IR 05000317/1979023
| ML19323D850 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 03/10/1980 |
| From: | Architzel R, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19323D836 | List: |
| References | |
| 50-317-79-23, 50-318-79-22, NUDOCS 8005220434 | |
| Download: ML19323D850 (9) | |
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O U. S. NUCLEAR REGULATORY COMMISSION To
OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
50-317/79-23 Report No.
50-318/79-22 50-317 Docket No.
50-318 DPR-53 C
License No.
DPR-69 Priority
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Category C
Licensee:
Baltimore Gas and Electric Company P.O. Box 1475 Baltimore, Maryland 21203 Facility Name:
Calvert Cliffs Nuclear Power Station, Units 1 and 2 Inspection At:
Lusby, Marvland Inspection Conducted:
December 26-28, 1979; January 2-4, 23-25, and 28-31, 1980 Inspectors:
S O. D. ) A, b
'31 o I8o R. Architzel, Residbnt Reactor Inspector date
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date date Approved by:
80bkb 3flol80 E. C. McCabe, Jr., Chief, Reactor Projects date Section No. 2, RO&NS Branch Inspection Summary:
Inspection on December 26-28, 1979, January 2-4, 23-25 and 28-31, 1980 (Combined Report Nos. 50-317/79-23 and 50-318/79-22)
Areas Inspected:
Routine, unannounced inspection by the newly assigned resident inspector (29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> Unit 1, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Unit 2) of plant operations and the power
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distribution episode (Unit 1) and review of reports.
The NRC Resident Office was
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officially activated on January 31, 1980.
Noncompliances:
None in two areas; two in one area (Infraction, not investiga-ting cause of Unit 1 PORV actuation (paragraph 4.d.); Infraction, Failure to correct grounds on vital DC Bus 21 (paragraph 4.e.)).
Region I Form 12 (Rev. April 1977)
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DETAILS 1.
Persons Contacted The following technical and supervisory level personnel were contacted:
E. R. Bauer, Assistant General Foreman - Maintenance J. A. Crunkleton, Foreman - Electrical Maintenance
- R. E. Denton, Nuclear Plant Engineer - Operations W. J. Lippold, Nuclear Engineer J. A. Mihalcik, Fuel Management Engineer P. G. Rizzo, Assistant General Foreman - Maintenance L. B. Russell, Chief Engineer - Nuclear Plant
- Present at the exit interview.
Other licensee employees were also interviewed.
2.
Piping and Instrumentation Drawings (P& ids)
As a result of the TMI accident, other incidents and excercises conducted in the NRC Operations Center, it became apparent that certain Nuclear
Power Plant P& ids are essential for the NRC Operations Center in Bethesda, Maryland.
The licensee was requested to provide the center a current set i
of P&ID's.
The licensee has provided the Operations Center with a complete set of Operations P&ID's (these deal with major plant systems) and has established control procedures to provide revisions in a controlled fashion.
No unacceptable conditions were identified.
3.
Unit 1 Power Distribution Anomaly A power distribution anomaly developed in the Unit 1 core in late September 1979.
The condition was discovered on October 22, 1979 and has been the subject of several telephone conversations, reports, and meetings between the NRC staff and BG&E.
The proximate cause of the episode was determined by the licensee to be a crud deposition on the fuel cladding resulting in Axial Shape Indices (ASI) and core differential pressures being different then expected.
A possible cause of the crud formation was determined to be instrument air leakage into the Chemical and Volume Control System (CVCS) during resin transfer operations.
A temporary fix to this air intrucion was to proce-durally isolate and vent the CVCS Ion Exchangers any time the applicable instrument air header was pressurized.
A planned permanent solution includes supplying this system with nitrogen instead of compressed air.
The inspector reviewed the notes of the Power Distribution Task Force, created by the Plant Operations Safety Review Committee to evaluate and direct licensee efforts to analyze and correct the anomaly.
These notes included a chronological history of the episode and minutes of the Task Force Meetings.
Principal Guidance and Engineering Support / Analyses /Recom-mendations were provided by the NSSS supplier.
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The inspector also reviewed the Safety Evaluation performed pursuant to 10 CFR 50.59 for Facility Change Request No. 79-3001, Hydrogen Peroxide Treatment of the Unit 1 Reactor Coolant System.
The inspector attended a meeting held on January 22, 1980 in Bethesda, Maryland between various members of the staff, and representatives of Baltimore Gas and Electric, Combustion Engineering, and Brookhaven National Laboratory.
The minutes of this meeting are in the Public Document Room.
The inspector reviewed the implementing procedure for the H 0 treatment,
RCP 1-1206, Special Procedure for RCS H 0 treatment, dated January 26, 1980.
The procedure contained verifica$ibn of an additional 1% (total of 2%) shutdown margin prior to initiation and provisions to ensure crud samples were mass spectrum / isotopically analyzed.
These items were requested to be verified by the licensee following the January 22, 1980 meeting.
The licensee performed the H 0, decrudding treatment coincident with a
Unit 1 shutdown in late January to implement THI-2 Short Term Lessons learned modifications.
A complete report is to be provided by the licens-ee concerning the Power Distribution Episode, documenting the event and the corrective actions taken.
This report will be reviewed during future NRC inspections (317/79-23-03).
4.
Review of Plant Operations a.
Plant Tour At various times during the inspection the inspector made tours of the facility.
These included the Control Room, Auxiliary Building (all levels, no High Radiation Areas) Turbine Building, Outside Peri-pheral Area, Security Buildings, Health Physics Control Points, Diesel l
Generator Rooms and Intake Structure.
The following observation and determinations were made:
Radiation controls established by the licensee, including posting
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of radiation areas, conditions of step-off pads and disposal of protective clothing were observed.
Control room manning was observed on several occasions during
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the inspection.
Systems and equipment in all areas toured were observed for the
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existence of fluid leaks and abnormal piping vibrations.
Seismic restraints and hydraulic snubbers were examined ca a
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l sampling basis to verify adequate installation and fluid levels.
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Plant housekeeping conditions, including general cleanliness
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conditions and storage of materials and components to preclede i
safety and fire hazards, were observed.
Control room and local monitoring instrumentation for various
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components and parameters were observed, including reactor power level, CEA positions and safety related valve position indication.
Whether proper access controls established.
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No items of noncompliance were identified.
b.
Review of Operating Logs, Records A review of logs and records was made to identify significant changes and trends, to assure required entries were being made, to verify Operating Orders conform to the Technical Specifications, to verify proper identification of abnormal conditions, and to verify confor-mance to reporting requirercents and Limiting Conditions for Operation.
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The following records were reviewed for the report period:
Shift Supervisors Log
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Unit 1 Control Room Operators Log (sample review)
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Unit 2 Control Room Operators Log (sample review)
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Nuclear Plant Engineer - Operations Notes and Instructions
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Unit 1 and 2's Control Room Daily Operating Logs (sampling review)
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NSSS Computer Log (sampling review)
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During a review of these logs at approximately 9:00 p.m. on January 29, 1980 the inspector noted that at approximately 11:45 a.m. on January 29, 1980 Power Operated Relief Valve (PORV) RC-404-ERV on linit 1 had inadvertently opened.
About 11:59 a.m. the 12B Reactor Coolant Pump started by itself.
Further review of these events identified the
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following subparagraphs.
c.
Reportability The inspector informed the licensee that the inadvertent opening of a PORV is an abnormal degradation of the Reactor Coolant Pressure
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Boundary (T.S. 6.9.1.8.c).
Relief valves are part of the RCS Bound-ary as defined by 10 CFR 50.2(v)(2)(iii).
A Licensee Event Report was prepared and submitted.
This item was discussed with the licens-ee prior to the expiration of the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed by the T.S.
The licensee's implementation of reporting requirements is unresolved and will be reexamined in future NRC Inspections (317/79-23-01; 318/79-22-01).
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PORV Actuation The inspector questioned the licensee concerning the cause of the
i inadvertent PORV actuation.
During cold shutdown conditions the PORV functions as MPT (Minimum Pressurization Temperature) relief with dual setpoints of 430 psig RCS pressure and 230*F RCS temperature.
On January 29, 1980, the reactor was at 258 psig and 130*F when the PORV lifted and reduced pressure to 169 psig.
The transient was
terminated by operator action (momentary use of the override shut feature of the PORV).
Reactor Coolent Pumps were secured when the RCS pressure dropped below their minimum operating values.
RCS flow was maintained at all times (a flush was in progress using 6000 gpm through the Shutdown Cooling System) and pressure was restored using all the Pressurizer Heaters.
No maintenance request was initiated
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to investigate this " spurious" event until the licensee was prodded by the Inspector and the necessity to " fill the blocks in" the LER form became evident on January 30, 1980.
During the processing of the Maintenance Request it was concluded that a technician removing pressure transmitter covers at the time and had in fact removed the initiating device's (PT-103-1) cover right before lunch on January 29, 1980, and had bumped the transmitter, causing the PORV to lift.
The technician had been instructed to be careful while removing the covers and was obtaining serial numbers of transmitters pursuant to NRC Bulletin 79-01, Qualification of Electrical Components.
Failure to initiate in a timely fashion, a determination of the cause of j
this loss of the primary coolant pressure boundary is an item of j
noncompliance (50-317/79-23-04).
e.
Reactor Coolant Pump Spontaneous Start About 11:45 a.m., January 29, 1980, during the pressure reduction following the inadvertent PORV actuation, the Unit 1 operator secured all Reactor Coolant Pumps in Unit I to assure Net Positive Suction Head requirements were not exceeded for the pumps.
At approximately 11:59 a.m. the 128 (Unit 1) Reactor Coolant Pump (RCP) started with-out operator action.
The operator immediately utilized the manual control switch " Pull to Lock" override feature to stop RCP 12B.
Maintenance action was initiated to investigate the cause of the inadvertent start and to assess potential damage to the pump and/or its motor.
The motor requires a separate lubricating oil lift pump i
to be running prior to starting to allow for proper clearance and l
bearing lubrication.
The control circuit design requires proper
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discharge pressure of the oil lift pump, inserting of a " synchronizing
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stick" into the control panel, and operation of the manual control switch to effect RCP startup.
During troubleshooting the licensee started the oil lift pump and then removed the control switch from
" Pull to Lock".
The pump s arted again, with the control switch in
"0FF" and the synchronizing stick not inserted.
Bearing temperatures
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and pusap vibrations were examined during the running of RCP 128 and the licensee concluded that no physical damage to the pump or motor had occurred.
(Since the RCP had been secured only about 10 minutes
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when it started spontaneously, it is possible that a film of oil was still coating the bearing surfaces.)
Examination of the control circuit revealed that the oil lift pump pressure switch casing was physically damaged and bent, causing a ground at the time of the spontaneous RCP 128 start, there were workers in the containment near the switch casing, and a possible cause of the switch ground was damage at that time incident to the work being performed.
The ground in the control circuit, which is powered from the No. 21, 125 VDC Battery Bus, coupled with another, uncorrected ground on the positive side of the bus, provided a cur-
rent path which bypassed the normal control circuit.
A simplified
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schematic of this spontaneous equipment operation event is appended to this report as Figure 1.
Investigation of the existing ground by the inspector revealed that Maintenance Request (HR) E-79-209 was initiated on November 5, 1979.
Investigation work was performed by electricians on November 5, 1979, without the ground being cleared.
On January 17, 1980, prior to the spontaneous start, further work was done on MR-E-79-209, and
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two partial grounds were cleared without the zero ground being found and cleared.
On January 29, 30 and 31,1980 the zero ground, caused by metal to metal contact, was located in Unit 1 annunciator circuit H-46 (served by the 1K02 feeder breaker).
The actual device grounded was Shutdown Cooling Differential Pressure Switch 1-PDIS-3825.
After clearing the zero ground, a partial ground was found on the negative side of the battery bus and cleared.
The inspector expressed concern that the ground had been allowed to exist so long without correction.
One of the bases for the assump-
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tion of validity of the single failure criteria is that the vital busses are ungrounded. The FSAR states in Chapter 8 that "...The 125 volt d.c. system and the 120 volt a.c. system are ungrounded and equipped with ground detectors.
Each of the four 125 vdc emergency power sources is equipped with the following instrumentation to
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enable continual operator assessment of emergency power source
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condition... DC ground indication..."
The inspector questioned the licensee concerning the severity of the
ground and its nature.
Various answers were received including a description as continuously present to intermittent.
The licensee also stated that complete troubleshooting for the ground would require that both nuclear powerplants be shutdown.
The inspector asked what criteria govern vital bus ground acceptability.
No criteria had been established by the' licensee.
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Ungrounded circuits have the advantage that development of one ground does not cause a loss of equipment.
A second ground is required to cause equipment to be tripped or shorted out.
Grounded circuits cannot tolerate such an unintentional ground without equipment loss.
But, an ungrounded circuit on which grounds are allowed to remain is susceptible not only to the loss of equipment from development of a second unintentional ground, but also to unplanned and uncontrolled equipment operation because of just such a circumstance as occurred in this case.
The following factors were identified as having contributed to this occurrence.
(1) Ground readings are not required to be taken at regular inter-vals (such as once per shift), documented, or reviewed by station management.
(2) Minimum acceptable ground readings are not specified.
(3) A ground ccadition was identified but not corrected until after an event preventable by prompt ground identification and cor-rection had occurred.
MR-E-79-209 documented an unsuccessful attempt to clear a ground on November 5-6, 1979 and indicated no further action until January 17, 1980, when two partial grounds were corected but the zero ground remained.
No further action to clear the ground was documented on the MR before RCP 128 started when it should not have done so.
After the zero ground was cleared, an additional partial ground was found on the negative side of the battery bus, making a total of at least 4 ground conditions identified on circuitry identified as ungrounded in the FSAR.
This is an item of noncompliance (50-317/79-23-02; 50-318/79-22-02)
with requirements to have appropriate procedures for an activity affecting quality and for failure to effect prompt corrective action (10 CFR 50, Appendix B, Criteria V and XVI).
5.
Plant Operations and Safety Review Committee The activities of the Plant Operations and Safety Review Committee during calendar year 1979 were reviewed.
Inspection activities in this area were not completed and will be documented in a future NRC Inspection Report.
6.
Review of Reports l
The following licensee reports were reviewed with no unacceptable condi-l tions identified.
November, 1979 Operations Status Reports for Calvert Cliffs No. 1
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Unit and Calvert Cliffs No. 2 Unit, dated December 13, 1979, revised December 19, 1979 (Unit II).
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December, 1979 Operations Status Reports for Calvert Cliffs No. 1
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Unit and Calvert Cliffs No. 2 Unit dated January 15, 1980.
Licensee Event Report 317/80-07/0lT, PORV RC-404-ERV Opening, dated
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February 12, 1980 was reviewed following the inspection.
Circum-stances surrounding this event were reviewed onsite as described in paragraph 4.
Licensee Event Report 317/80-06/03L, Control Element Assembly (CEA)
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60 dropped into the core.
Unit 1 was operating at 60% when CEA 60 dropped into the core at 1755.
Emergency Operating Procedure No. 11, CEA Malfunctions, was initiated and NRC notification was made via the " hot" line to the Bethesda Operations Center.
The Rod was restored to group height within the one hour allowed by T.S. 3.1.3.1.f (actual 50 minutes).
A power reduction was not required as the reactor was initially less than 70%.
The inspector discussea the event with the licensee and examined the current traces for the magnetic jacking devices and timing of the traces.
These traces were taken during the evening of January 3 after the CEA had been restored to group height.
All traces examined were normal and a cause was not determined for the rod drop during excercising or for subsequent slippage when attempts were made to restore the rod.
This item was previously unresolved (317/78-18-02; 318/78-12-02) and remains open pending further analysis by the licensee and NRC review of the licensee's actions.
7.
Unresolved Item An item about which more information is required to determine acceptability considered unresolved.
Paragraph 4.c contains an unresolved item.
8.
Exit Interview At the conclusion of the inspection the inspector held a meeting (see paragraph 1 for attendees) to discuss the inspection scope and findings.
The items of noncompliance and unresolved item were identified.
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FIGURE 1: SIMPLIFIED SCHEMATIC SHOWING REACTOR COOLANT PUMP (RCP)
SPONTANE0US START CIRCUIT DC Rit9 911 ppt Y Manual Control Switch
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Ground (s)
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Elsewhere on DC Synchronizing Equipment Stick
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Ground Flow Path
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011 Lift for
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Pump Current
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Pressure
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Switch \\
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125 V
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Battery
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No. 21
\\ Ground on Switch
\\RCP Start Device
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DC BUS RETURN
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