ML20137B083

Insp Rept 50-298/85-24 on 850801-0930.Violation Noted: Inadequate Operating Procedures,Unattended & Unlocked Security Records Storage Container & Failure to Meet NRC Reportability Requirements & Perform Surveillance Testing

ML20137B083
Person / Time
Site:	Cooper
Issue date:	11/13/1985
From:	Dubois D, Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20137B054	List: ML20137B050 ML20137B057 ML20137B083
References
50-298-85-24, NUDOCS 8511260176
Download: ML20137B083 (19)

See also: IR 05000298/1985024

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.g APPENDIX B

U. S. NUCLEAR REGULATORY COMMISSION

REGION IV ,

.NRC Inspection Report: 50-298/85-24 License: DPR-46

Docket: 50-298

Licensee: Nebraska Public Power District (NPPD)

P. O. Box 499

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Columbus, NE 68601

Facility Name: Cooper Nuclear Station (CNS)

Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska

Inspection Conducted: August 1-September 30, 1985

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Inspector: . N /0//E/86

D. L. DuBois, Senior Resident Inspector, (SRI)

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Date

. Approved:

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. (gdon Chief, Proje6t- Section A,

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Dats

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eactor P oject Branch

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Inspection Summary

Inspection Conducted August 1-September 30, 1985 (Report 50-298/85-24)

Areas Inspected: Routine, unannounced inspection of operational safety

verification, monthly surveillance and maintenance observations, licensee

action on previous inspection findings, complex surveillance, inservice

inspection, p'rimary containment integrated leak rate test, reactor coolant

system hydrostatic test, plant startup from refueling, startup testing,

security activities, emergency preparedness drills, notification of an unusual

event, and followup of Licensee Event Reports. The inspection involved

260 inspector-hours onsite by one NRC inspector.

Results: Within the 12 areas inspected, five violations were identified

(inadequate operating procedure, paragraph 6; unattended and unlocked security

records storage container, paragraph 7; failure to meet NRC reportability

requirements, paragraph 10; failure to perform surveillance testing according

to procedure, paragraph 12; and inadequate evaluation of surveillance test

results, paragraph 12).

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Details

1. Persons Contacted

Principal Licensee Personnel

• P. V. Thomason,-Division Manager of Nuclear Operations

• V. L. Wolstenholm, Quality Assurance Manager

• R. Brungardt, Operations Manager

• D. Reeves, Training Manager

+*J. Sayer, Acting Technical Staff Manager

+*R. Beilke, Chemistry and Health Physics Supervisor

• E. M. Hace, Plant Engineering Supervisor

• C. R. Goings, Regulatory Compliance Specialist

D. Norvell, Maintenance Manager

P. Ballinger, Reactor Engineering Supervisor

J. Scheuerman, Lead Reactor Engineer

J. L. Peaslee, Surveillance Coordinator

M. Unruh, Maintenance Planner

+J. M. Meacham, Technical Manager

J. Flaherty, Assistant to the Plant Engineering Supervisor

R. Windham, Emergency Planning Coordinator

H. Hitch, Acting Administrative Services Manager

R. Black, Acting Operations Supervisor

R. Deatz, Engineering Specialist

NRC Personnel

+R. E. Baer, Radiation Specialist, Region IV

The NRC inspector also interviewed other licensee and contractor personnel

including operations, engineering, maintenance, and administration.

• Indicates presence at exit meeting heid August 30, 1985

+ Indicates presence at exit meeting held September 30, 1985

2. Licensee Action on Previous Inspection Findings

(Closed) 8420-02 (Violation). The design function of the Standby Gas

Treatment System (SGTS) is to reduce and maintain the secondary

containment atmospheric pressure at a minimum of 0.25 inches of water

vacuum while directing all flow through the SGTS filtration units. The

SGTS as-built design flow rate is 1780 cfm. During preoperational

testing, the licensee observed that the design function of the SGTS was

accomplished at a flow rate of 1750 cfm. Following preoperational

testing, the licensee revised the value of SGTS fan design flow rate noted

in Survellance Procedure 6.3.19.4, "SGT Charcoal Filters Leak and Fan

Capacity Test,"Section VI, Subsections D and E, from 1780 cfm to

1750 cfm. The procedure revision was approved for implementation without

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performing a 10 CFR 50.59 analysis to determine if the preoperational test

value of 1750 cfm decreased the margin of safety of the SGTS.

The licensee incorrectly assumed that the preoperational test value of

1750 cfm should be considered design flow. On January '17,1985, the

licensee approved and implemented Revision 12 to Procedure 6.3.19.4 which

restored the design flow rate value of 1780 cfm to all affected

subsections of that procedure.

This item is closed.

(Closed) 8426-01 (Viola' tion). This item concerned tiu licensee's failure

to demonstrate operability of station batteries. This failure resulted

from the following causes:

. Insufficient understanding of battery technology.

. Technically inadequate proc.edures.

. Inadequately defined Technical Specification requirements.

. Lack of procedures and records.

The licensee-developed new procedures and revised established procedures

in this area as listed below:

. 2.2.24, Revision 11, "250V DC Electrical System"

. 2.2.25, Revision 11, "125V DC Electrical System"

. 2.2.26, Revision 6, "24V DC Electrical System"

. 6.3.15.1 Revision 14, " Station Battery Quarterly Check"

. 6.3.15.2, Revision 9, " Station Battery Service Test"

. 6.3.15.2A. Revision 0, " Station Battery Performance Test"

. 6.3.15.3, Revision 13, " Station, Diesel Fire Pump, CAS, and PMIS

Battery Weekly Check"

. 7.3.27, Revision 0, " Battery Equalizing Charge"

In a letter from Mr. L. G. Kunci (f4 PPD) to fir. D. B. Vassallo (f4RC-fiRR),

dated April 26, 1985, the licensee submitted Proposed Change flo.19 to the

CflS Technical Specification. This proposed change adequately defined

requirements concerning station battery testing.

The SRI determined that the above documents presently meet battery testing

requirements found in the vendors manual; IEEE Standards 308-1978 and

450-1980; and f4RC regulatory guides. Also, completed battery surveillance

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test records were reviewed and found to meet all present acceptance

criteria and requirements.

This item is closed.

(Closed) 8426-02(Violation). This item concerns the licensee's failure

to have a procedure for controlling battery charging. The SRI's review of

this item was performed in con. junction with preceeding item, 8426-01.

Licensee corrective actions in this area are included in the above

documentaion and are considered satisfactory.

This item is closed.

(Closed) 8511-01-(Violation). This item concerned the licensee's failure

to demonstrate secondary containment integrity prior to defueling

operations conducted during the period September 22-29, 1984. The CNS

Technical Specification requires. secondary containment integrity to be

maintained during handling of irradiated fuel within the secondary .

containment. Secondary containment integrity is defined in the Technical

Specification as follows:

'"1) At least one door in each access opening is closed.

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2) The standby gas treatment system is operable.

3) All automatic ventilation system isolation valves are operable

or secured in the isolated position."

Items 1 and 3 above were maintained during refueling operations. The

SGTS, item 2 above, is determined operable if it can reduce and maintain

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the secondary containment atmospheric pressure of 0.25 inches of water

vacuum. CNS Surveillonce Procedure 6.3.10.8 is performed to prove

operability of the SGTS. Although Procedure 6.3.10.8 was performed prior

to handling irradiated fuel and appeared to be satisfactory, the licensee

later discovered that the main condenser vacuum pumps had assisted the

SGTS in drawing the required 0.25 inches of water vacuum, thus

invalidating the test results.

Licensee corrective actions included the following:

. Procedure 6.3.10.8, " Secondary Containment Leak Test," was revised on

March 15, 1985, to include additional checks to insure the mechanical

vacuum pumps are off, main steam isolation valves are closed, and no

other operating or maintenance activities are in progress which would

affect the leak test.

. Procedure 6.4.8.7, "Off Gas Loop Seal Blowdown and Fill," was revised

on February 25, 1985, to include new steps which provide directions

for ensuring that SGTS discharge lines and loop seal drains are free

of water accumulation.

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. Successful completion of surveillance procedure 6.3.10.8 prior to

conducting any subsequent irradiated fuel handling operations.

. CNS Nonconformance Report (NCR) 003337, dated January 21, 1985, was

written to document this failure to adequately demonstrate secondary

containment integrity.

. Licensee Event Report (LER)85-001, "Defueling Operations Without

Secondary Containment Integrity," was submitted to the NRC on

February 14,'1985.

. Licensed operator training relevant to this item was completed.

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. NCR 003656, dated April 18, 1985, was written to document SGTS design

inadequacies which were identified by an independent

architect / engineering firm retained by the licensee.during March,

1985. As a result of that independent system design review, the

following modifications were made to the SGTS:

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a) Expansion sleeves located at the SGTS fans discharges and at the

crossover line between trains were replaced.

b) Additional bracing was added to the crossover line,

c) SGTS housing drain lines isolation valves were installed.

. LER 85-002, " Standby Gas Treatment System Design Deficiencies," was

submitted to the NRC on April 25, 1985.

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The SRI reviewed the licensee corrective actions listed above and

!- determined that all actions were completed satisfactorily prior to

refueling the reactor vessel during July 1985.

This item is closed.

3. Complex Surveillance and Inservice Inspection

The SRI completed a review of five procedures used by-the licensee to

perform complex safety-related Technical Specification required testing.

His review included the following documents:

. Surveillance Procedure 6.3.1.3, Revision 8, " Primary Containment

Integrated Leakage Test."

. Maintenance Procedure 7.0.8, Revision 0, " Hydrostatic Leak Test."

! . Special Test Procedure 85-1L, "ASME Class 1 Hydrostatic Test," dated

March 1, 1985.

. Special Test Procedure 85-15, " Recirculation System Flow Control

System," dated April 10, 1985.

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.. Surveillance Procedure 6.3.4.3,' Revision 20, "CS, RHR, and Diesel

Auto Start and Loading."

The NRC _ inspector witnessed the licensee's performance of the above

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Hydrostatic. test--August 1-2, 1985.

. Primary Containment Integrated Leak Test--August 6-13, 1985.

.. Recirculation System Flow Control Test--periodically during the power

ascension program.

. Diesel-Start and Loading Sequence--August 14, 1985 .

These reviews and observations verified that:

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. . . Testing was performed using approved procedures which were consistent

with regulatory requirements, industry standards, and the Technical

Specification.

,1 . Permanent or temporary procedure revisions were accomplished

, according to administrative requireraents and controls.

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.. Qualified personnel conducted the tests and performed the final

reviews and approvals of completed test data.

.- _The official test copy was available and used by test personnel.

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- 1.' Procedures contained the purpose, objectives, references,

a prerequisites, test equipment, precautions, limitations, and

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acceptance criteria. QA/QC hold points were established as

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appropriate.

. . Test equipment required by the procedures was calibrated and in

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. Procedures provided sufficient direction to accomplish necessary

evolutions.

. Systems were returned to normal lineup following completion of

testing.

The SRI performed independent measurements and calculations to verify the

licensee's data and test results.

No violations or deviations were identified in this area.

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4. Primary Containment Integrated Leak Rate Test

The SRI reviewed the licensee's procedure for a primary containment

integrated leak rate test-(ILRT) and witnessed the performance of the test

.as documented in paragraph 3 of this report. Additional reviews'and

observations included:

. A pretest' inspection of the primary containment building, containment '

inboard and outboard. isolation system valves, and the special systems

used to pressurize and vent the containment building during the ILRT.

. An ' independent valve position verification check prior to testing and

following satisfactory completion of the ILRT.

. ILRT walkdown of the control room consoles with particular attention

given to ready identification of the test boundary.

. Areas surrounding the primary containment were posted with

appropriate warning' signs and radiological requirements.

. Independent observation of containment parameters.

. Calculation of leakage rates and comparisons of results with licensee

data.

. " Monitoring of radioactivity release parameters during

depressurization.

. Observation of test instrumentation response and calculation of

leakage rate during the controlled leakage portion of the ILRT.

. Verification that Technical Specification requirements were met and

maintained during testing.

The CNS Technical Specification, Section 4.7.A.2.a. states that primary

containment integrity is confirmed if the leakage rate does not exceed the

equivalent of 0.635 percent of the primary containment volume per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

at 58 psig. The measured results of this ILRT was 0.3460 percent / day at

58 psig.

These reviews, observations, and independent verifications were conducted

to ensure that the ILRT was perfonned in accordance with the requirements

established in the CNS Operating License and Technical Specification.

No violations or deviations were identified in this area.

5. Reactor Coolant System Hydrostatic Test

The SRI reviewed the licensee's procedures for a reactor coolant system

hydrostatic test and witnessed the parformance of the test as documented

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in paragraph 3 of this report. Previous reviews were performed and

documented in NRC Inspection Report 50-298/85-18. This inspection

consisted of test performance observations including:

. Verification that the pressure boundary isolation valves were

maintained during the test.

. The reactor coolant system was protected from overpressure by code

safety valves set in the range of 1240-1250 psig.

. Water quality met chemistry requirements.

. The reactor coolant system was vented during filling operations.

. Pressurization temperature was kept above the' nil ductility

, transition temperature.

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. Hydrostatic test pressure was maintained at 1080 + 15 psig for a

minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to performing pressure boundary vessel,

piping,; pumps, valves and flanges inspections.

. Pressurization /depressurization and heatup/cooldown rates met

Technical Specification requirements.

. Test instrumentation response and data measurements.

. Verification of proper safety-related systems pressure switch

actuations/ reset for those switches that sense reactor coolant system

pressure in order to perform their safety functions.

. Verification that maintenance performed to reduce leakage did not'

. negate the performance of the test or test results.

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.._ On-the-spot procedure changes were approved and implemented as

permitted by administrative procedures.

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. Procedural steps accommodated the performance of the 10-year ISI

Class I hydrostatic test as well as testing all piping and components

replaced, modified, or repaired during the pipe replacement outage.

. A drywell inspection prior to and following the test.

. Availability of safety systems and maintenance of limiting conditions

for operations required by the Technical Specification,

i This hydrostatic test was performed in accordance with Section XI of the

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ASME Boiler and Pressure Vessel Code,1974 Edition. Test pressure was

1.10 times the CNS nominal operating pressure of 1005 psig.

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hydrostatic test was performed in accordance with licensee. commitments,.

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Specification requirements.

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No violations or deviations were identified in this area. .

6. Plant Startup From Refueling and Startup Testing

The SRI performed prestartup inspections relating to this area during

previous inspection periods. Results of those inspections were documented

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in NRC Inspection Reports 50-298/85-08, 85-11, 85-15, 85-16, and 85-18.

The following areas were subject to those inspections:

. Design changes and modifications

. Station procedures and drawings

. Document controls

. Vendor technical information program

. Plant maintenance

. Equipment / systems surveillance testing

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. Preoperational testing

. Inservice inspection and test program

. Power ascension testing

. Training

. Licensee management and safety committees

. Systems / components lineup verifications prior to and following

special tests and modifications

During-this inspection period, the SRI performed additional reviews,

independent verifications, and observations of the-following activities:

.. Inservice inspection of the reactor coolant ' system and primary

containment.

. Safety-related systems lineup verifications including automatic

depressurization, residual heat removal, core spray, primary

containment isolation, reactor core isolation cooling, standby liquid

control, emergency power distribution, nuclear. instrumentation, and

plant radiation monitoring systems.

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. Technical Specification surveillance test activities including source

range monitoring, high pressure coolant injection, diesel generator

load sequencing, mainsteam line isolation, air ejector high radiation

isolation, standby gas treatment, residual heat removal, and control

rod drive systems.

. Preoperational testing of high pressure coolant injection, reactor

core isolation cooling, reactor recirculation, and reactor level

control systems.

. .Prestartup checks.

.- Initial plant startup and power operation.

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. Power ascension testing of reactor recirculation flow control,

reactor feedwater and level control, and main turbine and generator

systems including auxiliary systems and components of each.

The SRI witnessed a reactor startup on August 20, 1985. This startup

followed an 11-month outage to replace all of the reactor coolant system

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piping. The following areas were observed prior to, during, and following

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that startup:

. A combined onsite and offsite safety committee's prestartup review

meeting on August 1-2, 1985.

. A joint meeting of the plant staff and NPPD Board of Directors

Nuclear Subcommittee on August 16, 1985, to determine the readiness

of the plant for startup.

. lianagement authorization for startup.

. Operable status of required systems.

. Systems and master startup checklists status.

. Crew shift manning.

. Usage of approved updated procedures.

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. Performance of startup and physics testing.

. Reactor and instrumentation response.

. Performance of special test procedures for equipment and systems

requiring operating reactor steam pressure and temperature

conditions.

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On August 20, 1985, the SRI reviewed Licensee Procedure 2.1.1, " Cold

Startup Procedure," Revision 39. He observed that the duty shift

supervisor had incorrectly documented that reactor water chemistry was

adequate for startup. The SRI had previously reviewed the chemistry

report and observed that reactor water pH was less than the Technical

Specification minimum permissible limit of 5.6. The shift supervisor was

immediately notified of his mistake and took prompt action to have another

sample taken, the results of which were within specifications. However,

10 CFR Part 50, Appendix B, Criterion V, requires that quality procedures

include quantitative or qualitative acceptance criteria for determining

that important activities have been satisfactorily accomplished. Neither

Procedure 2.1.1 nor the chemistry sample data sheet included criteria for

making that determination. This failure of the procedure and the sample

sheet to include quantitative acceptance criteria is an apparent violation

(50-298/8524-01).

These inspections and observations were perfonned to verify that plant

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systems and components were preoperational tested and aligned prior to

startup; approach to criticality, heatup, and power ascension were

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conducted in accordance with approved procedures; operational tests of

systems were conducted if required; and the reactor, steam plant and

electrical generation systems responded as designed.

l 7. Security Activities

The SRI acted as an observer during a licensee safeguards contingency plan

drill conducted by NRC Region IV security specialists on August 13, 1985.

This drill and other security inspection activities were documented in NRC

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Inspection Report 50-298/85-23.

During a routine security tour and inspection conducted on August 28,

1985, the SRI observed that safeguards material storage containers were

left unlocked and unattended in the licensee's security supervisor's

office. The security group was immediately notified of this occurrence

and dispatched a person to attend the area until the containers were

properly secured. 10CFR73.21(d)(2),requiresthatunattendedsafeguards

information be stored in a locked security container. The licensee's

failure to meet this requirement is an apparent violation

(50-298/8524-02).

8. Emergency Preparedness Drills

On July 26, 1985, an NRC consultant witnessed a small scale drill (Alert)

conducted by the licensee. The purpose was to test the response

capabilities of the operations support centers (OSCs). The scenario

required activation of the OSCs, notification of station personnel,

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l control room, and the exhibition of various tools and equipment required

i to correct equipment problems identified in the scenario.

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The SRI observed an emergency drill that was conducted and performed by

the licensee on August 10, 1985. Drill participants included only those

management personnel who are assigned to the control room (CR), technical

support center (TSC), and emergency operations facility (E0F) during an

emergency. The purpose of the drill was to demonstrate the ability to

properly classify the event, respond to the emergency situations presented

in the scenario, verify operability of communications equipment, and

properly activate and man the emergency facilities noted above. No major

problems were identified. The SRI was in attendance at a critique

conducted imediately following the drill.

' Additional licensee drills will be conducted in the immediate future, and

< will require full participation from all plant and corporate personnel

assigned' specific duties during emergency situations.' Also, activation of

all emergency response facilities and equipment will be required. The

annual full-scale emergency preparedness exercise is scheduled during

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October 1985, and NRC observations will be documented in NRC Inspection

Report 50-298/85-28.

No violations or deviations were identified in this area.

9. Notification of an Unusual Event

At 5:00 p.m. on August 24, 1985, the CNS operations shift supervisor

observed the actuation of a control room annunciator that indicated

trouble with the meteorological (MET) monitoring system. Subsequent

investigation indicated that the MET computer was on-line but output data

was not available. Following several unsuccessful attempts to reinitiate

the data function, the shift supervisor assumed the role of Emergency

Director and declared a Notification of Unusual Event (N0VE) at 5:30 p.m.

This decision was based upon CNS Emergency Plan Implementation Procedure

(EPIP)5.7.1, Attachment"C,""ClassificationGuide,"Section6.1,which

requires the licensee to classify and initiate a NOUE when a significant

lossofmeteorologicalassessmentcapabilityoccurs(e.g.,acompleteloss

of meteorological instrumentation).

At 5:54 p.m., the data acquisition and display functions of the MET

computer were restored. The NOUE was terminated by the Emergency Director

at 6:24 p.m. on August 24, 1985, af ter verifying that all MET

instrumentation continued operating in a stable manner.

The specific cause of the malfunction could not be determined, however,

certain indications pointed to a momentary loss or reduction of the

computer power supply voltage. A Design Change Request (DCR) 85-84 was

submitted for review and approval which would result in supplying the MET

computer from an uninterruptible power source. Procedure 5.7.1,

Section 6.1.2 was revised on August 29, 1985, to include the requirements

that communications with the National Weather Service must be lost

simultaneously with a loss of the MET system before it would be necessary

for the licensee to declare a NOVE.

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.The SRI interviewed shift personnel and the CNS Emergency Planning

Coordinator concerning this event. Also, he performed a review of the

following licensee logs, procedures, and reports:

. Control room logs for complete and timely entry of significant

infonnation pertaining to the event.

. Procedures and checklists applicable to the declaration and

termination of the event.

. Licensee followup report to the NRC in a letter from Mr. P. V.

Thomason (NPPD) to Mr. R. D. Martin (NRC-RIV) dated August 26, 1985.

. 5.7.1, Revision 4, " Emergency Classification"

. 5.7.2, Revision 4 " Notification of Unusual Event"

. 5.7.6, Revision 5, " Notification"

. 5.7.22, Revision 6, " Communications"

. 5.7.28, Revision 1, " Emergency Director"

The reviews and discussions were conducted to ensure that licensee

personnel performed all actions required by the CNS emergency procedures,

Technical Specification, and Emergency Plan.

No violations or deviations were identified in this area.

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10. Licensee Event Reports Followup

The following LERs are closed on the basis of the SRI's inoffice review,

review of licensee documentation, and discussions with licensee personnel:

. LER 85-001, "Defueling Operations Without Secondary Containment

Integrity."

. LER 85-002, " Standby Gas Treatment System Design Deficiencies."

The SRI documented in NRC Inspection Report 50-298/85-18, paragraph 7, his

observations and reviews cencerning an irradiated fuel bundle abnormal

handling operation. Specifically, an unchanneled fuel bundle was

suspendedandmovedwiththefuelbundlehandle.(bail)caughtandheldin

place by an outside corner of the fuel grapple head rather than being

properly grappled by the internal engagement mechanism.

The licensee made necessary temporary changes to refueling Procedure 10.25

prior to continuing fuel bundle movements within the spent fuel pool.

hCR 004688 was initiated on July 24, 1985, to document the occurrence.

However, the licensee did not submit a written report of this occurrence

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- - to the: NRC.' 10.CFRPart50.73(a)(2)(ii)(c)requiresthelicenseeto

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submit an LER within 30' days if any condition or event results in the-

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nuclear power plant being in a condition .not covered by the plant's

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3 operating:and emergency procedures. A review of licensee normal, ,

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J ' Jabnormal, and emergency procedures concluded that no procedure exists that

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instructions to recover from such an occurrence. This is an apparent

violation (50-298/8524-03), i

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11. Operational Safety Verification

, The SRI observed control room operations, instrumentation, controls,

. reviewed plant logs and records, conducted discussions with control room

personnel, and performed system walk-downs to verify that:

. Minimum shift manning requirements were met.

. Technical Specification requirements were' observed.

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. Plant operations were conducted using approved procedures.

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. Plant logs and records were complete, accurate, and indicative of

. actual system conditions and configurations.

. System pumps, valves, control switches, and power supply breakers

were properly aligned.

. Licensee systems lineup procedures / checklists, plant drawings, and

as-built configuratic. s were 'in agreement.

. Instrumentation was accurately displaying process variables and ,

protection system status to be within permissible operational limits

for operation.

. Plant equipment that was discovered to be inoperable or was removed

from service for maintenance was properly identified, redundant

equipment was verified to be operable, and applicable limiting

conditions for operation were identified and maintained.

. Equipment safety clearance records were complete and indicated that

affected components were removed from and returned to service in a

correct and approved manner.

. Maintenance work requests were initiated for equipment discovered to

require repair or routine preventive upkeep, appropriate priority was

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assigned, and work commenced in a timely manner.

. Plant equipment conditions such as cleanliness, leakage, lubrication,

and cooling water were controlled and adequately maintained.

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-. Areas of the plant were clean, unobstructed, and free of fire

hazards. Fire suppression systems and emergency equipment were

maintained in a condition of readiness.

. Security measures and. radiological controls were adequate.

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The SRI performed lineup verifications of-the.following systems:

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.. Reactor Recirculation-

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. 'High' Pressure Coolant Injection

. Reactor Core Isolation Cooling .

.. Core Spray

. Residual Heat ~ Removal

. Main Steam-

. Plant Radiation Monitoring

.- Nuclear Instrumentation

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.- Primary Containment -Isolation

. Automatic Depressurization

. Reactor Head Vents ,

. 4160V AC Emergency Power Distribution

. Emergency Diesel Generators 1

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. Standby Liquid Control

The tours, reviews, and observations were conducted to verify that

facility operations were performed in accordance with the requirenents

established in the CNS Operating License and Technical Specification.

No violations or deviations were identified in this area.

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12. Monthly Surveillance Observations

The SRI observed Technical Specification' required surveillance tests.

These observations verified that:

. . Tests were accomplished by qualified persor.nel in accordance with

approved procedures.

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Procedures conformed to Technical Specification requirements.

. -Test prerequisites were completed including conformance with

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applicable limiting conditions for operation, required administrative

approval', and availability of calibrated test equipment.

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Test data were reviewed for completeness, accuracy, and conformance

with established criteria and Technical Specification requirements.

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Deficiencies were corrected in a timely manner.

. The system was returned to service.

The following surveillance tests were selected and observed:

6.1.21, "SRM Calibration and Functional Test (Reactor Not In Run)"

6.2.4.1, " Daily Surveillance (Technical Specifications)"

6.3.1.3, " Primary Containment Integrated Leakage Test"

6.3.4.3, "CS, RHR, and Diesel Auto Start and Loading"

7.0.8, " Hydrostatic Leak Test"

85-1L, "ASME Class 1 Hydrostatic Test"

85-1N, "High Pressure Coolant Injection"

85-1R, "Feedwater Control System"

85-15, " Recirculation System Flow' Control System"

85-1Q, "SRM/RH/APRM Overlap Verification"

At 7:20 a.m. on August 29, 1985, the SRI observed that intermediate range

nuclear instrumentation channels D and F OPERATE / TEST switches were not in

the OPERATE position. The SRI reviewed completed CNS Surveillance

Procedure 6.1.3, "APRM System Excluding 15% Trip Function Test," that was

performed just prior to the preceding observation, and he noted that

step 10 required the affected switches to be placed in the OPERATE

position. The SRI immediately notified the shift supervisor of his

observations and the switches were returned to the OPERATE position. The

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failure to return the test switches to normal as directed by

Procedure 6.1.3 is an apparent violation (50-298/8524-04).

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During a review of completed Surveillance Test 6.2.4.1, " Daily

Surveillance (Technical Specifications)," performed by the licensee on-

--September 23, 1985, the SRI noted an incorrect logged data value.

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Specifically, the control room vent monitor gaseous activity reading that

was logged following completion of the source test was the same value as

that recorded during the source check instead of the lower background

value that existed prior to testing. The shift supervisor was notified of

the error and in accompaniment with the SRI verified the actual value to

be a background reading. The entered data was properly corrected shortly

thereafter. The SRI subsequently verified that the incorrect data entry

had been reviewed by the control room supervisor and shift supervisor and

neither of them identified it. The failure to adequately document, re-

view,andevaluatetestresultsisanapparentviolation(50-298/8524-05).

The reviews and observations were conducted to verify that facility

surveillance operations were performed in accordance with the requirement.

established in the CNS Operating License and Technical Specification.

13. Monthly 11aintenance Observation

The SRI observed preventive and corrective maintenance activities on

portions of the following systems / components:

. Service Water Pumps

. Service Water Booster Pumps

The observations were conducted to verify that:

. Limiting conditions for operation were met.

. . Redundant equipment was operable.

. Equipment was adequately isolated and safety tagged.

. Appropriate administrative approvals were obtained prior to

commencement of work activities.

. Work was performed by qualified personnel in accordance with approved

procedures.

. Radiological controls, cleanliness practices, and appropriate fire

prevention precautions were implemented and maintained.

. Quality control checks and postmaintenance surveillance testing were

performed as required.

. Equipment was properly returned to service.

These reviews and observations were conducted to verify that facility

maintenance operations were performed in accordance with the requirements

established in the CNS Operating I' cense and Technical Specification.

No violations or deviations were identified in this area.

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14. Exit Meetings

Exit meetings were conducted at the conclusion of.each portion of the

inspection. The NRC inspector summarized the scope and findings of each

inspection. segment at those meetings.

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