)

Scram Discharge Instrument Volume

6.1.14

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High Level

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MSIV Closure

6.1.6, 6.2.2.7.1

Turbine 1st Stage Pressure Permissive

6.1.11

Reactor Pressure Permissive

6.1.12

(functional)

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. Turbine Stop Valve Closure

6.1.32

-(calibration)

Reactor Water Level Low --Isolation

6.1.9

Main Steam Line Leak Detection

6.2.1.4.1, 6.2.1.6.2-

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Main Steam Line Pressure Low

6.2.1.4.3

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Condensor Vacuum Low

6.2.1.3.1

Drywell Pressure High - ECCS

6.2.2.1.2, 6.2.2.2.2

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Reactor Water Level Low - ECCS

6.2.2.1.1

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Reactor Pressure Low

6.2.2.1.3

Equipment Drain Sump Flow Meter

6.2.3.2

(calibration)

Main Control Room Ventilation

6.4.6.3

(actuation)

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Suppression Chamber Water Level

6.2.2.8.6

(calibration)

In the above review, completed test results were checked for conformance

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to Technical Specification requirements, selected procedures were reviewed

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for technical content to ensure that a satisfactory calibration would

result, and it was verified that the procedure actually used for testing

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.was the currently approved revision in force at the time of the test.

The

qualifications of several technicians performing these tests were reviewed

to verify they possessed' appropriate training and experience.

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No violations or deviations were identified during the foregoing review.

One procedural weakness was detected in Surveillance Procedure 6.2.2.7.1,

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" Main Steam Isolation Valve Limit Switch Calibration and Functional Test,"

Revision 5.

Technical Specifiretion Table 4.1.2 requires that these switches

undergo physical inspection each refueling outage; however, the supporting

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surveillance procedure does not apparently define any acceptance criteria

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or give guidance as to what constitutes a satisfactory inspection.

This

item was brought to the attention of the surveillance coordinator for retiew.

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The NRC inspector selected the following additional Technical Specification

surveillance requirements, which do not specifically require calibrations,

to determine if the associated safety-related components are periodically

calibrated:

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Technical Specification

Procedure or Method of Calibration

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SBLC Pump Discharge Pressure

Discharge pressure gage calibrated

4.4.A.2.b-

under Procedure 7.5.9.1

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Core Spray Pump Discharge Flowrate

S.P.6.2.2.4.1

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4.5.A.1.d

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Diesel Generator Jacket' Water Heater

Reading logged on station operators

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3.9.A.2

tours each four hours

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ECCS Discharge Pipe Fill System

Reading logged on station operators

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4.5.G

tours each four hours, low pressure

alarm switches calibrated under

Procedure 6.2.2.1.b

Reactor Vessel Bottom Head Drain

Recorder calibrated under

Temperature 4.6.A.l.a

Procedure 7.1.1

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Continuous Conductivity Meter

Procedures 8.5.1.4 and 8.5.3.4

4.6.B.2

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Meteorological Tower Wind Speed

Procedure 7.5.3.5

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6.6.1

Secondary Containment DP 4.7.C.1.a

Procedure 6.4.8.6

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Diesel Generator Fuel Oil Tank Level

Procedure 6.3.12.4

3.9.A.2

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Jet Pump Operability (DP reading)

Procedure 7.5.6.10

4.6.E.1

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The NRC inspector detected that calibration of the above components,

for which surveillance procedures are specified, are scheduled in the

master surveillance schedule.

The remaining components are entered

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into the station preventive maintenance program, and a monthly listing

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of items due for preventive maintenance is distributed to craft super-

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visors.

Completed preventive maintenance items are logged into the

system so that supervisors can be appraised of overdue items.

At the

time of the inspection, there were only a few items overdue, and these

were already scheduled for performance.

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The NRC incpector also revi ~ed the procedures associated with the cali-

brations or calibration chec/s related to the foregoing technical speci-

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fications to determine that appropriate acceptance criteria are included,

and that the procedure will result in a suitably accurate calibration.

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There were no deficiencies detected in this area.

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The licensee's program for the control, storage, and calibration of test

equipment used in the performance of Technical Specification calibrations

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was reviewed.

All station test equipment-is entered in a master list

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(Procedure 7.1.2, Revision 5, " Calibration Procedure for Test Equipment").

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y4,,-mv*

,,ym,--,.--y-,,-,e,m,,.,,m,,m-,c,,,,,,,,,w,e,-,,-,m-,w-,

,,,-.,e,im-,,,,wv,twm-,-*,.v--n+,,w-%.,~.,,4

gw.gre---,-, wee-

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Calibration frequencies are included in this procedure.

The NRC inspector

civiewed the records for three pieces of test equipment (one thermometer,

one Wallace-Tiernan pneumatic calibrator and one 0-30 psig Ashcraft pres-

sure gage) and determined that the calibration records for these instruments

were traceable to the National Bureau of Standards, and that storage and

care of the instruments were suitable.

No deficiencies were identified in

this area of the inspection.

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The NRC inspector observed the quarterly calibration of the service water

radiation monitor under Procedures 6.3.7.4, Revision 3 and 7.5.6.8,

Revision 4._

The calibration appeared to be performed in accordance with

the procedure, and no equipment or procedural inadequacies were identified.

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

6.

Requalification Training

The NRC inspector reviewed the licensee's program for requalification

training.

The licensee's requalification training program had been

approved for the NRC by P. F. Collins letter of August 18, 1976.

The NRC

inspector found that the licensee had prepared a revised requalification

training program, dated September 1980.

The NRC inspector found that the

licensee had implemented portions of this revised requalification training

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plan without Commission approval.

10 CFR 50.54, paragraph i-1 states, in

part, "Notwithstanding the provisions of (10 CFR) 50.59 the licenste shall

not, except as specifically authorized by the Commission, make a change in

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an approved operator requalification program by which the scope, time

allotted for the program or frequency in conducting different parts of the

program is decreased." Contrary to this, the NRC inspector found that, for

written examinations given covering material presented in the requalification

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program lecture series, the minimum acceptable grade had been lowered from

80% to 70%.

The NRC inspector noted from the licensee's records two examples

in which individuals who had been examined on the content of requalificatior,

lectures had made grades between 70% and 80%, but had not been rescheduled

for lectures as required by the approved requalification training plan.

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This is an apparent violation. (8114-02)

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The NRC inspector founa a second example of this apparent violation.

Contrary to 10 CFR 50.54, the last licensee annual examination of licensed

operators and senior operators did not examine tt.em on applicable portions

of 10 CFR, Chapter 1.

This apparently had resulted from the fact that the

licensee changed some of the titles of the program from those listed in the

approved training plan.

Specifically, the approved requalification training

plan required " Technical Specifications" and " applicable portions u.' la CFR,

Chapter 1" to be covered.

The licensee had changed these two items to

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" Administrative procedures, conditions, and limitations" in the 1980 revision

of the requalification training plan.

The NRC inspector found that, f'r the

annual requalification examination given in March 1981, there had been no

questions related to 10 CFR, Chapter 1.

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The NRC inspector reviewed the annual requalification examination and the

results thereof.

Other than the items noted above, the annual examination

was satisfactory in the NRC inspector's judgement.

The NRC inspector

noted that the licensee's evaluation of this examination had been very

detailed.

The licensee had not only examined the operators responses and

grades in each section of the examination, but had also analysed the

responses to each question.

Licensee representatives stated that this

had been done in order to pinpoint specific areas of weakness that might

exist in an examination area, even though all operators scored above 80%

in the area.

The result of this analysis was a comprehensive and detailed

training schedule for 1981-82.

The NRC inspector found that at the time

of the inspection the licensee had already fallen behind this schedule.

The NRC inspector noted that two individuals, who were required to attend

specific requalification training as a result of the annual examination,

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had not attended this training despite the fact that it had been given

four times.

The NRC inspector noted that missed training lectures provided

an additional workload for the training staff. The licensee training staff

numbered two.

The ratio of on-site personnel to each trainer was approxi-

mately 75:1, a very high ratio.

The NRC inspector noted that the trainers

were also involved in training personnel for initial licensing but did

receive some assistance in this and other training from other site personnel

(e.g., health physics techniciaris taught HP indoctrination).

The small size

of the training staff and the fact that the licensee was already behind

schedule for requahfication training created doubt in the NRC inspector's

mind as to whether the licensee would meet his minimum requalification

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training requirements.

This is considered an unresolved item pending

subsequent review later in this training cycle.

(8114-05)

The NRC inspectc . noted that the licensee's method for recording infor-

mation on reactivity manipulations was for the personnel on watch to

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record pertinent information on forms in the control room.

A member of

the training staff then later transcribed these records to permanent

records.

A similar system was used to record emergency or abnormal actions.

The NRC inspector found that the last recorded events in these categories

at the time of the inspection were September 3, 1981, and June 19, 1981,

respectively.

The NRC inspector established that neither of these dates

were the latest for the respective categories.

This is considered to be

an unresolved item pending further review of the licensee's method for

recording reactivity manipulations, which may change when the list of reac-

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tivity manipulations to be recorded changes.

(8114-06)

The NRC inspector was shown licensee evaluations of licensed operators and

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senior operators.

These evaluations emphasized the evaluation of operators

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during abnormal and emergency conditions.

These evaluations appeared to

meet the requirements of 10 CFR 55, Appendix A, paragraph 4.C and the

licensee's approved training plan.

When the NRC inspector probed this

area, he determined that the licensee had a second system of operator

evaluation and feedback to training which was run by the Operations

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Supervisor.

Records were reviewed which showed training in the form of

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additional procedure review, lectures, or required reading.

This training

was the apparent result of needs perceived as the result of operator

evaluation, although these evaluations were not recorded as such.

The NRC inspector expressed concern to licensee management that, while the

dual system of evaluations technically appeared to meet the evaluation

requirements of 10 CFR 55, Appendix A, it appeared to miss the intent of

this requirement in that the real evaluation of operators was not recorded

as such.

The NRC inspector also determined that the licensee was conducting,

or had scheduled, training in heat transfer, fluid flow, thermodynamics, and

mitigation of accidents involving a degraded core.

The NRC inspector also

determined that all these categories except the last.one had been incl::ded

in the revised training program of September 1980.

A separate training pro-

gram for this last category had been provided for the licensee's operating

staff.

7.

Fire Protection / Prevention Program Elements and Implementation

The objective of this portion of the inspection was to ascertain whether

the licensee's program for fire protection and prevention was in con-

formMce with 10 CFR 50, Appendix R requirements; Technical Specification

requirements; and approved industry codes and standards as committed to

in the licensee's Fire Hazard Analysis, and to verify that the program was

being implemented.

The following elements of fire protection and prevention were included in

this area of the ir.spection:

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administration of the program

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administrative controls for combustible materials, welding and cuttir;g

operations, housekeeping and maintenance, and modification activities

relative to their impact on fire protection

fire brigade organization and training

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fire protection and prevention training for persons not assigned to

fire brigade

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maintenance, test, and inspection procedures for fire protection

equipment

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program for internal audits of the fire protection and prevention

program

At the corporate level, the licensee provides support for the station fire

protection and prevention program through a fire protection engineer.

This

individual performs annual inspections of the fire protection program for

the corporate level Safety Review and Audit Board as required by Technical Specification 6.2.8.a.

The NRC inspector reviewed the results of the latest

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fire protection inspection performed for SRAB (inspection dates July 20-23,

1981).

The scope of this inspection appeared to cover the areas required

by the Technical Specifications and identified several apparent procedural

weaknesses for which, the inspector was informed, procedure changes were

being drafted; an item dealing with control of cutting operations is dis-

cussed in the following paragraphs.

Responsibility for fire protection at the station is assigned to the

Station Fire Chief.

At the present time, this duty is assigned to the

Engineering Supervisor.

The Station Fire Chief supervises the training

of the fire brigade and backup support force (mechanical maintenance

personnel).

Training in the past has consisted of quarterly fire drills

for each fire brigade and an annual " hands-on" session for all brigade

and backup personnel which has included a classroom briefing on fire

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fighting techniques and principles and practice against an actual con-

trolled fire.

This training program is outlined in a document entitled

" Cooper Nuclear Station Fire Protection Training Program," dated July 10,

1980.

During the inspection, the NRC inspector was informed that this

program was being revised to include requalification requirements for fire

brigade and support personnel.

The new program will continue quarterly

fire drills for each fire brigade with an annual lecture and " hands-on"

session for each operator.

Security force. personnel (who provide two

members of the fire brigade) and mechanical maintenance personnel will take

part in the lecture and " hands-on" sessions once each two years with the

st'pulation that new employees in each of these categories attend the first

available session after. reporting to the site.

The proposed revised program

appeared to meet the requirements of 10 CFR 50, Appendix R.

The NRC inspector determined that lesson plans were not yet available for

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the proposed fire protection training program.

He was informed that these

would be prepared as the program was performed.

As the revised program

has not yet been approved and implemented, the inspector indicated that

the final, approved version would be reviewed at a future date to assure

continued conformance to Appendix R of 10 CFR 50.

(0 pen Item 8114-07)

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Fire brigade training records for 1980 were reviewed and revealed that

fire protection training had been conducted as required.

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Fire protection and prevention training for the balance of the plant

staff is included in the general employee training program.

Plant

Procedure 5.4.1, " General Fire Procedure," Revision 12 indicates that any

person discovering a fire should notify the control room of the type (i.e. ,

Class A, B, C, or D), magnitude, and location of a fire in order to aid the

fire brigade in de?.ermining the best initial response to a fire.

It was

not readily apparer,t that employees, other than those who are trained as

part of the fire brigade or backup support group, received periodic instruc-

tion as to the types of fire hazards that might be found at the plant and

consequent classes of fire that could be experienced.

The licensee conducts

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periodic safety meetings of the plant staff and the NRC inspector was

informed that fire hazards are discussed as part of the overall safety

program.

The NRC inspector questioned several employees and determined

that they understood their responsibilities under Procedure 5.4.1.

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inspector had no further questions on this item.

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The NRC inspector reviewed licensee administrative procedures to deter-

mine that controls for combustible materials and operations that might

-

raise fire hazards were specified.

Controls of such materials and

a;tivities were spelled out with the exception of c;tting operations.

Such operations using oxyacetylene torches. raise obvious fire hazards.

The inspector noted that this item had been identified during the fire

protection inspection conducted for the SRAB (discussed in the second

2

paragraph of this section).

The corrective action suggested to provide

control over cutting operations was a permit system similar to that in

use for weldir.g activities and the inspector was informed that a proce-

dure change was being drafted for this.

The permit system to control

'

cutting operations will be reviewed during a future inspection.

(0 pen

item 8114-08)

i

Preventive maintenance, test, and inspection of fire protection equipment

is specified in the Technical Specifications for all fire protection

i

systems and equipment associated with safety-related components and

areas (Technical Specification:: 4.14 through 4.20).

These requirements

are contained in the licensee's surveillance Procedures 6.4.5.1 through

6.4.5.17.

The following Technical Specification requirements for fire

7

protection equipment were reviewed to determine that technically adequate

and approved procedures existed.

Techncial Specification

Procedure

4.14.A

6.4.5.5, Rev. 7, " Fire Detection System

Semiannual Inspection"

4.14.8

6.4.5.6, Rev. 4, " Fire Detection System

Circuitry Operability"

4.15.A.1

6.4.5.1, Rev. 26, " Fire Detection System '

Monthly Inspection"

!

4.15.A.2

6.4.5.1, Rev. 26, " Fire Detection System

Monthly Inspection"

4.15.A.3

6.4.5.2, Rev. 17, " Fire Detection System

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Annual Inspection"

4.15.A.6.a

6.4.5.1, Rev.17, " Fire Detection System

Annual Inspection"

4.15.A.6.c

6.4.5.9, Rev. 3, " Diesel Fire Pump

Inspection

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_ - _ _ _ _ _ _ _ _ _ __

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_ - _ _ _ _ _ _ _ _ _ _ _ _

.

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4.15.A.7.b & c

6.4.5.11, Rev. O, " Diesel Fire Pump

Battery Inspection"

4.16.A.2

6.4.5.16, Rev. 0, " Fire Protection

System Eighteen Month Inspection"

4.19.A.1 & 2

6.4.5.16, Rev. 0, " Fire Protection

System Eighteen Month Inspection"

The results of completed surveillance tests were reviewed to determine

if testing had been completed at the required interval and that system

performance was satisfactory.

Procedure

Dates

6.4.5.1 (monthly)

8/7/81, 7/10/81, 6/15/81

6.4.5.2 (annual)

7/21/81, 3/30/81, 2/20/81,

2/15/81, 2/12/81 (various sections)

6.4.5.3 Fire Protection

8/28/81, 8/21/81, 8/14/81, 8/7/81,

Pump Weekly Operability

7/28/81

Test

6.4.5.4 Fire Pumps Six Month 1/19/81, test for second half of

Flow Test

1981 performed on 7/21/81 under

6.4.5.2

6.4.5.5 (semiannual)

8/18/81, 2/13/81

6.4.5.6 (semiannual)

5/15/81

6.4.5.7 Diesel Generator CO2 2/11/81

Operability (semiannual)

The NRC inspector toured accessible areas of the plant to verify that

fire protection equipment was installed and appeared to be operable

as required and that proper housekeeping was being observed.

It ws3

noted that several areas of the station are posted as no smoking

areas, specifically the Reactor Building and station battery rooms in

the Control Building basement.

The NRC inspector did note, however,

that the cable spreading room was not posted as a nonsmoking area and

several cigarette butts were found during the tour of this safety-related

area.

The inspector could not identify any requirement prohibiting

smoking in this area.

He expressed his concern that smoking was not con-

trolled in the cable spreading room to the station superintendent during

the exit interview.

The station superintendent said that this area was

routinely used as passage to and from the Control Building basement, and

that it would not therefore be posted as a nonsmoking area; however,

smoking would be discouraged for workers in the cable spreading room.

No violations or deviations were identified in this part of the inspection.

14

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_ - _ _ - _ _ - _ _ _ _ _ . -

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8.

Tests and Experiments

The NRC inspector reviewed the licensee program for the control of tests

and experiments. The NRC inspector reviewed eighteen Special Procedures

(SP) and Special Test Procedures (STP) conducted during the period

1979-1981.

The NRC inspector concluded that the majority of these were

not " tests or experiments" within the context of 10 CFR 50.59.

The NRC

inspector noted that each SP and STP was reviewed by the Station Operations

Review Committee (SORC).

Additionally, it was found that a "10 CFR 50.59

Reportability Analysis" had been prepared for each SP and STP.

This

reportability analysis included an unreviewed safety question analysis

and the basis for this determination.

The NRC inspector questioned licensee

representatives as to whether the reportability analysis form was slanted

toward design changes vice test or experiment, but concluded from this dis-

cussion and the records reviewed that the licensee's use of the reportability

analysis was adequate.

The NRC inspector reviewed licensee Procedure 1.12, Revision 5 (8/24/81),

" Procedure and Reports-Use and Format." This procedure provided instructions

for the preparation, review and approval of SPs and STPs.

The NRC inspector

noted that this procedure did not contain provisions to implement the require-

ments of Technical Specifications, section 6.2.1.A.4.b for SCRC review of

certain test results.

The NRC inspector noted from other records and inter-

views that the 50RC dic' conduct reviews of some completed SPs and STPs.

There were no examples identified of an SP or STP wherein the results of a

test required 50RC review and such review was not accomplished.

This is

considered an unresolved item pending licensee action to promulgate or to

revise procedures so that the Technical Specification review requirement

is implemented.

(8114-09)

9.

Quality Assurance Program

The NRC inspector reviewed the licensee's quality assurance program.

The

NRC inspector noted that there had been no major changes either implemented

or requested in the Quality Assurance program since the last inspection of

this area.

The NRC inspector reviewed the following licensee procedures:

QAI-5, Rev. 13 (1/12/81), " General Guidelines - Quality Assurance

Audits"

QAI-6, Rev. 8 (5/29/81), " Personnel Qualifications and Training for

QA Assignments"

1.7, Rev. 13 (12/18/80), " Station Maintenance"

Records of sixteen completed maintenance items were reviewed to verify that

the licensee was implementing quality control checks in accordance with his

procedues.

There were no violations or deviations identified in this portion of

the inspection.

15

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10. Audits

The NRC inspectors reviewed the licensee's audit program as executed by the

Quality Assurance (QA) department.

Four QA audits were reviewed.

One of

tnese audits was " Training" (No. 81-8).

This audit was conducted in February

and March 1981.

Facts identified in this audit matched those found by NRC

inspectors; but the conclusions reached did not apparently lead to discovery

of potential violations existent.

For example, Audit 81-8 identified that

the I&C Supervisor maintained on-the-job training records, but missed the

fact that many of these records had never been incorporated in permanent

training records as required by Technical Specifications.

The NRC inspector

found a seco,d example.

Audit 81-8 described the requalification training

program and M.ated that a revised requalification training had been sub-

mitted fo approval.

The NRC inspector noted that the description-of the

requalification program in effect usea verbatim language from the revised

(i.e., unapproved program) to describe the subjects covered in the annual

!

examination.

This change.in requalification programs resulted in an

apparent violation described in paragraph 6 of this report.

The NRC

inspector expressed contem to licensee management that QA audits were

identifying facts but missing issues of regulatory concern.

No violations or deviations were identified during this portion of the

,

inspection.

'

11.

SRAB Audit Activities

The purpose of this inspection effort was to review the completed audits

of the off-site Safety Review and Audit Board (SRAB) which are required

by Technical Specification 6.2.1.B.

,

The aforementioned Technical Specification requires the SRAB to audit

,

f

selected aspects of plant operation such that all nuclear safety-related

activities are covered in at least a two year period.

This specification

further requires that periodic reviews of the audit programs should be

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performed by the Board at least twice a year to ensure that such audits

!

are being accomplished in accordance with the regirements of Technical

l

Specifications.

For ease in executing the audit program, the SRAB has defined fourteen

audit areas to cover the nuclear safety-related activities at the station

7

l

and a two year cycle during which each area is to be audited at least

l

once.

Refueling activities are audited at each ann'!al refueling.

The

j

current two year cycle began on January 1, 1980, to run through December 31,

1981.

An audit schedule with assigned audiars was drawn up by the SRAB

i

in April 1981 (SRAB Meeting 65).

I

The inspector reviewed the status of the audit progru and determind tn'

following: of fourteen audits scheduled during the two year cycle 1980-bM,

only four audits had been completed.

The table below indicates the curru.

status of SRAB audits:

16

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1980-1981 Cycle

Audit Areas

Date of Last Audit

Scheduled For

Notes

Station Security

March 1980

(1)

Conformance With

March 1978

5/80

(2)

Regulations and

Operating License

Conditions (Including

Surveillance Testing)

Station Operations

January 1978

5/80, 10/81

(2),(3)

Major Maintenance and

August 1979

12/80

(1)

ISI

Chemistry, Health

August 1981

Physics and

Environmental

Monitoring

Radioactive Waste

January 1979

12/80

(1)

Treatment and

Disposal

'

Design Changes

August 1978

8/80

(1)

Nuclear Materials

January 1980

9/81

(1)

Safeguards

QA Program

December 1980

Refueling

May 1981

Industrial Safety

September 1981

8/80

Emergency Plans

November 1978

(1)

!

Fire Protection

July 1981

Training

January 1980

2/81

(1),(4)

Notes:

r

(1) Audit performed to satisfy 1978-1979 cycle requirements.

(2) Audit performed to satisfy 1976-1977 cycle requirements.

(3) An audit was scheduled for May 1980 and apparently done, but a

<

i

report was never prepared due to lost audit notes.

j

(4) A memo of August 25,-1981, rescheduled this audit to November 1981.

I

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.

The above status indicates clearly that the licensee has failed to adhere

to his schedule for completion of audits, and for several safety-related

audits (items 2, 3, and 7 in the foregoing table), the interval since the

last SRAB audit has been three years or more.

The NRC inspector then reviewed records of SRAB meeting minutes to deter-

mine if the licensee had identified this problem through periodic review

of the audit program as required by the Technical Specifications.

Minutes

for SRAB meetings 63 through 70 inclusive (December 4,1979, through

August 3,1981) indicated that the SRAB had discussed SRAB audits on only

two occasions.

The first occasion was during SRAB meeting 65 (April 18,

1980) where the status of audits fc' the 1978-1979 cycle was discussed;

and the second was during SRAB meeting 70 where the results of a QA

department surveillance of SRAB activities were discussed.

The NRC inspector reviewed the QA department surveillance report referred

to above.

The report was dated April 28, 1981, for an audit conducted on

April 22-24, 1981.

The report identified two major discrepancies in the

SRAD audit program.

First it noted that SRAB audits were not being con-

ducted as scheduled; this finding was nearly identical to that of the NRC

inspector reported above. The second finding was that audit reports were

not being issued in a timely manner with an interval of up to one year from

the time of the audit. These findings were discussed with the Director of

Licensing and Quality Assurance during the exit interview for this portion

of the inspection.

He indicated to the NRC inspector that this report had

received considerable attention from the SRAB during the most recent SRAB

meeting (No. 70 of August 3, 1981).

During this meeting, the SRAB Chairman

stressed to board members their responsibilities for completing the assigned

audits as scheduled; this was followed up on by memorandum to the assigned

auditors.

Additionally, the licensee has issued revised operating proce-

dures for the SRAB.

These procedures went into effect on July 31, 1981,

and specify that audit teams will be required to perform their assigned

audits within the time limits set out in the audit schedule.

Further,

the procedure requires that an auait report be prepared as soon as

practicable after the audit.

The NRC inspector took note of these corrective actions; but pointed out

that an apparent violation of the Technical Specification requirement

that SRAB periodically review the status of the audit program had

occurred.

The failure to meet this requirement was, in the NRC inspector's

judgement, a contributing factor to the failure of SRAB to perform the

required audits as discussed in the foregoing paragraphs.

(8114-01)

The NRC inspector expressed his concern about the apparent failure of

SRAB to perform audits as required and lack of periodic SRAB review

of the audit program status.

The inspector noted that similar findings

had been the subject of previous NRC inspection reports.

In IE

Inspection Report 50-298/75-12 (paragraph 4), it can be observed that,

18

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". . . . The inspector also noted that audits of (sic) the Safety Review

and Audit Board (SRAB) were not being conducted during the time frame

scheduled by the licensee . . . ." .This item did not result in enforcement

action; however, during IE Inspection 50-298/77-07 (referfence IE Inspection

' Report 50-298/77-07, paragraph 2), the NRC inspector found that of twelve

audits set out by SRAB for accomplishment in the two year cycle, January 1976

to December 1977, only four had been accomplished as of the date of the

inspection (June 22-24,1977).

Further it was noted that the SRAB minutes

could not establish that the periodic review required by the Technical

Specifications had been conducted.

The report notes that this failure to

review constituted a violation of Technical Specification 6.2.1.B.

In view of the fact that the conduct of the SRAB audit program had previ-

ously resulted in enforcement action, the NRC inspector stated that careful

review and clearly formulated corrective action would be in order for this

item.

12.

Organization and Administration

The NRC inspector reviewed the licensee's on-site organization.

The

NRC inspector compared the licensee's Technical Specifications section 6.1, " Organization," to licensee Procedure 1.2, Revision ~7

(1/18/80), " Station Organization and Responsibility." There were no

conflicts identified between these two.

The NRC inspector did detect,

however, that Procedure 1.2 did not state to whom positions described

reported to.in each case.

For example, Procedure 1.2 did not state to

'

whom the Shift Technical Advisor (STA) reported.

The NRC inspector

found that the licensee had trained both engineers and shift super-

visors as STAS.

The NRC inspector noted that the licensee had advised

the NRC (NPPD letter LQA 8000439 of October 31,1981) that, ". . . at

times, SR0 licensed personnel are filling the STA position."

At the time of the inspection, the NRC inspector noted that the Shift

Supervisor was assigned as the STA exclusively.

The NRC inspector

also noted that Procedure 1.2 designated the shift supervisors as " Acting

Fire Chief" in event of a fire when the designated Fire Chief (who was

the Engineering Supervisor) was not on site.

This was also a require-

ment of licensee Procedure 1.6, Revision 20 (3/25/81), " Personnel

Equipment and Safety." The licensee had committed to change this acal

tasking by July 1, 1982 (NPPD letter LQA 8100218 of June 26,1981).

There were no violations or deviations identified in this portion of

the inspection.

13.

Unresolved Items

Unresolved items are matters about which more information is required

in order to ascertain whether it is an acceptable item, a violation,

.or a deviation.

Unresolved items are discussed in this report as

indicated below:

19

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1

Ite.n Number

' Paragraph-

Description

'

1

[

8114-04

4

Torus Saddle Welds

l

8114-05.

6

Completion of Requalification Training

l

8114-06

6-

Records _of Reactivity Manipulations

1

4

8114-09

8

Review of Completed Tests / Experiments

14. Exit Interview

. Exit interviews were conducted September 4,17, and 18,1981, with those

,

personnel denoted in paragraph 1 of this report. At these exit interviews,

the NRC inspector summarized the scope and findings of their inspection.

'

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ML20032C602

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