ML20137B083
| 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: |
| References | |
| 50-298-85-24, NUDOCS 8511260176 | |
| Download: ML20137B083 (19) | |
See also: IR 05000298/1985024
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APPENDIX B
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U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
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.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:
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D. L. DuBois, Senior Resident Inspector, (SRI)
Date
. Approved:
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. (gdon Chief, Proje6t- Section A,
Dats
eactor P oject Branch
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8511260176 851114
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Inspection Summary
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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.
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Technically inadequate proc.edures.
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Inadequately defined Technical Specification requirements.
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Lack of procedures and records.
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The licensee-developed new procedures and revised established procedures
in this area as listed below:
2.2.24, Revision 11, "250V DC Electrical System"
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2.2.25, Revision 11, "125V DC Electrical System"
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2.2.26, Revision 6, "24V DC Electrical System"
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6.3.15.1 Revision 14, " Station Battery Quarterly Check"
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6.3.15.2, Revision 9, " Station Battery Service Test"
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6.3.15.2A. Revision 0, " Station Battery Performance Test"
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6.3.15.3, Revision 13, " Station, Diesel Fire Pump, CAS, and PMIS
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Battery Weekly Check"
7.3.27, Revision 0, " Battery Equalizing Charge"
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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
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containment. Secondary containment integrity is defined in the Technical
Specification as follows:
'"1)
At least one door in each access opening is closed.
2) The standby gas treatment system is operable.
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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
the secondary containment atmospheric pressure of 0.25 inches of water
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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
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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
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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
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conducting any subsequent irradiated fuel handling operations.
CNS Nonconformance Report (NCR) 003337, dated January 21, 1985, was
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written to document this failure to adequately demonstrate secondary
containment integrity.
Licensee Event Report (LER)85-001, "Defueling Operations Without
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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
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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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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,
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SGTS housing drain lines isolation valves were installed.
LER 85-002, " Standby Gas Treatment System Design Deficiencies," was
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submitted to the NRC on April 25, 1985.
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The SRI reviewed the licensee corrective actions listed above and
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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
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Integrated Leakage Test."
Maintenance Procedure 7.0.8, Revision 0, " Hydrostatic Leak Test."
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Special Test Procedure 85-1L, "ASME Class 1 Hydrostatic Test," dated
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March 1, 1985.
Special Test Procedure 85-15, " Recirculation System Flow Control
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System," dated April 10, 1985.
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Surveillance Procedure 6.3.4.3,' Revision 20, "CS, RHR, and Diesel
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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.
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Primary Containment Integrated Leak Test--August 6-13, 1985.
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Recirculation System Flow Control Test--periodically during the power
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ascension program.
Diesel-Start and Loading Sequence--August 14, 1985 .
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These reviews and observations verified that:
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. Testing was performed using approved procedures which were consistent
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with regulatory requirements, industry standards, and the Technical
Specification.
Permanent or temporary procedure revisions were accomplished
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according to administrative requireraents and controls.
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Qualified personnel conducted the tests and performed the final
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reviews and approvals of completed test data.
_The official test copy was available and used by test personnel.
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Procedures contained the purpose, objectives, references,
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prerequisites, test equipment, precautions, limitations, and
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acceptance criteria. QA/QC hold points were established as
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appropriate.
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. Test equipment required by the procedures was calibrated and in
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service.
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Procedures provided sufficient direction to accomplish necessary
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evolutions.
Systems were returned to normal lineup following completion of
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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 '
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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
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following satisfactory completion of the ILRT.
ILRT walkdown of the control room consoles with particular attention
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given to ready identification of the test boundary.
Areas surrounding the primary containment were posted with
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appropriate warning' signs and radiological requirements.
Independent observation of containment parameters.
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Calculation of leakage rates and comparisons of results with licensee
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data.
" Monitoring of radioactivity release parameters during
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depressurization.
Observation of test instrumentation response and calculation of
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leakage rate during the controlled leakage portion of the ILRT.
Verification that Technical Specification requirements were met and
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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
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maintained during the test.
The reactor coolant system was protected from overpressure by code
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safety valves set in the range of 1240-1250 psig.
Water quality met chemistry requirements.
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The reactor coolant system was vented during filling operations.
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Pressurization temperature was kept above the' nil ductility
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transition temperature.
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Hydrostatic test pressure was maintained at 1080 + 15 psig for a
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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
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Technical Specification requirements.
Test instrumentation response and data measurements.
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Verification of proper safety-related systems pressure switch
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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'
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negate the performance of the test or test results.
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On-the-spot procedure changes were approved and implemented as
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permitted by administrative procedures.
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Procedural steps accommodated the performance of the 10-year ISI
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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.
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Availability of safety systems and maintenance of limiting conditions
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for operations required by the Technical Specification,
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This hydrostatic test was performed in accordance with Section XI of the
ASME Boiler and Pressure Vessel Code,1974 Edition. Test pressure was
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1.10 times the CNS nominal operating pressure of 1005 psig.
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This inspection was conducted to ensure that the primary system
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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.
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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
in NRC Inspection Reports 50-298/85-08, 85-11, 85-15, 85-16, and 85-18.
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The following areas were subject to those inspections:
Design changes and modifications
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Station procedures and drawings
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Document controls
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Vendor technical information program
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Plant maintenance
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Equipment / systems surveillance testing
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Preoperational testing
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Inservice inspection and test program
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Power ascension testing
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Training
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Licensee management and safety committees
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Systems / components lineup verifications prior to and following
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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
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containment.
Safety-related systems lineup verifications including automatic
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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
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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
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core isolation cooling, reactor recirculation, and reactor level
control systems.
.Prestartup checks.
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Initial plant startup and power operation.
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Power ascension testing of reactor recirculation flow control,
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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
piping. The following areas were observed prior to, during, and following
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that startup:
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A combined onsite and offsite safety committee's prestartup review
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meeting on August 1-2, 1985.
A joint meeting of the plant staff and NPPD Board of Directors
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Nuclear Subcommittee on August 16, 1985, to determine the readiness
of the plant for startup.
lianagement authorization for startup.
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Operable status of required systems.
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Systems and master startup checklists status.
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Crew shift manning.
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Usage of approved updated procedures.
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Performance of startup and physics testing.
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Reactor and instrumentation response.
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Performance of special test procedures for equipment and systems
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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
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startup; approach to criticality, heatup, and power ascension were
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.
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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
Inspection Report 50-298/85-23.
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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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establishment and maintenance of communications between OSCs and the
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control room, and the exhibition of various tools and equipment required
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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
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assigned' specific duties during emergency situations.' Also, activation of
all emergency response facilities and equipment will be required.
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annual full-scale emergency preparedness exercise is scheduled during
October 1985, and NRC observations will be documented in NRC Inspection
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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
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infonnation pertaining to the event.
Procedures and checklists applicable to the declaration and
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termination of the event.
Licensee followup report to the NRC in a letter from Mr. P. V.
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Thomason (NPPD) to Mr. R. D. Martin (NRC-RIV) dated August 26, 1985.
5.7.1, Revision 4, " Emergency Classification"
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5.7.2, Revision 4 " Notification of Unusual Event"
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5.7.6, Revision 5, " Notification"
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5.7.22, Revision 6, " Communications"
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5.7.28, Revision 1, " Emergency Director"
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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.
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No violations or deviations were identified in this area.
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
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Integrity."
LER 85-002, " Standby Gas Treatment System Design Deficiencies."
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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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nuclear power plant being in a condition .not covered by the plant's
operating:and emergency procedures. A review of licensee normal,
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' Jabnormal, and emergency procedures concluded that no procedure exists that
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, covers the fuel. handling methodology described above, nor do they provide
instructions to recover from such an occurrence. This is an apparent
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11. Operational Safety Verification
The SRI observed control room operations, instrumentation, controls,
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. 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.
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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
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actual system conditions and configurations.
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System pumps, valves, control switches, and power supply breakers
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were properly aligned.
Licensee systems lineup procedures / checklists, plant drawings, and
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as-built configuratic. s were 'in agreement.
Instrumentation was accurately displaying process variables and
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protection system status to be within permissible operational limits
for operation.
Plant equipment that was discovered to be inoperable or was removed
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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
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affected components were removed from and returned to service in a
correct and approved manner.
Maintenance work requests were initiated for equipment discovered to
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require repair or routine preventive upkeep, appropriate priority was
assigned, and work commenced in a timely manner.
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Plant equipment conditions such as cleanliness, leakage, lubrication,
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and cooling water were controlled and adequately maintained.
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Areas of the plant were clean, unobstructed, and free of fire
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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:
Reactor Recirculation-
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'High' Pressure Coolant Injection
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Reactor Core Isolation Cooling
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Residual Heat ~ Removal
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Main Steam-
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Plant Radiation Monitoring
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Nuclear Instrumentation
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Primary Containment -Isolation
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Automatic Depressurization
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Reactor Head Vents
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4160V AC Emergency Power Distribution
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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
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The SRI observed Technical Specification' required surveillance tests.
These observations verified that:
. Tests were accomplished by qualified persor.nel in accordance with
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approved procedures.
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Procedures conformed to Technical Specification requirements.
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-Test prerequisites were completed including conformance with
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applicable limiting conditions for operation, required administrative
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approval', and availability of calibrated test equipment.
Test data were reviewed for completeness, accuracy, and conformance
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with established criteria and Technical Specification requirements.
Deficiencies were corrected in a timely manner.
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The system was returned to service.
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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
failure to return the test switches to normal as directed by
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Procedure 6.1.3 is an apparent violation (50-298/8524-04).
During a review of completed Surveillance Test 6.2.4.1, " Daily
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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
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Service Water Booster Pumps
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The observations were conducted to verify that:
Limiting conditions for operation were met.
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. Redundant equipment was operable.
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Equipment was adequately isolated and safety tagged.
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Appropriate administrative approvals were obtained prior to
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commencement of work activities.
Work was performed by qualified personnel in accordance with approved
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procedures.
Radiological controls, cleanliness practices, and appropriate fire
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prevention precautions were implemented and maintained.
Quality control checks and postmaintenance surveillance testing were
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performed as required.
Equipment was properly returned to service.
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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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