IR 05000317/1978038
| ML19289D656 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/09/1979 |
| From: | Conte R, Dante Johnson, Zimmerman R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19289D651 | List: |
| References | |
| 50-317-78-38, NUDOCS 7903140025 | |
| Download: ML19289D656 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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0FFICE OF INSPECTION AND ENFORCEMENT
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Regio-n I 50-317/78-38 Report No. 50-318/78-34
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50-317 Docket No. 50-318 DPR-53 C
License No. DPR-69 Priority Category c
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Licensee:
Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, Maryland 21203 Facility Name:
Calvert Cliffs Nuclear Power Plant Units 1 and 2 Inspection at:
Lusby, Maryland dnd Management Meeting at:
NRC Region I Office Inspection conducted:
December 12-15, 1978 and Management Meeting c nd ted:
Inspectors:
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'R.J. Cony,ReactfdrInspector ida1(e igned Y WL%ba l
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R R iD rm
, Rdictor Inspector
/dat/e si ned
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Y 7f D. F. JoInson, Reactor Inspector
'dat6 sighed Approved by:
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'H. B. Kister, Chief, Nuclear Support date signed Section No. 2, R0 & kS Branch Inspection Summary:
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Inspection on December 12-15,1978 (Combined Rt. cort Nos. 50-317/78-38; 00-318/78-34)
Areas Inspected: Routine, unannounced inspection of administrative controls of safety related maintenance; documented activities associated with completed safety related maintenance; qualification records of selected individuals who performed safety related maintenance and calibrations; surveillance calibrations of safety related components and equipment required by Technical Specifications including associated procedures and completed data; performance of selected surveillance calibrations; calibration data for selected measuring / test equipment; calibrations required by Technical Specifications of components and equipment associated with safety related systems and/or functions including procedures and completed data; licensee action on previous inspection findings; review of licensee event reports; review of plant operation and facility tour; personnel identification and dosimetry. The inspection involved 40 inspector-hours onsite for Unit 1 and 37 inspector-hours onsite for Unit 2 by three regional based inspectors.
Resul ts : Of the 11 areas inspected, no items of noncompliance were found in 9 areas; 2 apparent items of noncompliance were found in two areas (Infraction - failure to comply with Technical Specification 3.0.4, Paragraph 3; and failure to include or reference instructions for functional test and document functional acceptability for safety related maintenance, Paragraph 6).
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In addition, on January 4,1979, a management meeting was held with Baltimore Gas and Electric Company representatives at the NRC Region I Office to discuss the Office of Inspection and Enforcement Inspection program at the Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
The meeting involved 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> by five members of the Region I staf.
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DETAILS
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1.
Persons Contacted E. Baur, Electrical Shop Foreman
- J. Carroll, Performance Engineer - Operations
- J. Carter, Supervisor Quality Control Unit
- J. Hill, Shift Supervisor
- J. Lemons, Nuclear Plant Engineer - Maintenance M. Miernicki, Performance Engineer - Maintenance J. Moreira, Instrument and Control Foreman, Unit 2
- P.
Rizzo, Assistant General Foreman for Electrical and Instrumen-tation Maintenance
- L. Russell, Chief Engineer D. Sheranko, Maintenance Foreman, Unit 1 W. Whitaker, Maintenance Foreman, Unit 2 The inspector also interviewed other licensee employees, including members of the technical and operations staff.
- denotes those present at the esit interview.
2.
Licensee Action on Previous Irspection Findings (Closed) Infraction (317/78-22-03, 318/78-16-03): The inspector verified that a training session was held on August 14, 1978, during which Rad-Chem technicians were re-instructed in the use of procedure RCP-2-402.
Special emphasis was placed on the use of rejection stickers and the prevention of use of instrumentation that is out-of-calibration.
In addition, instruments which have been rejected are now segregated and not accessible for use.
(Closed) Infraction (317/78-22-04, 318/78-16-04): The licensee has prepared procedure RCP-3-605 " Supply and Utilization of Quality Breathing Air," Revision 0, November 16, 1978, which provides for the control and maintenance of respirable air.
(Closed) Infraction (318/78-19-03): Disciplinary action was taken with personnel involved in the event where the SIT was at a level greater than allowed by T.S. and a letter reviewing this event and reminding them of their responsibilities in operating the facility in strict compliance with Technical Specification requirements was sent to all Shift Supervisors and Senior Control Room Operators (See Details, Paragraph 3).
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(Closed) Infraction (318/78-20-01): This item describes the failure by the licensee to obtain grab samples of the containment atmosphere while the containment particulate radioactivity monitoring system was out of service. The licensee issued a letter to all Control Room Operators with copies routed to Shift Supervisors and Senior Control Room Operators emphasizing the responsibilities of all licensed operators to be aware of Limiting Conditions for Operation and verifying compliance with action statements of the Technical Specifications (See also Details, Paragraph 3).
3.
Failure to Comoly with Technical Specifications During a routine reactor trip recovery, the reactor was made critical with Technical Specification action statements 3.7.3.1 and 3.7.5.1 in effect. The No. 11 component cooling system and the No.11 saltwater sub-system were inoperable.
The operators' actions taken in this event are in violation of Technical Specification 3.0.4 Limiting Condition for Operation.
Similar events involving the operation of the facility in violation of Technical Specification requirements have been cited previously in NRC inspections 50-317/
78-26; 50-318/78-20; and 50-317/78-25; and, 50-318/78-19 conducted on September 5-8, 1978, and September 12-15, 1978, respectively.
This item of noncompliance is categorized as an Infraction (78-38-09).
4.
Review of Licensee Event Reoorts (LER's)
The inspector reviewed LER's to verify that details of the events were clearly reported including the accuracy of the description of cause and adequacy of corrective action.
The inspector determined whether further information was required from the licensee and whether the event had generic implications.
The following Unit 1 LER's were reviewed:
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LER 78-23/3L, dated November 6,1978; Failure of the contain-ment atmosphere particulate monitor due to moisture accumulated as a result of high ambient containment humidity with subse-quent condensation inside the detector.
LER 78-42/3L, dated September 7, 1978; Failure of a thermal
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detector in the halon fire suppression system in the cable spreading room.
Failure was caused by a ground in the detecto.
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LER 78-43/lT, dated September 26, 1978; a small fish impingement developed on the travelling screens as a result of oxygen depletion in the bottom waters in the intake area.
LER 78-44/3L, dated October 20, 1978; Flow rate of the spent
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fuel pool exhaust fans were discovered to be less than Technical Specification requirements. The cause was due to excessive leakage of the discharge dampers.
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LER 78-45/3L, dated October 20, 1978; Isolation transformer
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failure in the SGIS channel resulting in a less conservative trip point.
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LER 78-46/3L, dated September 12, 1978; Radioactive liquid waste release rate exceeded Environmental Technical Specification limits.
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LER 78-47/3L, dated October 20, 1978; Loss of No. 21 13 KV service bus due to a tripped breaker.
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LER 78-48/3L, dated Cctober 26, 1978; No. 36 CEA dropped into the core while exercising due to voltage spike in the 15 volt DC power supply.
LER 78-49/3L, dated Notember 16, 1978; Containment particulate
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and gaseous radiation runitors were discovered to be inoperable due to a pump failure.
LER 78-50/3L, dated October 30, 1978; Release ratio of radio-
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active liquid effluents exceeded Environmental Technical Specification limits.
LER 78-51/3L, dated November 17, 1978; Failure of a solenoid
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valve on main steam isolation valve hydraulic accumulators rendering No. 11 MSIV inoperable.
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LER 78-52/3L, dated November 3,1978; The No.11 auxiliary feed pump was declared inoperable due to a faulty throttle valve.
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LER 78-53/3L, dated December 7, 1978; The containment particu-late radiation monitor failed low due to disconnected signal cabl.
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The following Unit 2 LER's were reviewed:
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LER 78-20/3L, dated July 20, 1978; Failure of Channel C of the reactor protection system due to a failed diode in the regulating circuit.
LER 78-27/3L, dated September it, 1978; CEA No. 53 dropped into
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the core during normal operatior as a result of a voltage spike in the 15 VDC power supp7y.
LER 78-28/3L, dated September 14, 1978; The containment
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particulate radiation monitor was discovered to be reading low (off scale). The cause of f ailure was a bad tube connector and isolator component assembly board.
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LER 78-29/3L, dated September 7, 1978; A failed detector in the halon fire suppression system was discovered.
The thermal detector for Zone 5 in the cablu spreading room was found to be grounded.
LER 78-30/3L, dated September 25, 1978; The CEA motion inhibit
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was found inoperable for CEA position deviation during performance of CEA partial movement testing.
LER 78-31/lT, dated September 8,1978; Boron concentrations in
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21B and 228 safety injection tanks were discovered to be less than Technical Specification requirements.
Reduced concentrations were caused by dilution from in leakage of reactor coolant through the outlet check valves.
LER 78-32/3L, dated September 21, 1978; The level in 21B
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safety injection tank was discovered to be three inches above Technical Specification limits.
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LER 78-33/3L, dated October 20, 1978; While filling the refueling pool from the refueling water tank the Recirculation Actuation System (RAS) was actuated resulting in the trip of both LpSI pumps causing a loss of shutdown cooling.
LER 78-34/3L, dated November 2, 1978; During a surveillance
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test, twelve of the sixteen safety valve relief setpoints were found to be outside the allowed tolerance.
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LER 78-35/lX, dated October 31, 1978; While in cold shutdown,
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both shutdown cooling pumps were lost due to cavitation caused by air entry from the purification system resin blowdown operation.
LER 78-36/lT, dated November 2,1978; No. 21 Main Steam Isolation
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Valve failed to shut during routine testing.
The solenoid actuation valves were found to be sticking.
LER 78-37/3L, dated November 10, 1978; The wide range logarithmic
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neutron flux monitor Channel A was discovered indicating two decades lower than Channels B, C, and D.
The cause was attributed to a bad connector.
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LER 78-38/lT, dated November 3,1973; Contrary to Technical Specification requirements, the containment inner door was open from October 27, 1978 until repairs were affected on October 31, 1978.
LER 78-41/3L, dated November 24, 1978; The refueling water tank
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was inadvertently drained below Technical Specification limits due to a local sample valve that was left open after drawing a routine sample.
The inspector's findings regarding licensee event reports were acceptable.
The inspector verified by review of representative records and by interviews with licensee personnel that appropriate corrective action has been taken with regard to the above LER's.
However, the inspector pointed out to the licensee (by reference to specific LER's) that corrective actions and measures to prevent recurrence are not always addressed. The inspector further stated that if cause of event is unknown at the time and later identified, supplemental reports are required to fully describe final resolution of the occurrence.
The licensee concurred with the inspector's comments and stated that all future LER's submitted would be reviewed in more detail to ensure the above areas are adequately described.
No items of noncompliance were identifie.
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5.
Administrative Control af Safety Related Maintenance The inspector reviewed the licensee's administrative procedures for the control of safety related maintenance to verify that the controls required by Technical Specifications and ANSI N18.7 and ANSI N18.1 had been established. These procedures and the above standards and specifications were then used as the acceptance criteria for the review of individual maintenance activities discussed in paragraph 6 below. The following procedures were reviewed.
Quality Assurance Procedure - 14, Plant Maintenance, Revision
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12, September 28, 1978.
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Calvert Cliffs Instruction (CCI) - 200B, Maintenance Requests, August 31, 1977.
CCI - 201A, Calvert Cliffs Nuclear Power Plant Maintenance
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Procedures, November 21, 1977.
CCI - 204A, Functional Test Procedures, March 31, 1977.
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CCI - 207A, Control of Safety Related Spare Parts, November
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15, 1976.
CCI - 1070, Area and System Cleanliness Requirements, July
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1, 1977.
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CCI - 206B, Personnel and Material Accountability, November 10, 1976.
No items of noncompliance were identified.
6.
Review of Safety Related Maintenance Activities a.
The inspector reviewed documentation associated with a sampling of maintenance activities to verify the following items.
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Applicable limiting conditions for operation were met while components or systems were out of service.
Required administrative approvals were obtained prior
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to commencing wor.
Approved procedures appropriate to the task were used.
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Activities were properly supervised and inspected.
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Functional testing / calibration was performed where required, prior to returning the system to service.
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Maintenance was accomplished by qualified maintenance personnel.
For those items denoted by (*), that QC purchase order
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and receipt inspection documentation was available and complete, b.
The following maintenance activities, including associated maintenance requests, procedures, and functional tests, were reviewed.
Unit 1 M. R. 0-78-313, Shutdown Cooling Heat Exchanger Leakage.
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- M..R. M-78-358, No. 4 Reactor Coolant Pump Seal Replace-
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ment.
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M. R. M-78-144, Service Water Relief Valve to No. 14 Containment Cooler 1RV-1593-Test.
M. R. 0-78-1677, No.12 Control Room A/C Unit Inoperable.
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M. R. M-78-79, No 11 Auxiliary Feed Water Pump Overspeed
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Test.
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M. R. 0-78-420, I-ST-148 Boric Acid Check Valve Leaking.
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M. R. M-78-397, Access Hatch Building, Sprinkler Head Replacement.
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M. R. 0-78-1072, Diesel Fire Pumo Gland Leakage - Idle Pump.
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M. R. 0-77-762, Fire Protection FP-139 Valve Replacemen.
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M. R. 0-78-620, No. 11 Shutdown Cooling Heat Exchanger
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Leaks.
M. R. 0-78-2132, No. 12 Diesel Generator Oil Leak.
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M. R. 0-78-765, No.11 Control Room A/C - Receiver Level Low.
M. R. M-78-221, Auxiliary Building Exhaust Fan No.11 -
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Filter Change.
M. R. 0-78-2349, No.11, No.12, No.13 Component Cooling
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Pumps - Excessive Leakage.
M. R. M-78-223, Spent Fuel Pool Exhaust Fan No.12 Filter
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Replacement.
M. R. IC-78-033, 1-CV-5464 Valve Leaking Excessively.
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M. R. 0-78-1562, Control Element Assembly No. 22 -
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Withdrawal Inoperative.
M. R. E-78-069, Reactor Trip Switchgear U-1, Anti-Pump
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Feature Failure.
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- M. R. M-78-3024, Pressurizer Vapor Sample Valves -
Replacement.
M. R. M-78-235, Replace Component Cooling Heat Exchanger
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Outlet Cover.
M. R. M-78-4058, Containment Tendon Stressing Washer
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Damaged.
M. R. M-78-188, Repair Interlock on Containment Door.
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M. R. IC-78-078, Refueling Water Tank Level Calibratio s
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Unit 2 M. R. M-78-2466, Hose Station 45-22 Inoperable.
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- M. R. M-78-2284, No. 22 Charging Pump - Replace
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Suction and Discharge Valves.
M. R. 0-78-1243, No. 22 Main Steam Isolation Valve
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(MSIV) - Inoperative 2SV4048.
M. R. M-77-2296, No. 21 MSIV - Disassemble Clean and
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Inspect 2SV4043.
M. R. 0-78-3217, No. 22 MSIV - Readjust Pilot Pressure.
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M. R. 0-78-3218, No. 22 MSIV - Limit Switch Adjustment.
M. R. M-77-2297, No. 21 MSIV - Disassemble Clean and
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Inspect 2SV4044.
M. R. M-77-2295, No. 21 MSIV - Disassemble Clean and
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Inspect 2SV4042.
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M. R. M-78-2128, No. 21, No. 22 HPSI Pumps - Oil Replacement.
M. R. 2-IC-78-189, Boric Acid Pumps' Discharge and Suction
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Pressure Gage Calibrations.
M. R. 0-78-3870, No. 21 Boric Acid Pump Discharge Gage
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High - Calibration Check performed.
M. R. 0-78-3891, No. 21 Boric Acid Pump Discharge Gage -
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replaced.
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M. R. 0-78-1077, No. 22 Boric Acid Pump Discharge Pressure Gage Leaks - corrected itself, no recurrent problem.
c.
During the review of Maintenance Requests (MR) it was determined that certain M. R.'s:1) did not include or reference instructions for equipment return to normal operating status, and/or 2) did not contain documentation of equipment functional acceptability for the operations departmen.
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The following M. R. 's did not contain documentation of functional acceptability, 'and no record substantiating the completion of the operability test could be located: M. R. 0-78-765 (Maintenance Foreman's instructions required a one hour trial run prior to return of equipment to service), and M. R. M-78-2128 (did not include or reference instructions for equipment return to normal operating status).
The following M. R.'s did not include or reference instructions for returning equipment to normal operating status, and did not contain documentation of equipment functional acceptability.
However, test acceptability was substantiated through review of completed surveillance tests: 0-78-3217, 0-78-3218, M-77-2295, M-77-2296, M-77-2297, M-78-4058, 0-78-620, and M-78-4058.
The following M. R.'s did not include or reference instructions for equipment return to normal operating status; however, functional acceptability was documented as required: 0-78-2132, 0-78-2349, and M-78-188.
The above conditions represent an apparent noncompliance (deficiency) with T.S. 6.8.1; ANSI N18.7-1972, section 5.3.5; CCI-200B, August 1977 (317/78-38-01 and 318/78-34-01).
During discussions, the licensee stated that the above problem was considered documentary in nature, and that post maintenance testing was accomplished concerning the M. R. 's in question.
Based on conversations with Maintenance and Operations personnel, the inspector noted that there appears to exist, in some cases, a difference of opinion between the two departments as to whether a post maintenance test is required, and if so, the type of test required depending on the kind of maintenance performed.
The licensee acknowledged the inspector's comment.
d.
Associated with the review of M. R.'s, reference was made to various logs in an effort to verify that post maintenance testing was accomplished satisfactorily. Through this process it was determined that the Unit 1 control room log, dated May 11, 1978 through November 6, 1978, could not be located.
This item is unresolved pending completion of actions by the licensee to locate the 109. book and subsequent NRC:RI review (317/78-34-02).
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7.
Administrative Controls for Calibration Administrative controls were reviewed to determine the licensee's program for implementing requirements associated with the control of safety related calibrations, as specified in Technical Specifi-cations, Section 6, Regulatory Guide 1.33, Quality Assurance Require-ments; and, ANSI N18.7, Administrative Control for Nuclear Power Plants.
The following documents were reviewed:
CCI 104B, Change 2, November 20, 1978, Surveillance Test
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Program; CCI 120A, April 13, 1977, Calibration Program for Measuring
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and Test Equipment;
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CCI 135, December 1,1977, Administration of Inservice Inspection;
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CCI 204A, Change 1, May 11, 1977, Functional Test Procedures; CCI 205C, August 23, 1978, Setpoint Control Procedure;
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CCI 209A, March 28, 1977, Test Equipment Calibration Procedures; CCI 211B, September 12, 1978, Preventive Maintenance Program;
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and, CCI 215, August 21, 1978, Vibration Preventive Maintenance.
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No items of noncompliance were identified.
8.
Surveillance Calibration of Safety Related Components and Eouioment Required by Technical Specifications a.
Calibration procedures / data were reviewed on a sampling basis to verify the following:
Calibration frequency requirements have been met;
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Applicable system status during component calibration was
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in conformance with Technical Specification limiting conditions for operations; Procedure format provided detailed stepwise instructions;
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Procedure review and approval were as required by Techni-
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cal Specifications; Trip points of calibrated components were in conformance
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with Technical Specification requirements; and, Technical content of procedures was sufficient to result
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in satisfactory component calibration.
b.
Selected Technical Specification (TS) surveillance requirements, associated test procedures and data (indicated by date of per-formance)arelistedbelow.
Unit 1 TS 4.4.6.1: M-536-1, Revision 1, January 26, 1978, Containment
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Sump Level Calibration; Data; February 8,1978.
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TS Table 4.3-6 Item 3: M-530-1, Revision 1, January 24, 1978, RCS Cold Leg Temperature Calibration, Data: February 10, 1978.
Unit 2 TS Table 4.3-2 Item 1.C and TS Table 4.3-1 Item 4 and 9:
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M-510-2, Revision 3, September 6,1978, Reactor Protective System Calibration Check Section V, Pressurizer Pressure TM/LP Calibration Check (includes safety injection function);
Data: September 20,1978, and M-210 B-2, Revision ll, March 30, 1978, Reactor Protective System Functional Test,Section V, Pressurizer Pressure TM/LP Functional Check; Data: November 22, October 19, September 5, and August 2, 1978.
TS Table 4.3-1 Item 5 and TS Table 4.3-2 Item 1.B: M-520-
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2, Revision 0, August 4,1976, ESFAS Calibration,Section II, Automatic Removal of Pressurizer Pressure and Steam Generator Pressure Blocking Signal Verification Test and Section III, Containment High Transmitter Calibration Check (SIAS); Data: January 6, 1978, and M-220-2, Revision 3, June 30, 1977, ESFAS Functional Test (Section II and III only); Data: November 22, October 19, September 5, and August 2, 197.
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TS Table 4.3-6 Item 5; M-532-2, Revision 2, June 28,
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1978, Pressurizer Level 2C43 Calibration Check; Data:
September 21, 1978.
TS Table 4.3-10 Item 10: M-535-2, Revision 1, January 24,
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1978, Feedwater Flow Calibration; Data: July 16, 1978.
TS 4.4.9.3.1.a and b: M-572-B-2, Revision 0, September 6,
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1978, Pressurizer Relief Valve (FRV) Channel Calibration; Data: September 16,1978; and M-672-B-2, Revision 0, September 6,1978, Pressurizer Relief Valve Functional Test; Data: October 17, 1978.
No items of noncompliance were identified.
9.
Calibration Required by Technical Specifications of Components and Equipment Associated with Safety Related Systems and/or Functions a.
The calibration program (addressed in Regulatory Guide 1.33 and ANSI N18.7) for components associated with safety related systems was reviewed on a sampling basis.
These components are used to monitor system parameters to comply with Technical Specification (TS) safety limits; limiting conditions of operation and surveillance requirements.
The following were verified:
Specific requirements have been established for the above
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calibrations including schedules and frequencies; Procedures have been reviewed and approved in accordance
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with the Technical Specifications, contain acceptance criteria consistent with the Technical Specifications, and contain detailed instructions commensurate with the complexity of the calibration; and, Technical content of procedures are adequate to perform a
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satisfactory calibration.
b.
The selected TS parameter, associated instrument calibration procedures (where appropriate) and data (indicated by date of performance)arelistedbelow.
Unit 1 Refueling Water Tank Level (TS 3.1.2.7 and TS 3.5.4), Pre-
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ventive Maintenance (PM) 1-52-I-A-17, Functional Test Instruction (FTI) 105, Revision 1, May 1,1973, Loop Calibration Check Procedure; Data: Channels 4142 and 4143 on March 25, 197.
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Cable Spreading Room Halon System Pressure (TS 4.7.11.3),
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No data reviewed (see below).
Fire Protection High Pressure Pump Automatic Start Pressure
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(0-PS-6220 E and 0-PS 6224) and Fire Main Header Pressure (0-PR-6224) (TS 4.7.11.1.1), PM l-13-I-A-5 and 6, FTI-101, Revision 0, March 30, 1973, Alarm Setpoint Procedure and FTI 105 (title addressed above), Data: 0-PS-6220 and 6224 on October 18, 1978 and 0-PR-6224 on October 18, 1978.
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Fire Protection System Flow (0-FI-6223) (TS 4.7.ll.l.l.f.2 PM-1-13-I-A-9, FTI-106, Revision 1, August 20, 1974, Receiver Calibration / Calibration Check; Data: October 19, and December 7,1978.
Discharge and Suction Pressure Gages for High Pressure
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Safety Injection Pumps (HPSI) (TS 3.5.3 and 4.0.5) PM 1-52-I-A-18 through 20, FTI-150 (title previously addressed);
Data: Discharge Pressure Gages May 18, 1978; Suction Pressure Gage for No.11 HPSI Pump on July 7,1978; Suction Pressure Gages for Nos.12 and 13 HPSI Pumps on August 24, 1978.
Unit 2 Pressurizer Temperature and Spray Line Differential
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Temperature (TS 3.4.9.2), PM 2-64-I-R-1, No data review (seebelow).
Diesel Fuel Oil Level (TS 3.8.1.1), FM l-23-I-2YR-1,
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May 31, 1978, Fuel Oil Storage Tank Level (one point check of mechanical indicator per PM schedule for last refueling, September to November,1978).
Boric Acid Storage Tank Volume (TS 3.1.2.7), PM 2-40-
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I-A-8 and 9, FTI 105 (title previously addressed),
Data: Channel 206 on August 10, 1977 and August 1, 1978 and Channel 208 on February 8, 1978 and August 1, 1978.
Discharge and Suction Pressure Gages for Boric Acid (BA)
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Pumps (TS 4.1.2.6), PM 2-40-I-A-4 and 5 (for discharge gages only - see below), FTI-106 (title previously addressed);
Data: No. 21 BA Pump Suction Gage on September 1,1976 and December 14, 1978; No. 21 BA Pump Discharge Gage on August 13,1976; No. 22 BA Pump Suction Gage on September 1,1976 and December 14,1978; No. 22 BA Pump Discharge Gage on August 16, 1976 and December 14, 197.
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c.
During the review it was determined that the Preventive Main-tenance (PM's) Card and Functional Test Instructions (FTI's)
constitute the procedure for control of the calibrations of these type instruments. However, these two documents collectively lack certain procedural aspects for the calibration of in-line instruments / gauges.
Specifically, the PM card and the FTI do not provide for on-shift operations personnel formal release of the equipment for preventive maintenance and for the documen-tation of functional acceptability by the on-shift operational personnel.
In some instances, it was noted that a maintenance request (MR) number (which does provide for the operation department interface as discussed above) was documented on the calibration data sheet or the MR was attached to th data sheet. However, this was done primarily in cases for corrective maintenance and not for routine calibrations.
The licensee representative acknowledged the above and stated that this issue had been addressed previously and it appeared to be resolved through.the use of PM weekly schedules.
However, the inspector stated that these schedules do not appear to schedule PM's on a shift basis.
It was noted that administrative
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controls for the PM system require the permission'of the shift supervisor prior to the performance of a PM but these controls do not provide for documentation of this release of equipment.
The licensee representative stated that this area warranted further review. This is unresolved pending review by the licensee and subsequent NRC:RI inspection in this area (317/78-38-03 and 318/78-34-02),
d.
It was determined that certain instruments could not be cali-brated due to physical restrictions.
Specifically, for Pressurizer Temperature, tap-in points for the loop calibra-tions are physically inaccessible.
This is being reviewed to determine the feasibility of establishing a support structure in the area or to use other instrumentation, that is accessible for calibration, to comply with the TS limit on Pressurizer Temperature heatup rate and Spray Line Differ-ential Temperatur.
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Also, the pressure gages for the Cable Spreading Room Halon System are unisolable from the respective storage bottles.
The licensee representative stated that this system is under review by the Office of Nuclear Reactor Regulation (NRR)
and that modifications may be warranted in which consideration will be given to isolable pressure gages.
In the interim, the Fire Inspector has been notified to provide bottle fittings with isolable pressure gages when these bottles are sent off site for recharging through vendor services.
This area is unresolved pending the calibrations of instruments used to comply with the Limiting Conditions of Operations in TS for the Halon System Pressure and Pressurizer Temperature, and subsequent NRC:RI review (317/78-38-04 and 318/78-34-03).
e.
The suction gages (not ambered) for the Boric Acid Pumps do not have corresponding PM cards for periodic calibrations.
However, these gages are temporary and have been calibrated in tie past on a periodic basis.
The licensee representative stated that all gages associated with Inservice Inspection (ISI) Surveillance Procedures are being reviewed under a Facility Change Request and that a result of this review will be to assure PM's exist for periodic calibrations of these gages.
This is unresolved pending the incorporation of all ISI gages into the PM system and subsequent review by NRC:RI (317/78-38-05 and 318/78-34-04).
f.
During the review of calibration data it was noted that several records for these calibrations were stamped with 1 year retention periods. This was observed for calibration records gages /!r.stru-ments in the Fire Protection System and Boric Acid System.
The Technical Specifications, Section 6, requires a 5 year retention for such records.
The licensee representative acknowledged the above and stated that the present Quality Assurance System for maintaining plant records is keeping all documents on a permanent basis.
It was indicated that this may not be true in the future.
The inspector did not notice any instance of missing calibration records since the establishment of this calibration program.
The inspector stated this area will be reviewed in a subsequent inspection (317/78-38-06 and 318/78-34-05).
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10.
Calibration and Control of Test Equipment a.
The calibration and control of test equipment used as standards in the calibration of components identified in paragraphs 8 and 9 above were reviewed on a sampling basis to verify the following:
Establishment and adherence to calibration schedules;
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Maintenance of calibration records identifying standards
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used which have traceability to the National Bcreau of Standards or other independent testing organizations; Proper storage and labeling of test equipment; and,
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Adequate control of test equipment including recordkeeping.
b.
The selected devices are listed below:
Fluke 8600 A Digital Volt Meter (DVM), Serial Number
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(S/N) 8182; Weston Volt Ohm Milliameter (V0M), S/N 7466;
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Keith Model 174, S/N 8567, (primary standard);
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Wallace and Tiernan Pneumatic Tester, S/N 14435;
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Perma Cal Gauge, S/N 5016;
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Heise Gage, S/N 17479;
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Mansfield and Creen Modified Deadweight Tester, S/N
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7101 (primary standard);
Fluke 8600 A DVM, S/N 8180;
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Fluke 8600 A, DVM, S/N 8184; and,
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Fluke 887 A Tester, S/N 8533, primary standard.
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No items of noncompliance were identifie.
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11.
Performance of Calibration / Functional Tests The inspector witnessed the performance of Unit 1, M-210-B-2, Revision 7. September 6,1978, Reactor Protection System Functional Test,Section II (Power Range) on December 14, 1978.
This was to verify performance consistent with approved procedure and to verify the implementation aspects for the control of test equipment.
No items of noncompliance were identified.
12. Technician Qualification The inspector reviewed the qualification records of selected personnel having responsibility for calibration and maintenance of safety related systems and components to verify that the individuals experience level and training were in accordance with ANSI N18.1, Selection and Training of Nuclear Power Plant Personnel, Section 4.
During this review, it was determined that one individual assigned to the Instrument and Control (I&C) Group (employed by the licensee August, 1978) lacks the required 2 years working experience in the I&C specialty. This person was qualified by the licensee to per-form certain Technical Specification Surveillance Procedures at a Level I capability in accordance with ANSI 4.5.2.6, Draft 3, October,1974, Qualification of Inspection, Examination, and Testing Personnel for the Construction Phase of Nuclear Power Plants.
ANSI 4.5.2.6 provides for optional experience requirements other than the 2 years working experience.
Based on a review of the individual's background records, it appeared to the inspector that the Level I requirements of ANSI 4.5.2.6 were met for this employee.
However, the inspector emphasized that Technical Specifications reference ANSI 18.1-1971 for the qualifications of onsite organization personnel and the Operational Quality Assurance Program (FSAR Section 1.c) commits to both ANSI Standards.
Further, ANSI 4.5.2.6 sddresses primarily the quality assurance experience levels for quality assurance personnel and it is, therefore, inappropriate to use ANSI 4.5.2.6 to qualify onsite organization personnel not in the quality assurance area. This position is consistent with previous guidance from NRR on the same issue.
Based on a review of selected records of calibratic safety e
related instruments and on discussions with licenses asentatives, it appeared the subject individual has not performed 1 ourveillance Procedures above, even those for which the person is qualified as documented on Certificate of Qualificatio.
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The licensee representative acknowledged the above and stated that this area will be reviewed.
It was also stated that until resolution of this issue, the subject individual will not be permitted to per-form any TS Surveillance Procedures or other safety related functions in other than a training capacity.
Further, it was stated that upon completion of the above review appropriate programmatic changes will be made.
This is unresolved pending completion of action by the licensee as stated above and subsequent NRC:RI review (317/78-38-07 and 318/78-34-06).
13.
Personnel Identification and Dosimetry On December 13, 1978, during discussions with personnel, the inspector observed that one individual (A) was wearing another individual's (B) identification badge.
Both individuals are members of the onsite organization with the same protected and vital area access eligibility.
However, along with the badge, individual A was wearing individual B's personnel dosimetry (Thermoluminescent Dosimeter - TLD).
The wrong badge /TLD had been issued approximately 7:00 a.m. that day and individual A returned to the guard house at approximately 11:00 a.m.
for the proper badge /TLD.
Subsequently, the licensee conducted an analysis of the event with the following conclusions.
No other employees were involved in the mix-up.
Individual B
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had not reported to the site that day; but had he reported, the error probably would have been caught.
Individual A had not gone into any radiation areas and, therefore,
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there would be no reason to suspect erroneous dosimetry readings on either individual A's or B's TLD.
The cause appears to be an error on the part of the guard who
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issued the wrong badge.
Both badges were in close proximity to each other in the file rack.
The licensee representative stated that security procedures are being revised in this area to preclude recurrence of this event.
However, no definite plan could be presented to the inspector at the time of the inspection since review by both the onsite organiz-ation and security organization is warranted.
This is unresolved pending completion of action by the licensee as stated above and subsequent NRC:RI review (317/78-38-08 and 318/78-34-07).
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14.
Facility Tour At various times during the period of September 12-15, 1978, the inspector conducted tours of the following accessible plant areas:
Auxiliary Building
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Turbine Building
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Control Room
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Outside Peripheral Areas
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New Security Building
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During the inspection tours, observations relative to the following items were acceptable:
plant housekeeping and cleanliness;
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posting of radiation areas, access control, and currency of
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survey; security access controls;
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control room manning and assignment of licensed operators
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pursuant to the Technical Specifications; and,
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selected monitoring instrumentation - operability and para-
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meters within license limits.
No items of noncompliance were identified.
15.
Unresolved Items Unresolved items are findings about which more information is required in order to ascertain whether they are acceptable items, items of noncom-pliance or deviations.
Unresolved items disclosed during the inspec-tion are discussed in paragraphs 6, 9c, 9d, 9e, 9f, 12, and 1.
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16.
Exit Interview The inspectors met with licensee representatives (denoted in para-graph 1) at the conclusion of the inspection on December 15, 1978.
The inspector summarized the scope and findings of the inspection.
Subsequent discussions of the inspectors' findings occurred in telephone conversations between licensee and NRC:RI on December 15, 19, and 20, 1978.
17. Management Meeting a.
Purpose On January 4,1979, a management meeting was held with Baltimore Gas and Electric Company representatives at the NRC Region I Office to discuss the Office of Inspection and Enforcement Inspection program at the Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
b.
Attendees The following listed persons were present at the exit interview:
Baltimore Gas and Electric Company A. E. Lundvall, Jr., Vice President, Supply V. R. Evans, Vice President, General Services J. N. Bullock, Manager, Electrical Production C. C. Steinhagen, Manager, General Supervisor Security Service L. B. Russell, Chief Engineer, Calvert Cliffs Nuclear Power Plant J. G. Deegan, Supervisor, Security Programs R. E. Denton, Nuclear Plant Engineer - Operations, Calvert Cliffs Nuclear Power Plant NRC Office of Inspection and Enforcement, Region I J. M. Allan, Deputy Director G. H. Smith, Chief, Fuel Facility and Materials Safety Branch J. W. Devlin, Chief, Security and Investigation Section, Safeguards Branch R. R. Keimig, Chief, Reactor Projects Section No.1, Reactor Operations and Nuclear Support Branch D. F. Johnson, Project Inspector, Reactor Operations and Nuclear Support Branch
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c.
Acenda The following areas were addressed by RI and discussed with the licensee's representatives during the course of the meeting:
1.
Purpose of Meeting 2.
Emergency Planning, Environmental Monitoring Program and Health Physics - inspection findings for 1978 and enforcement history 3.
Security - Recent inspection findings and enforcement history 4.
Reactor Operations - Recent inspection findings in the area of apparent personnel errors and licensee corrective action to precl'ude recurrence 5.
Closing Remarks d.
Conclusion It was mutually agreed by both parties that the meeting was beneficial and that it provided a clear understanding of the Office of Inspection and Enforcement Inspection program and the licensee's actions to improve the effectiveness of the management control system to enhance compliance with NRC regulations.