IR 05000317/1980005
| ML19320D324 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/30/1980 |
| From: | Architzel R, Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19320D318 | List: |
| References | |
| 50-317-80-05, 50-317-80-5, 50-318-80-05, 50-318-80-5, NUDOCS 8007210247 | |
| Download: ML19320D324 (20) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AN') ENFORCEMENT 50-317/80-05 Region I-
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50-318/80-05 Report No.
50-317 Docket No.
50-318 C
OPR-53 C
Category License No. DPR-69 Priority
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Licensee:
Baltimore Gas and Electric Company
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P. O. Box 1475 Baltimore, Maryland - 21203 Facility Name:
Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Inspection at:
Lusby, Maryland Inspection co ducted: April 7-30, 1980 Inspectors:
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C 30 ()ArchitzeT,UResident Reactor Inspector date signed
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date signed date signed Approved by:
dObkh f/T0/60 E. C. McCabe, Jr., Chief, Reactor Projects date signed Section No. 2, RO&NS Branch Inspection Summary:
Inspection on April 7-30, 1980 (Combined Report Nos. 50-317/80-05 i
and 50-318/80-05)
Areas Inspected:,' Routine, onsite regular and backshift inspection by the resident inspector (20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, Unit 1; 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, Unit 2). Areas inspected
included the control room and the accessible portions of the auxiliary, turbine, service, and intake buildings; radiation protection; physical security; fire protection; plant operating records; spent fuel shipment, loss of coolant accident procedure and associated training review, and reporting to the NRC.
Resul ts: No items of noncompliance were identified during this inspection.
- Region I Form 12
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DETAILS 1.
Persons Contacted The following licensee personnel were contacted:
E. R. Bauer, Assistant General Foreman - Maintenance S. Davis, Performance Engineer R. Denton, Nuclear Plant Engineer - Operations W. Gibson, Nuclear Plant Engineer - Performance W. Lippold, Nuclear Engineer M. Miernicki, Engineer, PMD J. Mihalcik, Fuel Management Engineer P. Rizzo, Assistant General Foreman, Maintenance
- L. Russel, Chief Engineer - Nuclear Plant R. Sheranko, General Foreman - Nuclear R. Wenderlich, Performance Engineer J. Yoe, Plant Training Specialist R. Mallick, General Electric Co. Service Representative
- Present at the exit interview.
Other licensee employees were also contacted.
2.
Licensee Action on Previous Inspection Findinc,
(Closed) Unresolved Item (317/79-08-03; 318/79-07-03); Review Disinfection of Respiratory Protection Equipment.
This item concerned the licensee practice of donning respirators, following disinfection, to perform a negative pressure check and subsequent bagging of the respirator.
The licensee now wipes the mask's sealing surface with cetylcide (.a disinfectant) subsequent to the negative pressure test.
(Closed) Noncompliance (317/79-13-01; 318/79-10-01) Inadequate Posting Pursuant to 10 CFR 19.11.
The licensee responded to this item in letters dated November 9, 1979 and January 4, 1980.
The inspector observed the new posting of these notices which now contain references to the Facility licenses and the Operating Procedures. The notices are posted at both guard houses and the turbine building elevator / control room entrance.
Responsibility to annually review these postings has been delegated-to the Plant Radiation Safety and Chemistry Engineer.
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(Closed) Unresolved Item (317/79-14-02; 31'8/79-11-01). Addi tional-Measures to Preclude Misalignment of FCCS Valves. The licensee has incorporated as a Shift Turnover check (Control Room Operator Shift Turnover Sheet,. Step E) an additional measure to ensure that ECCS valves remain correctly positioned.
This check verifies that the ECCS valves are in position by matching handswitch and indication light position to either green (closed) or red (open) dots installed on the Control Boards. The Shift Turnover Checks were implemented on December 28, 1979 via NPE-0 Standing Instruction 79-8.
(Closed) Unresolved Item (318/78-19-07):
Review Retest Results of STP-0-65-2. As noted when this item was reinspected (Report 318/79-01), the retest results for the STP are considered lost.
The licensee's method of tracking and recording retest data on pumps and valves which fail IWP and IWV acceptance criteria has been formalized and incorporated in CCI-104D, Revision 3, Surveillance Test Program.
The inspector reviewed the revised test procedure and noted that Section VII, Conduct of Inservice Testing of Pumps and Valves, now contain the following requirements:
If a pump test reveals a parameter to be within the " Alert
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Range" the pump is tested at 15 day intervals until the cause of the deviation is determined and corrected.
Documentation of the retesting is accomplished on procedure attachment 18, Pump Supplemental Test Sheet.
If a pump or valve is not tested during the performance of its
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applicable STP the Surveillance Test Coordinator - Operations is responsible for insuring the component is tested prior to
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return to service. Performance is documented on procedure attachment 20, Supplemental Test Sheet.
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If a valve stroke time exceeds its alert acceptance criteria
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the valve will be retested at least monthly until the completion of corrective maintenance. A Valve Supplemental Test Sheet, procedure attachment 19, documents the results.
If a valve exceeds its action acceptance criteria it is declared inoperable.
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The Surveillance Test Coordinator - Operations is required to
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maintain a Supplemental Test Sheet Log, depicting the current status of all pumps'and valves which have failed their respective Inservice Tests.
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The inspector had no further questions on the licensee's methoa of documenting failed / delayed Inservice Fump and Valve-Test results.
(Closed) Unresolved Item (317/79-05-01; 318/79-05-01):
Valve Stroke Times Not Specified.
The insector reviewed STP-0-5-1, Revision 6, Auxiliary Feedwater Test, and noted that a maximum stroke time had been determined (20 seconds) and incorporated into the procedure acceptance criteria.
In addition, the licensee stated that this criterion had been incorporated in the valve stroke acceptance criteria for all full stroke tested safety-related valves.
(Closed) Unresolved Item (317/78-25-05; 318/78-19-05):
New Reference Values. As noted when this item was reinspected (317/79-12; 318/79-09), the licensee's procedures have been revised to incorporate new reference values for pump parameters.
The inspector reviewed procedure CCI 104D, Revision 3, Surveillance Test Program.
The procedure now contains, in section VII. B, provisions for making a determination of what Inservice Testing is required following maintenance and repairs to ASME Code Section XI pumps and valves.
(Closed) Unresolved Item (317/78-38-08; 318/78-34-07): WRONG TLD/ Identification Badge Issued.
Licensee badge issuance procedures have been revised to include better controls / checks prior to issuance.
These procedures will minimize the possibility of issuance of the wrong badge.
(Closed) Unresolved Item (317/78-23-01; 318/78-17-01):
Review of Completion of Licensee Actions Pursuant to I.E. Circular 78-08, Environmental Qualification of Electrical Equipment.
This item concerned the review of licensee actions with respect to IEC 78-08.
That circular has been superseded by IE Bulletin 79-01 and associated revisions and supplements.
Review of licensee actions in this area will be documented pursuant to IEB 79-01.
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'(Closed) bnresolved Item (318/79-16-01):
Safety Evaluation to be performed to Support Unit 2 Refueling.
The inspector reviewed the licensee's Safety Evaluation Report, Unit 2, Cycle 3 Reload Core
' Design and Fuel Design, performed pursuant to 10 CFR 50.59 and ~
dated November 15, 1979.
The licensee determination was made that no unreviewed ' safety question existed nor did the change require a change in the Facility Technical Specifications.- The Safety Evaluation
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was reviewed by the Plant Operations and Safety Review Committee (Meeting 79-172) and the Off Site Safety Review Consnittee (Meeting 79-13).. The reload fuel (Batch 2E) was identical in mechanical and metallurgical design and expected performance to that for cycle 2
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(previously analyzed).
No unacceptable conditions were identified during the inspector review of the Safety Evaluation.
c-(Closed) Unresolved Item (318/79-10-02):
Repair Leaking Safety Injection Tank Drain Valves.
The Safety Injection Tank Drain Valves were repaired during the fall,1979 Refueling outage.
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Shiftwise makeup and special leakage monitor logging are no longer
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(Closed) Unresolved Item (317/78-34-03; 318/78-31-03):
Evaluation
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of Diesel. Fuel Oil Day Tank Level Verific-lon.
This item had been
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left unresolved pending further licensee ansideration of installation of a more positive method of determining that minimum oil level is maintained (T.S. Surveillance 4.8.1.1.2.a.1, Monthly verification of at least 375 gallons (T.S. LC0 3.8.1.1.b.1).
The licensee has determined that the current method of verification, performed
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weekly during Surveillance Test STP 0-8-0, Diesel Generator Weekly Test, Revision 5, is satisfactory.
This test verifies and documents
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Day Tank level in excess of the minimum by observation of auto start of the fuel oil transfer pumps during diesel testing.
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tra%.er pumps provide fuel oil to the day tank and are started by actuation of a level switch physically installed in the tank above the minimum required volume.
Failure of the transfer pumps to start during routine diesel surveillance results in test failure.
,The inspector concurred that this is an acceptable method of verifying minimum Diesel Fuel Oil Day Tank level.
(Closed) Inspector Follow up Items (317/77-23-02; 318/77-22-02):
Revision of the Technical Specifications.
This item concerned the discovery by the licensee that if the RCS, SIT, and RWT boron concentrations were at the allowed maximum following a LOCA, the T.S. required minimum 75 cubic feet of TSP (trisodium phosphate dodecahydrate) would be insufficient to raise the sump water pH to 7.0; The licensee committed to administratively keep maximum baron
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concentration below 2,200 ppm (T.S. Limit was 2,700 ppm) and request t:
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a T.S. change stipulating the maximum.
The referenced change was requested in a BG&E letter dated November 13, 1978 and the Unit I and II Technical Specifications were amended (nos. 38 and 21, respectively)onApril 24, 1979 to require RWT and SIT baron concentrations less than 2,200 ppm.
(0 pen) Unresolved Item (317/78-25-11; 318/78-19-11) Revise Visual Snubber Inspection Procedure to Incorporate Piston Measurement.
The licensee stated that they had recently received a Bechtel formula to allow calculation of acceptance criteria to ensure the snubber has room for thermal growth without hitting the mechanical stops, and was performing field examination to determine validity of the formula.
The licensee further stated that measurement of the piston extension would be incorporated on a 40 month interval, per IWB-2500 examination category B-K-2 requirements, and that the surveillance program would be in place and started during the up-coming (Fall,1980) refueling outage for Unit I.
(Closed) Unresolved Item (317/79-12-03; 318/79-09-03); Hanger Settings. Not Verified.
The inspector reviewed Procedure NDE-5.702,
" Visual Exanination of Piping Systems and Attached Components" Revision 6 dated August 3,1979.
Procedure Attachment No. 3, Examination Data Sheet, was revised to required verification of support settings for variable and constant type spring hangers and snubbers.
In addition, the inspector reviewed Facility Change Requests Nos. 79-88 and 79-109.
These FCRs documented the licensee's review of Class 2 and 3 snubbers which resulted in Technical Specification changes to delete 15 Unit 1 Snubbers and add 15 Unit 2 snubbers.
(Closed) Unresolved Item (317/79-12-01; 318/79-09-01):
Temperature Effects on Snubber Operability.
The inspector reviewed a letter to i
BG&E from Bechtel Power Corporation dated November 7,1979 documenting the acceptability of the dynamic performance of Grinnell Hydraulic Suppressors between the ranges of 1"/ min to 40"/ min, Lockup Velocities.
Bechtel performed an analysis and indicated that the worst case
environment (Main Steam Tunnel,155 F) would result in 1cckup velocities less than the 40"/ min stated in the Grinnell guidelines.
(Closed) Noncompliance (317/79-23-04):
P1RV Inadvertent Actuation.
The licensee responded to this item in a letter dated April 3, 1980. As noted in the response the licensee determined the cause of the inadvertent actuation (physical disturbance of the control loops).
The inspector reviewed a Memorandum dated April 11, 1980 from the Nuclear Plant Engineer - Operations to all Shift Supervisors t
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addressing the categorization of inadvertent PORV actuation as a
" prompt" reporting requirement and directions to ensure that maintenance requests are initiated as soon as possible whenever operations requests repairs to safety related equipment.
(Closed) Unresolved Item (317/79-05-02; 318/79-05-02):
Installation of Low Pressure Switches on Auxiliary Feedwater Pump Suctions.
The inspector reviewed the maintenance requests associated with FCR 78-41, Installation of higher range pressure switches (MR Numbers 2-I&C 76-210,1-I&C-75-344, IC-79-054, and IC-79-090).
These MRs included the calibration results for the new 8arksdale suction pressure switches and observation of switch performance following a full flow trip test. ~ (lower range switches had failed during these conditions).
The tests were conducted satisfactorily.
(Closed) Open Item (317/80-03-02; 318/80-03-02): Adequacy of NRC Resident Inspector Office.
The licensee has provided a trailer within the protected area for use as the NRC Resident Office.
The trailer contains adequate space for the anticipated site manning level.
Inspector transfer to the new office was effected during this inspection period.
(Closed) Open Item (317/80-MR-02; 318/MR-02):
February, 1980 Monthly Report.
The March,1980 Operations Status Report requested recipients to correct the February 29, 1980 Average Daily Unit Power Level sheets (Unit 1 and 2 mw's were reversed).
(Closed) Unresolved Item (318/77-18-03):
RCP Seal Pressure Sensing Line Weld Failures.
The licensee has completed Facility Change Requests Nos.77-147 and 78-1028.
These FCRs added flexible hoses in the sensing lines for the RCP seal pressures.
These modifications have lessened the possibility of seal sensing line weld failures due to vibration induced cracking.
Update LER 79-03/0lX was submitted on April 25, 1980 delineating several of the past occurrences and completion of corrective action.
(0 pen) Unresolved Item (317/78-38-05; 318/78-34-04):
Establish Preventive Maintenance Cards for Gages.Used with Inservice Inspection Surveillance Procedures.
The Facility Change Requests to install gages for ISI Surveillance testing have not received implementation approval.
The licensee stated that PM cards would be established following installation of the gage..
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3.
-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 Peripheral Area, Security Buildings, Health Physics Control Points, Diesel Generator Rooms, Service Building and Intake Structure.
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In addition the inspector made rounds and observed the log taking
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with the outside operator during the day shift on April 11, 1980 and with the Unit 1 Auxiliary Building Operator during the evening shift on April 15, 1980.
The following observations and determinations were made:
Radiation controls established by the licensee, including
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posting of radiation areas, conditions of step-off pads and disposal of protective clothing were observed.
Control Room manning was observed on several occasions
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during the inspection, including observation of shif t turnover and panel walkdowns.
Systems and equipment in all areas toured were observed for
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the existence of fluid leaks and abnormal piping vibrations.
Seismic restraints and hydraulic snubbers were examined on a
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sampling basis to verify adequate installation and fluid levels.
Plant housekeeping conditions, including general cleanliness
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conditions and storage of materials and components to preclude 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 were established.
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On April 16 the inspector noted that the fire watch required to be stationed following discovery of three holes in the fire barrier b? tween the control room and dosimetry room was designated
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as the Unit No. 2 Control Room Operator.
The breaches in the fire barrier were in the overhead near the control room door, not actually visible from the control room operator's desk.
The inspector stated his concerns regarding adequacy of the fire watch and burdening of the control room operator (Unit 2)
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The licensee posted a dedicated fire watch in the control room until the fire barrier was repaired.
l During the tour with the Auxiliary Building Operator (Unit 1)
the inspector noted excessive lubricating oil on the floor in the Unit 1 Charging Pump Room Area.
This item was discussed
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with the licensee and the oil cleaned up.
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will be monitored on a continuing basis.
On April 16,1980 Unit 2 letdown flow was lost when Letdown Containment Outer Isolation Valve 2 CV-516 failed closed.
The valve is air operated, fails closed on loss of air pressure and is located inside containment.
Investigati'on outside containment could not determine the cause of the valve closure so the licensee decided to shutdown and enter containment to investigate. The inspector observed the plant shutdown from 100% power to the tripping of the main turbine between 4:45 p.m. and 7:30 p.m.
The shutdown was performed in an orderly fashion and in accordance with plant operating procedures.
Inability to dilute (letdown capability was lost) to stay critical at hot zero power inorder to overcome xenon poisoning resulted in completion of the shutdown during the evening shift.
Investigation of the Isolation Valve revealed a broken air line to the control vaPve, which was repaired and the reactor restarted.
No items of noncompliance were identified.
b.
Review of Operating Logs, Records A rev_iew 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 conformance to reporting requirements and Limiting Conditions for Operation.
The following records were reviewed for the report period:
Shift Supervisor's Log
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Unit 1 Control Room Operators Log
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Unit 2 Control Room Operators Log
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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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RAD-CON Smooth Log - April 12-15, 1980
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Unit No.1 Plant Transient and Operating Cycles Book - April
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entries No items of noncompliance were identified.
4.
Loss of Coolant Accident (LOCA) Procedure Review The inspector reviewed the licensee's LOCA procedures to verify that they were modified to incorporate the Combustion Engineering (C-E)
Owners Group Guidelines resulting from a review of the Three Mile Island accident.
The procedures were reviewed by comparison with the evaluations and.
guidel.ines contained in enclosures 1 and 2 to the letter dated Novtuber 14, 1979 from D. F. Ross, Jr., Director, Bulletins and Orders Task Force to G. E. Liebler, Chairman, Combustion Engineering Owners Group, and titled " Evaluation of Operator Guidelines for Small-Break Loss-of-Coolant Accidents in C-E Designed Operating Plants."
The following procedures were reviewed against the guidelines (procedures apply to both units):
E0P-5 Loss of Reactor Coolant, Revision 10 dated February 7,
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A0P-12 Loss of Coolant Accident Long Term Cooling, Revision 2
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dated February 14, 1980.
AOP-5 ECCS Long Term Cooling Core Flush, Revision 2 dated October
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25, 1978.
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E0P-6 Steam Generator Tube Rupture, Revision 4, dated February
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a.
Comparison to Guidelines The licensee's procedures generally contained the symptoms, imediate actions, subsequent actions and precautions identified in the CE owners group SBLOCA guidelines.
Deviations from the guidelines were either determined to be acceptable or unresolved
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as noted below by the inspector following review and discussions with the licensee.
The procedures do not differentiate initially between a
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small and large break LOCA. Precautions specifically appropriate to the small break LOCA are incorporated in appropriate supplementary actions in E0P-5 and actions in A0Ps 5 and 12.
This was determined to be acceptable.
The licensee's procedures do not contain, as a symptom, the
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guidelines' diagnostic chart.
The licensee stated that incorporation of such a chart would require an entirely new approach to Emergency Procedure Implementation, and that present procedures would require addition of the chart individually because a " general" emergency procedure philosophy was not utilized at Calvert Cliffs.
The licensee further stated that operators had been extensively trained in LOCA and steam break procedures and would not encounter problems differentiating between a steam break or LOCA. Based upon these discussions and operator interviews, the inspector determined that not including a diagnostic chart was acceptable.
Operators demonstrated knowledge of the different symptoms for steam line breaks, LOCAs and the special case LOCa, Steam Generator Tube Rupture.
E0P 5 did not contain guideline immediate action No. 4,
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verification of main feedwater flow or establish Auxiliary Feedwater Flow. The licensee stated that this was an unexpected complication for a LOCA and the step is currently contained as a Suppleaentary action.
The inspector stated that this immediate action was normally just a verification and appropriate.
The licensee stated E0P 5 would be revised to include Guideline Immediate Action 4.
This then is open pending completion of licensee actio M
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0 E0P 5 contained the criteria (50 F subcooling, pressurizer
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level indicated, unless inadvertent) for securing SIAS as the first supplementary action (caution statement). The licensee stated, and the inspector agreed, that securing of SIAS was not necessary as an immediate action and was appropriately placed as a supplementary action, i.e. it is neither a necessary nor desireable imediate action to discontinue operation of the safety injection system.
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Guideline precaution number 1, ensuring cooling services
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available prior to restarting the Reactor Coolant Pumps, was not addressed.
The licensee stated that Reactor Coolant Pumps are.not restarted once stopped as a supplementary action in any LOCA procedure.
Restarting of the Reactor Coolant Pumps following a LOCA would be reviewed and approved by the P0RSC as a recovery action procedure.
This was determined to be acceptable.
Guideline precaution numbers 3, (utilizing all available
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indications to diagnose event because accident may cause irregularities in a particular instrument), 7 (general precaution addressing temperature instrument utilization, RCP motor current, differential pressures to determine if RCS voiding is occurring) and 10 (high coolant activity may preclude utilization of shutdown cooling due to Auxiliary Building radiation levels), were not incorporated.
The licensee stated these precautions would be added to the procedures.
This is unresolved pending completion of licensee action.
Guideline symptom 4(i), Tave decreasing or at saturation,
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was not incorporated.
The licensee stated that this would be addressed by adding, as a symptom, few or zero
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degrees subcooling on the subcooled Margin Monitor since this instrument has recently been added and provides a direct monitor of RCS saturation..This item is unresolved pending completion of licensee action.
Revision of the E0Ps and A0Ps to include the items above is unresolved (317/80-05-01; 318/80-05-01).
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b.
Systems Considerations The inspector reviewed the procedures with respect to systems installed and modified by the licensee in response to STLL (Short Terms Lessons Learned).
The inspector noted that the operators were aware of system instrumentation modifications (PORV position indication, automatic AFW starting, subcooled margin monitors, ESFAS Reset Interlocks).
The irJpector noted that E0P-5 includes checklists to ensure that appropriate lines have been isolated following containment isolation.
Although the RAS (Recirculation Actuation System) will operate automatically at 30 inches in the Refueling Water Tarik, E0P-5 directs a manual shift in anticipation at the 4 foot level.
Appropriate precautions are incorporated to prevent deadheading of the HSPI pumps after recirculation commences.
The inspector noted that Guideline Followup Action No. 2, initiate simultaneous hot and cold leg injection within two hours, cannot be accomplished because CCNPP does not have hot leg injection capability. The licensee utilized A0P-5 ECCS Long Term Cooling Core Flush, within eight hours of a LOCA, to prevent the buildup of RCS boric acid to a level which woe'd result in crystylization.
No unaccepable conditions were identified.
c.
Operator Training and Interviews The inspector reviewed records of training received by operators relating to LOCA procedures and revisions implemented following the TMI-2 accident.
Interviews by the inspector were conducted with on shift Senior Operators and Operators and with staff licensed Senior Operators.
The Operators know the procedure imediate actions and understood the procedure supplemental actions, precautions and symptoms.
Documentation of the following formal training was reviewed by the inspector.
Formal Classroom Lectures given to all Licensed Operators
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during the weeks of July 9, 16, and 23, 1979 at the C-E Simulator, Windsor, Connecticut.
These lectures included j
Accident Analysis, Accident Identification, and TMI Pro-cedure Modifications.
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E0P-5, Reactor Loss of Coolant and A0P-12 Loss of Coolant
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Accident Long Term Cocling Procedure Reviews on January 7, 8, or 9, 1980 (attended by all licensed operators on one of these days).
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Additional NSSS Analysis Lecture given at the C-E Simulator,
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including E0P-5 walkthrough during February, March, and April 1980.
(Licensed operators only.
Senior Operators have been attending a special curriculum at i;he Charles County Comunity College to upgrade knowledge level to fulfill Shift Technical Advisor requirements.)
E0P-5, Loss of Coolant Walkthroughs between August,1979
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and November, 1979.
(NPE-0 gave to Shif t Supervisors; Shift Supervisors gave to Senior Control Room Operators; Senior Control Room Operators gave to Control Room Operators; and the Training Specialist (SCRO) gave to licensed plant staff.)
No unacceptable conditions were identified.
5.
Review of Events Requiring One Hour Notification of the NRC The circumstances surrounding the following events involving prompt (one Hour) notification of the NRC via the dedicated ENS (0PX)
telephone were reviewed.
April 21,1980:
Unit No. I tripped from full power at 11:16
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a.m. The inspector proceeded to the control room and observed the licensee's post trip actions.
The ESF were not actuated by the Reactor Trip, important plant parameters were verified to be normal-post trip and plant safety systems functioned as designed.
The cause of the trip was determined to be a sequential loss of Reactor protection System (RPS) Motor Generator (MG)
Sets (No.11 followed in approximately 20 seconds by No.12).
Testing showed that the second MG could not withstand the loss of the first.
Voltage regulator adjustments corrected the sequential tripping problem, but MG set output voltage oscillations persisted.
Restart was effected at 10:30 p.m. on April 22,
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April 25, 1980:
Unit No. I tripped from full power at 1:16
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p.m. The inspector proceeded to the control room and observed the licensee's post trip actions.
The ESF were not actuated by the Reactor Trip, Important plant parameters were verified to be normal.
Post trip and plant safety systems functioned as designed.
The cause of the trip was again determined to be loss of the RPS MG sets, this time during troubleshooting / adjusting of the voltage regulators with a vendor representative present.
MG set 12 had been taken off the line-between 12:48 p.m. and 13:08 p.m to adjust the voltage regulator. The set' was paralleled back on service and then No. 11 MG taken off at 13:15 p.m.
followed by a-reactor trip at 13:16 p.m.
Although continued adjustment of the voltage regulators were made and a series of tests performed demonstrating satisfactory loss of one MG set, the licensee noted that oscillations still existed with the machine outputs. The inspector (observed part of the MG set testing and noted the or*.put oscillations which were much more severe than output oscil.ations occurring on Unit 2).
Because the oscillations were worse when the sets were operated in parallel versus singly, the licensee decided to restart with only RPS MG set No.11 on the line.
During the start up, oscillations were noted on No.11 MG set output so No.12 MG set was used instead.
The reactor was made critical again at 00:12 a.m. on April 26, 1980.
Licensee continuing investigation of this problem revealed that, due to the highly cyclic and inductive nature of the MG set load (CEA Magnetic Jacks),
similar voltage oscillations had been encountered at other stations and had been corrected with the installation of a voltage regulator of a different design.
The licensee ordered new voltage regulators from the vendor and replaced the existing regulators when the unit was shutdown for other reasons on April 30,1980.
The new regulators appeared to completely correct the oscillation problem.
The licensee stated that the voltage regulators for Unit No. 2 RPS MG sets will also be replaced with the new design during a shutdown period.
The inspector stated that this item (50-318/80-05-03) will be followed by the NRC.
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No items of noncompliance were identified.
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6.
Shipment of Spent Fuel The inspector reviewed various documents and observed certain activities relating to a shipment of spent fuel from the licensee to Battelle Memorial Institute, West Jefferson, Ohio.
In accordance with 10 CFR 73.72 the licensee sent a letter to the NRC, Region I on April 10, 1980.
The required telephone communication was made to the resident inspector who reviewed the letter (Subject:
Notification of Special Nuclear Materials Shipment).
The inspecter also reviewed a letter dated February 26, 1980 to BG&E from tne NRC Physical Security Licensing Branch, Division of Safeguards, NMSS.
This letter affirmed that initial LLEA notifications had been made pursuant to 10 CFR 73.37(a)(2). The inspector noted that the route had been revised to avoid heavily populated areas (the approved route was verified to be in the driver's possession on April 18, 1980 by the inspector). The inspector noted that provisions for coping with threats and safeguards emergencies had been developed.
The activities associated with loading of the spent fuel cask after arrival on site were observed. These activities were accomplished in accordance with Procedure FH-33, " Procedure for Use of Spent Fuel Shipping Cask Type NLI 1/2" reviewed by the P0RSC on April 14,1980 (Meeting 80-58). The inspector noted that the Auxiliary Building Crane Shift Checks and Daily Checks were performed on April 15, 1980, prior to lifting the cask from the trailer to the washdown pit.
The inspector reviewed the radiological requirements specified on SWP 80-342, dated April 15, 1980, " Load Fuel Rod Shipping Basket into Shipping Cask and Ship Cask," and observed the implementation of the requirements during various portions of the transfer.
On April 18, 1980 prior to the shipment leaving the site the inspector wit-nessed the placement of lead seals on the trailer.
The inspector also verified that, in addition to the driver, the transport vehicle was to have one escort. The inspector interviewed the driver concerning radiological protection and examined the beta-gamma instrument which accompanied the truck.
Independent measurements were made by the inspector.
The following were noted:
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Location Limit Observed
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Cask Contact 200 mR 15 mR (neck)
10 ft 20 mR
<1 mR CAB 2 mR
.05 mR
2 Surface Contamination 2,200 dpm/100 cm 1000 dpm/100 cm *
- Licensee swipe surveys.
The driver tested both the radio telephone and CB radio in. the presence of the inspector and discussed operation of the truck's immobilization device which was observed by the inspector.
No unacceptable conditions were identified.
However, a problem with the trailer was identified by the driver enroute.
During the conduct of the two hour inspections it was noted that the trailer main I-beam was cracked.
The driver contacted LLEA who contacted a welder to respond to the scene and make' emergency repairs. After repairs were made, the shipment proceeded to a truck stop for further evaluation.
National Lead (NL), BG&E and Combustion Engineering reviewed the weld repair in a telephone conference on April 19. A NL Representative was dispatched to the scene to examine the repair and follow the shipment to Battelle. The licensee notified the NRC of the shipment status.
An engineering evaluation was performed on the weld repair and the shipment was resumed without further incidents, arriving during the evening on April 19, 1980. Appropriate NRC personnel were informed of the Trailer I-beam problem.
7.
In Office Review of Licensee Event Reports (LERs)
The inspector reviewed LERs submitted to the NRC:RI office to verify that the details of the event 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, whether generic implications were indicated, and whether the event warranted onsite followup.
The following LERs were reviewed:
LER (Unit No.)
Dated: Subject 79-01/3X(2)
4/25/80: Update Report.
RCS Leakage from socket weld on RCP 21A middle seal pressure sensing line.
Modified sensing lines and supports have been installed.
l 79-03/1X(.2)
4/25/80: Update Report.
RCS leakage from weld on RCP 22A lower seal pressure sensing line. Modified sensing lines and supports have been installed.
79-52/3X(.1)
4/25/80:
Update Report. Containment Gaseous and Par-ticulate RMS failed due to loss of sampling pump. Spare pump-installed.
Cause of failure was graphite particles from the pump deposited on the bearings causing seizure.
Manufacturer (Conde Pump Model #6) contacted and stated this is normal end of life failure mode for this pump when pumping ga.
i
LER (Unit No.)
Dated: Subject 80-05/3L(1)
4/25/80:
No.12 Control Room A/C Unit Tripped, No.11 Remained Operable.
Unit repaired (failed diaphram -
unipressure valve) and returned to service in 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.
80-11/3X(1)
4/E5/80:
Update Report.
Shutdown Cocling Suction Header Relief Valve Improperly Set.
Reports completion of corrective action to revise the setpoint control procedure to ensure a copy of all setpoint changes are routed to surveillance test coordinators.
80-13/3L(1)
3/25/80:
Channel D Containment High Pressure Trip Setpoint Drift (High) caused by failed signal isolator.
FCR 80-25 originated to replace Industrial Grade Capacitors with MILSPEC devices to improve reliability. Estimated completion date January 1,1981.
80-13/3X(1)
4/30/80:
Update Report.
Channel D Containment High Pressure Trip Setpoint Drift (High). This update report was requested by the inspector to include the magnitude of the setpoint drift to allow evaluation of the consequences.
(Required e 0 psig; as found 4.036.)
80-14/3L(1)
4/ 10/80:
RC Drain Tank and Quench Tank 02 Sample Valves Failed to Shut.
Corrosion products found on relay terminals.
80-15/3L(1)
4/ 17/80:
No.11 SG Auxiliary Feed Pump Taken Out of Service to Repair Broken Oil Sight Glass.
Update LER to be submitted following completion of FCR to add oil
<
petcocks for sampling purposes.
l 80-16/3L(1)
4/25/80:
Channel A Wide Range Nuclear Instrument Inoperability Caused by +_ 15 VDC Power Supply Failure.
80-17/3L(1)
4/25/80:
Surveillance Testing Determined ECCS Exhaust Train Charcoal Filter Removal Efficiency to be 88%
versus 90% required; original filter (1975) were replaced.,
!
80-19/3L(1)
4/25/80:
Channel B Loss of Load Trip Bypassed for Corrective Maintenance to Replace 24 VDC Auxiliary Master Trip Relay.
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n
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LER (Unit No.)
Dates: Subject 80-13/3L(2)
4/2/80:
Steam Generator Isolation Channel ZF pressure input signal bypassed.
Signal Isolator determined to be nonlinear, FCR 80-25 initiated to upgrade original
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isolators; to be completed by January 1, 1981.
80-16/3L(2)
3/28/80:
No. 21 Emergency Diesel Generator Slow Start Time. Governor replaced with a new unit.
80-17/3L(2)
4/10/80: No. 22 Saltwater / Service Water heat exchanger removed from service to repair a tube leak.
80-18/3L(2)
4/10/80: No. 23 Charging Pump Suction Line relief valve line discovered to have a leaking weld. Weld repair was made and a system design change, to be inspected by the NRC, has been initiated to correct recurring vibration induced metal fatigue cracks in the charging pump suction lines.
Modification to be completed during the fall,1980 refueling outage.
80-20/3L(2)
4/18/80: Channel D RPS High Power, ASI, TM/LP and Low Flow Trips Bypassed when Channel D High Power went into a trip condition.
80-21/3L(2)
4/18/80: No. 23 Saltwater Pump Did Not Start With a SIAS Signal.
Relay Socket Pin Found Broken.
No items of noncompliance were identified.
The inspector discussed the discovery, at other plants, of reports that Rosemont Model 1152 and 1151 A Pressure Transmitters may fail to normal output with an over or under ranged input. The licensee stated that Rosemont Pressure Transmitters were not utilized in Safety-Related Applications at CCNPP, however Rosemont 1151 Differential Pressure Transmitters were utilized for Reactor Coolant Pump diffbrential pressure (4 total each unit, and reactor vessel differential pressure (1 each unit). The licensee stated that this item would be further investigated and reviewed by the P0RSC.
This item will be reinspected by the NRC (317/80-05-02; 318/80-05-02).
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8.
Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted by the licensee pursuant to Technical Specification 6.9.1 and 6.9.2 were reviewed by the inspector.
This review included the following considerations: the report includes the information required to be reported by NRC requirements; test results and/or supporting information are consistent with design predictions and performance specifications; planned corrective action is adequate for resolution of identified problems; detemination whether any information in the report should be classified as an abnormal occurrence; and the validity of reported infonnation. Within the scope of the above, the following
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periodic reports were reviewed by the inspector:
March,1980 Operations Status Reports for Calvert Cliffs No.1 Unit
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and Calvert Cliffs No. 2 Unit, dated April 15, 1980.
Results of Steam Generator Eddy Current Examinations at Calvert Cliffs,
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Units 1 and 2, dated March 24, 1980.
Technical Specification 4.4:5.5.b requirement for inclusion in Annual Operating Report (discontinued).
Results for No.12 Steam Generator (Unit 1) and No. 21 Steam Generator (Unit 2) revealed no service induced degradations, no tubes plugged in Calvert Cliffs Steam Generators since they were placed in service.
9.
U_nresolved Items-Unresolved items are matters about which more information is required to determine whether they are acceptable, items of noncompliance or deviations.
Unresolved items addressed during this inspection are discussed in Para-graph 4 of this report.
10.
Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the~ inspection scope and findings. A summary of inspection findings was also provided to the licensee at the conclusion of this report period.