IR 05000324/1987006
| ML20210C022 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/29/1987 |
| From: | Fredrickson P, Garner L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20210B956 | List: |
| References | |
| 50-324-87-06, 50-324-87-6, 50-325-87-06, 50-325-87-6, IEB-83-06, IEB-83-6, NUDOCS 8705060068 | |
| Download: ML20210C022 (17) | |
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' UNITED STATES
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D NUCLEAR REGULATORY COMMISSION o
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REGION li h
101 MARIETTA STREET. N.W.
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. ATLANTA, GEORGI A 30323
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Report Nos. 50-325/87-06 and 50-324/87-06 Licensee: Carolina Power and Light Company P. O. Box 1551.
Raleigh, NC 27602
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Docket Nos. 50-325 and 50-324 License Nos. DPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Conducted: March 1 - April 3, 1987 Inspectors: SMn ig2,9/D g/W.H.'Ruland
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Approved By:
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P.'E. Fredrickson, Section Chief
/Date 'Si gn'ed Division of Reactor Projects SUMMARY Scope:
This routine safety inspection involved the areas of followup on previous enforcement matters, maintenance observation, surveillance observa-tion, operational safety verification, ESF system walkdown, onsite Licensee Event Reports (LER) review, review of special reports, circular followup, followup on inspector identified and unresolved items, onsite followup of events, and environmental qualification of electrical equipment.
Results: No violations or deviations were identified.
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B705060068 870429 PDR ADOCK 05000324
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a DETAILS 1.
Persons Contacted Licensee Employees P. Howe, Vice President - Brunswick Nuclear Project C. Dietz, General Manager - Brunswick Nuclear Project T. Wyllie, Manager - Engineering and Construction J. Holder, Manager - Outages R. Eckstein, Manager - Technical Support E. Bishop, Manager - Operations L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)
R. Helme, Director - Onsite Nuclear Safety - BSEP J. O'Sullivan, Manager - Maintenance G. Cheatham, Manager - Environmental & Radiation Control
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J. Smith, Manager - Administrative Support K. Enzor, Director - Regulatory Compliance A. Hegler, Superintendent - Operations W. Hogle, Engineering Supervisor B. Wilson, Engineering Supervisor B. Parks, Engineering Supervisor R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)
W. Dorman, Supervisor - QA W. Hatcher, Supervisor - Security
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R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)
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C. Treubel, Mechanical Maintenance Supervisor (Unit 1)
R. Poulk, Senior NRC Regulatory Specialist D. Novotny, Senior Regulatory Specialist
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W. Murray, Senior Engineer - Nuclear Licensing Unit
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Other licensee employees contacted included construction craftsmen,
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engineers, technicians, operators, office personnel, and security force
members.
2.
Exit Interview (30703)
The inspection scope and findings were summarized on April 3,1987, with the vice-president and general manager.
Three unresolved items were discussed (see paragraphs 6, 12 and 13).The licensee acknowledged the findings without exception. The licensee did not identify as proprietary
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any of the materials provided to or reviewed by the inspectors during the inspection.
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3.
Followup on Previous Enforcement Matters (92702)
(CLOSED)
Violations 325/83-11-03 and 324/83-11-02, Failure to Properly i
i Implement Procedures, and 324/83-20-03, Failure to Follow Procedure. The inspector reviewed the licensee's response to the violations and examined
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implementation of the corrective actions by reviewing the procedures and training records referenced.
Operating Instruction 01-13, which was issued on August 2, 1983, adequately addressed the concerns of the violations issued.
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(CLOSED) Violation 325/83-11-04, Failure to Make Timely Prompt Report.
The inspector reviewed the corrective actions taken for this violation and found them to be adequate. The inspector noted that the event reported in LER 1-87-01 was also not promptly reported.
However, it is the inspector's opinion that this is not symptomatic of a programmatic problem in this area.
(CLOSED) Violation 325/83-11-05, Failure to Submit a Special Report. The inspector reviewed the corrective actions associated with the Limiting Conditions of Operation (LCO) for fire barrier seals.
In addition, the inspector reviewed the active fire barrier seal LCO log and discussed maintenance of the log with a senior fire protection system specialist.
The licensee appears to have adequately addressed this violation.
(CLOSED)
Violation 325/83-11-06 and 324/83-11-03, Providing a Material False Statement. The inspector reviewed the licensee's corrective actions regarding the subject of this violation; preparation and maintenance of an accurate Q-List.
In addition, the. inspector reviewed the current Q-List and the procedures governing its contents and discussed their implementa-tion with the project engineer assigned to this task. Although the total program has not been completed, the corrective actions taken to date and the current program appear adequate.
These items are administratively closed.
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(CLOSED) Violation 324/83-20-02, Failure to Make Prompt Notification of a Significant Event. The licensee failed to r'ecognize with the Steam Jet Air Ejector (SJAE) radiation monitor inoperable, a 10CFR50.72 report should have been made within one hour. While the program has not been completely successful in eliminating late notification as evidenced by the late reporting noted in LER 1-87-01, the corrective actions have made significant improvements in evaluation and notification of events. The inspector has no further concerns with this event.
(OPEN)
Violation 325/84-27-01, Inadequate 10CFR50.59 Review.
The inspector reviewed the closeout package and interviewed selected engineer-ing personnel. While there has been increased training in the area of 10CFR50.59 review, the inspector was unable to determine that sufficient improvements had been made in the area.
This item will remain open pending further review.
(CLOSED)
Violation 325/85-04-01, Failure to Complete the Applicable Portions of the Post Reactor Trip Investigation Report Prior to Restart of Unit 1.
The inspector reviewed the licensee's response to the violation and the subsequent revision to Operating Instruction 01-22, Plant Incident
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and Post Trip Investigation, to preclude recurrence of this oversight.
The revision to 01-22 places more emphasis on completing the documenta-tion.
(CLOSED) Violation 325/85-16-01 and 324/85-16-01, Inadequate Surveillance Test Procedure for the Refueling Hoist Slack' Cable Interlock.
The inspector reviewed Procedure Test PT-18.1, Revision 20 and the applicable portions of the Technical Specifications (TS).
PT-18.1 addresses the concerns expressed in the Inspection Report No. 85-16 and is consistent with the TS.
(CLOSED)
Violation 325/86-15-01 and 324/86-16-02, Failure to Promptly Correct Deficiencies With the Designated Jumper Control Program.
The inspector reviewed the licensee's response dated August 14, 1986.
A review of operating instruction 0I-29, Operations Internal Audits, Revision 10, dated February 6, 1987, verified that the procedure was revised as committed. Step 4.11.1.2 of 01-29 requires, "If any jumper is not accounted for, an investigation will be conducted with the results presented to the Plant Nuclear Safety Committee (PNSC) prior to logging the jumper as missing."
On March 1, 1987, the inspector audited the designated jumper log.
No problems were discovered.
No violations or deviations were identified.
4.
Maintenance Observation (62703)
The inspectors observed maintenance activities and reviewed records to verify that work was conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The inspectors also verified that:
redundant components were operable; administrative controls were followed; tagouts were adequate; personnel were qualified; correct replacement parts were used; radiological controls were proper; fire protection was adequate; quality control hold points were adequate and observed; adequate post-maintenance testing was performed; and independent verification requirements were implemented.
The inspectors independently verified that selected equipment was properly returned to service.
Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenance.
The inspectors observed / reviewed portions of the following maintenance activities:
87-AISJ1 Snubber 1-E11-90SS267 Disassembly.
86-CAIY2 Actuator Rebuild for 2-E41-F008.
87-NTF091 Calibration of Diesel Generator No. 4 Differential Pressure Switches per OPIC-PS00.
s 87-NWA091 Calibration of Diesel Generator No. 4 Lube Oil Switch 2-LO-PS-6534-4 per MI-03-3W1.
No violations or deviations were identified.
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5.
Surveillance Observation (61726)
The inspectors observed surveillance testing required by Technical Specifications.
Through observation and record review, the inspectors verified that:
tests conformed to Technical Specification requirements; administrative controls were followed; personnel were qualified; instru-mentation was calibrated; and data was accurate and -complete.
The inspectors independently verified selected test results and proper return to service of equipment.
The - inspectors witnessed / reviewed portions of the following test activities:
2MST-AMI27M AMI Suppression Pool Temperature Monitor Channel Functional.
2MST-PCIS34R Primary Containment Isolation System Low Main Steamline Pressure Instrument Channel Response Time Test.
OI-3.2 Control Operators Daily Surveillance Report.
PT-12.2D No. 4 Diesel Generator Monthly Load Test.
No violations or deviations were identified.
6.
Operational Safety Verification (71707)
The inspectors verified conformance with regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system status.
The inspectors verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met.
Control room, shift supervisor, clearance and jumper / bypass logs were reviewed to obtain information concerning operating trends and out of service safety systems to ensure that there were no conflicts with Technical Specifications Limiting Conditions for Operations. Direct observations were conducted of control room panels, instrumentation and recorder traces important to safety to verify operability and that parameters were within Technical Specification limits. The inspectors observed shift turnovers to verify that continuity of system status was maintained. The inspectors verified the status of selected control room annunciator.
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Operability of a selected Engineered Safety Feature (ESF) train was verified by insuring'that:.each accessible valve in the flow path was in its ' correct position; each power supply and breaker, including control-room fuses, were aligned for components that must activate upon initiation signal; removal of power from - those ESF motor-operated valves, so identified by Technical Specifications, was completed; there was no leakage of major components; there was ' proper lubrication' and cooling water available; and a condition did not exist which might prevent fulfillment of the system's functional requirements.
Instrumentation essential to system actuation or performance was verified operable by observing on-scale indication and proper instrument valve lineup, if accessible.
The inspectors verified that the licensee's health physics policies / -
procedures were followed.
This included a review of area surveys,
- radiation work permits, posting, and instrument calibration.
The inspectors verified that:
the. security organization was properly ~
manned and security personnel were capable of performing their assigned functions; persons and -packages were checked prior to entry into the protected area (PA); vehicles were properly authorized, searched and -
escorted within.the PA; persons within the PA displayed photo identifica-tion badges; personnel in vital areas were authorized; and effective compensatory measures were employed when required.
The inspectors also observed plant housekeeping. cantrols, verified position of certain' containment isolation valves, checked a clearance, and verified the operability of onsite and offsite emergency power sources.
a.
Instrument Equalizing Valve Found Open The licensee found the equalizing valve for the Unit 2 Division II Core Spray (CS) sparger break detection differential pressure instrument (2-E21-dPIS-N004B) open instead of closed. The instrument was inoperable with the equalizing valve open.
TS 4.5.3.1.c.2, requires the above instrument to be calibrated every 92 days to demonstrate core spray system operability. The above surveillance requirement is implemented by 2-MST-CS21Q.
This Maintenance Surveillance Test (MST) was performed satisfactorily on March 14, 1987, and the instruments, both N004A and N004B, were satisfactorily returned to service. Step 7.7.5.5 required the technicians to close the instrument equalizing valve.
Independent verification was used as required when closing the valve.
The inspector reviewed the Maintenance Experience Report (MER) the licensee had prepared concerning the event.
The licensee had concluded that the technicians had correctly returned N004A to service.
This assertion was based on an auxiliary operator's-statement that he observed N004A operable on March 19, 1987, three
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days after completion of the MST. The equalizing valve was found open by a different auxiliary operator on March 21, 1987 at 8:45 a.m.
The Division II core spray system was declared inoperable upon discovery and returned to operable status when the equalizing valve
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was shut at 12:45 p.m.
The inspectors have concluded, based on the MER review and discus-sions with plant personnel, that the cause of the mispositioned valve remains indeterminate. Vandalism or multiple personnel errors cannot be ruled out. The inspectors will continue to monitor for repeat events of this nature.
b.
During the night of March 18, 1987, the Licensee discovered that both the normal and alternate 4160 KV breakers to the motor-driven fire pump were OFF. Operating procedure OP-41, Fire Protection and Well Water System, Revision 27, dated January 19, 1987, requires the breakers to be in the ON position. Operating personnel restored the breakers to their correct position.
With both breakers 0FF, the motor-driven fire pump was not operable as required per Technical Specification 3.7.7.1.a.
Action statement
"a" of Technical Specification 3.7.7.1 requires the pump to be restored to operable status within 7 days or issue a report to the Commission within 30 days stating the plans and procedures to be used to provide for the loss of redundancy in the system.
The Licensee has initiated an investigation into when and how the breakers were turned off. This investigation along with corrective actions will be documented in an Operating Experience Report (0ER). The inspector plans to review the OEk when issued. This is an Inspector Followup Item: Review Motor-Driven Fire Pump Breaker Misalignment OER (325/87-06-01).
No violations or deviations were identified.
7.
ESF System Walkdown (71710)
On March 31 through April 3, 1987, the inspectors verified that all major valves and electrical breakers associated with emergency diesel generators 1 and 3 were aligned in accordance with operating procedure OP-39, Diesel Generator, Revision 34, dated March 27, 1987, or under clearance such that the operability of the diesel generators was not affected. At the time of the inspection, air compressor No. 2, associated with the starting and control air system on diesel generator No. 1, was removed for repair. The inspector verified that valving had been realigned to allow air compressor No. I to pressurize the No. 2 air receiving tank. The inspection included walking down the fuel oil, lube oil, Jacket water cooling, and starting air systems. Portions of the demineralization and service water systems which support the diesel engines were also inspected. Hoses and belts were verified to be in good physical condition.
In general, the diesel engines appear to be well maintained and in good working order. The inspectors reported the following conditions to the licensee:
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o On March 31, circuit 14 in 120 V AC panel 1A-DG was found tripped.
Operations reset the breaker.
Per drawing 9527-LL-93041, sheet 15, circuit 14 supplies power to the generator control panel lighting and space heaters.
On March 31, diesel generator No.1 exhaust silencer moisture drain o
valve, DIE-V1, was found nearly closed.
Page 55 of OP-39 requires the valve to be open. An auxiliary operator (AO) stated that the valve was 10 to 20% open. Using a 2 to 3 foot long cheater bar he closed the valve about half a turn. Again using the cheater bar, he managed to open the valve about 75% before the handwheel threads stripped out. Work request 87-AKFQ1 was issued to repair the valve.
Review of plant records showed that the valve was checked and
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independently verified as open on January 23, 1987.
Because of the difficulty involved in manipulating the valve, the inspector feels that the January check may have been done improperly.
This possibility was pointed out to the licensee during the exit meeting.
Failure of the valve to be full open has no effect on the operability of the diesel ' generator as other drain paths were available as described below.
On April 4, the inspectors observed that the bottom of the exhaust o
silencers were rusted through in places. In addition, the inspector found pieces of the baffling material in the bottom of one of the silencers. The licensee does not consider the condition as adversely affecting the operability of the diesels.
o The inspector found that valve DSA-V25, Air Tank A pressure root valve to DG-PS-6524-1 and DG-PS-6525-1, listed on OP-35, page 45, was labeled as DSA-V23 in the field.
Drawing 9527-D-2265, sheet 1A, shows the correct number to be DSA-V25.
Correction of the material condition of the exhaust silencers is an Inspector Followup Item:
Repair of Diesel Generator Exhaust Silencers (325/87-06-02).
No violations or deviations were identified.
8.
Onsite Review of LERs (92700)
The listed LERs were reviewed to verify that the information provided met NRC reporting requirements. The verification included adequacy of event description and corrective action taken or planned, existence of potential generic problems and the relative safety significance of the event.
Onsite inspections were performed and concluded that necessary corrective actions have been taken in accordance with existing requirements, licensee conditions and commitments.
(CLOSED) LER 1-79-40, Re-analysis of Safety Related Hangers. This item is covered under Bulletin 79-07 and 79-14 programs.
The inspector reviewed the LER package. This item is administrative 1y close.
(CLOSED)
LER 1-83-31, Minimum Number of Operable Intermediate Range
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Monitor (IRM) Channels as Required by TS Was Not Met. The IRMs were repaired and the completed work packages were reviewed.
(CLOSED)
LER 1-83-45, The Steam Supply Inboard Isolation Valve for
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Reactor Core Isolation Cooling (RCIC), E51-F007, Did Not Fully Open. The inspector reviewed the completed work package which included the suspected root causes and the subsequent corrective actions.
In addition, the inspector reviewed the revised Maintenance Instruction, MI-16-537, and the applicable maintenance records.
(CLOSED) LER 1-83-47, Low Control Rod Drive (CRD) Accumulator Pressure.
The inspector reviewed Maintenance Instruction MI-10-5212 and the completed work package.
(CLOSED)
LER 1-83-58, Reactor Vessel Steam Dome Pressure Instrument Failed to Meet Response Time Due to Incorrectly Set Damping Adjustment.
The licensee corrected the damping adjustment and checked the balance of the instruments, including the spare instruments in stock. The inspector reviewed the licensee's corrective actions and completed work package.
(CLOSED) LER 1-84-02, Reactor Scram Due to High Pressure. The inspector reviewed the completed work package and determined that certain welding records associated with the repair of the Electro-Hydraulic Control (EHC)
system could not be located.
The balance of the work package was acceptable.
The LER is closed.
However, until the records can be located, it is an Inspector Followup Item:
Documentation of Welding Associated with LER 1-84-02,(325/87-06-03).
(CLOSED) LER 1-84-08, Thermal Overload of the Control Building Emergency Air Filtration System Fan Motors.
The inspector reviewed the revised Performance Test PT-35.4.5, Training Records, and the completed work package.
(CLOSED) LER 1-84-09, Isolation of Core Spray Minimum Flow Valve. The inspector reviewed the reports from General Electric and Byron Jackson and the completed modification package.
(CLOSED) LER 1-84-15, Inadequate Review of Effluent Monitoring Surveil-l lance Requirements. The inspector reviewed the applicable revisions to Periodic Tests and the completed work package. In addition, the inspector reviewed the change to the TS.
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LER 1-84-19, Reactor Building Roof Ventilation Monitor Inoperable.
The inspector reviewed the completed work package and
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Periodic Test PT-04.1.6P.
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(CLOSED) LER 1-84-26, Reactor Scram Due to Main Turbine Trip and Stop Valve Fast Closure.
The inspector reviewed Maintenance Instruction MI-03-401A and the completed work package.
(CLOSED) LER 1-84-33, Failure of Chlorine Detection System to Meet FSAR and TS Design Criteria. The inspector reviewed the final report on the evaluation of the BSEP chlorine detection system.
(CLOSED) LER 1-84-34, Reactor Scram Due to Spurious Upscale Trips of IRMs E and H.
The inspector reviewed the completed investigation report and the work package.
(CLOSED) LER 1-84-36, Manual Isolation of the Control Building Heating Ventilating Air Conditioning (HVAC) System While Investigating Leaks in the Chlorine System Storage Location.
The inspector reviewed the subsequent investigation reports and noted the minor chlorine leaks were located and repaired.
The work package appears complete.
(CLOSED)
LER 1-85-16, Primary Containment Valve Found Leaking During Local Leak Rate Testing. Several valves were found to be leaking, all of which were on lines below the minimum water level of the torus and did not represent a containment atmospheric leak path. The inspector reviewed the completed work package.
(CLOSED)
LERs 1-85-20, }-85-46 and 1-85-49, Inadequate Response Time Testing. The inspector riviewed the completed LER work packages and the final resolution contain41 in the LER 1-85-49 package. The information providedappearsadequate]
(CLOSED) LER1-85-30,IshlationofPrimaryContainmentGroups2,6and8 While Hanging Clearance for Main Steam Isolation Valve (MSIV) Outboard MSIV Actuator Power.
The' inspector reviewed the completed work package, Operating Procedure OP-3: (Revision 13), and the 120 V breaker label project.
(CLOSED)
LER 1-85-35, Idadequate Surveillance Test Procedure for the Refueling Hoist Slack Catie Interlock. See Violations 325/85-16-01 and 324/85-16-01 closeoutwriijeup.
(CLOSED) LER 1-85-36, Inhtiation of Standby Gas Treatment (SBGT) System and Isolation of Reactor. Building Ventilation System Due to Personnel Error.
The appropriate [ersonnel received counseling concerning the importance of thorough Orint research prior to hanging equipment clearances. The inspector, reviewed the training records.
(CLOSED)
LER 1-85-39, Prfmary Containment Group 1 Isolation and Core Spray Injection During Refueling.
The reactor was defueled, with the vessel head remcied, when procedural and personnel errors caused a depressurization 4,f the variable leg which resulted in a false low level 3
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signal and the subsequent injection of an estimated 25,000 gallons of, water. The inspector reviewad the corrective actions proposed in the LER report and the records of their implementation.
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(CLOSED)
LER 1-85-41, Isolation of Control Building HVAC System. The inspector -reviewed the licensee's completed work package and explanation of the event.
(CLOSED) LER 1-85-45, Reactor Protection System (RPS) Actuation Due to Spurious Upscale Electronic Noise Spike on IRM D Channel While the Unit Was in Refueling. The inspector reviewed the completed work package.
(CLOSED) LER 1-85-47, Primary Containment Groups 1 and 8 Isolation and Automatic Core Spray Initiation During Refueling, Due to a Pressure Spike in the Common Reference Leg Due to an Inadequate Procedure. The procedure has been corrected to preclude future recurrences.
(CLOSED) LER 1-85-51, RPS Actuation During Refueling Caused by Employee Bumping the Signal Cable of Local Power Range Monitor (LPRM) 3A-04-29 While Performing Work in Close Confines Underneath the Vessel.
The inspector reviewed the completed work package, (CLOSED) LER 1-85-54, Automatic Actuation of Control Building Emergency Air Filtration System During Refueling, Caused by a Cable Spreading Room
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Fire Alarm Actuation Which Resulted from Grinding in the Cable Spreading Room. The inspector reviewed the proposed corrective actions.
(OPEN)
LER 1-85-59, Reactor Scram Resulting from a False Low Reactor Water Level Which Occurred Due to a Pressure Spike on the Instrument's Common Reference Leg. This scram was complicated by the lockout of diesel generator No. 4 due to an air leak in the sensing line for lube oil pressure and the trip of the RCIC turbine due to a release of the trip latch mechanism. The inspector reviewed the LER package and portions of the corrective action documentation.
This item remains open.
(CLOSED) LER 2-82-55, Review of High Energy Line Break (HELB) Analysis.
The inspector reviewed the summary of the report from United Engineers and Constructors, Inc. covering HELB, the completed work package and the applicable 10CFR21 report. These adequately address the issue covered by this LER.
(CLOSED) LER 2-82-143, Reactor Water Cleanup Differential Flow Indicator Reading Downscale. This resulted from a plugged instrument line.
The line was cleared and the instrument was returned to service.
(CLOSED) ' LER 2-83-30, Trip of Both Reactor Recirculation Pumps Due to Level Disturbance in the Common Reference Leg. The cause was attributed to a spurious signal.
The inspector reviewed the completed LER packag.
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(CLOSED) LER 2-83-96, Valve Body Cracks in the Augmented Off-Gas (A0G)
Fire Protection Standpipe System. These items were added to the plant freeze protection system. The inspector reviewed FP-26 and has no further Concerns.
(CLOSED) LER 2-85-09, Primary Containment Group 6 Isolation During' Flush of Reactor Vessel Bottom Head Drain Line. Due to an inadequate flushing procedure, the licensee failed to address changing area radiation levels due to crud movement.
The licensee has revised the special flush procedure, for this specific case, to address this issue.
(CLOSED) LER 2-85-12, Reactor Scram Due to Flow Perturbation. The cause of the scram was not determined. All. systems functioned as designed. The inspector reviewed the two scenarios postulated.
There are no further questions at this time.
No violations or deviations were identified.
9.
Review of Special Report (90713)
(CLOSED) Special Report 325/NRE-81-02 and 324/NRE-81-02,13 Fire Barrier Penetrations Were Exceeded. The inspector reviewed the special report per TS 3.7.8.a, dated October 26, 1981, which delineated the problem involving the fire barrier penetrations. The inspector reviewed the current LC0 practices and the corrective actions associated with this special report.-
No violations or deviations were identified.
10. NRC Circular Followup (92703)
(CLOSED) NRC Circular 325/79-CI-24, Proper Installation and Calibration of Core Spray Pipe Break Detection Equipment.
The inspector reviewed MST-CS21Q, EER 79-499, and GE SIL 300 which covered the information in the circular.
No violations or deviations were identified.
11.
Followup on Inspector Identified and Unresolved Items (92701)
(CLOSED)
Inspector Followup Item 325/80-27-03, Review of DG Testing Requirements. The inspector reviewed the technical specifications and the information contained in the inspection report closecut package.
Activities in this area were found to be satisfactory.
(CLOSED)
Inspector Followup Item 325/82-08-02 and 324/82-08-02, Develop-
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Freeze Protection Checklist. The inspector reviewed 01-43, Revision 1, Freeze Protection and Cold Weather Bill, and determined it contained the information discussed in this inspector followup ite.
(CLOSED)
Inspector Followup Item 325/82-17-01 and 324/82-17-01, Correlation of In-Service Testing (IST) Required Testing with PT Test Procedure Requirements.
The inspector reviewed Engineering Procedure ENP-17, Revision 1, Pump and Valve In-Service Testing, and determined it
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contained the information discussed in this inspector followup item.
(CLOSED)
Inspector Followup Item 324/83-20-04, Docun:ent Discrepancies Existed in the Requirement for Independent Verification Between AP Section 11.5 and Temporary Instruction TI-104 Section 1.3.
TI-104, Revision 6, was issued to correct the discrepancies.
(CLOSED) Inspector Followup Item 325/83-30-02 and 324/83-30-02, Review of Leak Detection Procedure. The licensee performed the review of the leak detection system as compared with the guidelines of Regulatory Guide 1.45.
The results indicate that the leakage detection system's accuracy met the guidelines.
(CLOSED)
Inspector Followup Item 325/83-42-01 and 324/83-42-01, Resolu-tion of Senior Reactor Operator (SRO) Acting Simultaneously as Shift Technical Advisor (STA). AP, Revision 85 issued on March 28, 1984, stated that the SR0 and STA are separate positions.
(CLOSED)
Inspector Followup Item 325/84-04-02 and 324/84-04-02, Review of Proposed Long Term Corrective Actions Associated with Residual Heat Removal (RHR) Service Water (SW) pressure switches.
The inspector reviewed the report which addressed the long term corrective action for the RHR SW pressure switches.
(OPEN)
Inspector Followup Item 325/84-07-02 and 324/84-07-03, Revise PT-46.4 to Incorporate Acceptance Criteria Required to Verify the One Eighth Inch Water Gauge Positive Pressure in the Control Room Every 18 Months. The inspector reviewed Revision 4 of PT-46.4 which was issued on April 18, 1984, to verify the acceptance criteria was added.
Performance of PT-46.4 has shown that 1/8" is not obtainable. The matter is currently under review by NRR.
This item remains open pending establishment of a new limit and subsequent implementation into procedures.
(CLOSED)
Inspector Followup Item 325/84-08-01 and 324/84-08-01, Quality Assurance (QA) Response for I.E.Bulletin 83-06 Forwarded to NRC by June 1, 1984. The inspector confirmed the response was forwarded by the committed date.
(CLOSED)
Inspector Followup Item 325/84-30-01 and 324/84-30-01, Lack of Engineering Interface and Interdisciplinary Review of Maintenance Instructions. The licensee has completed their review of the engineering interface and interdisciplinary review of maintenance instructions and determined that, with the improvements that have been made in implementa-tion of the existing procedures, no changes were required. The inspector concurs with this positio..
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(CLOSED)
Inspector Followup Item 325/85-09-01 and 324/85-09-01, Independent Verification During Routine and Corrective Maintenance Testing.
TMI Action Item I.C.6 addressed the issue of independent verification. AI-59 and AP Section 11.7 are now consistent with current maintenance policy and TMI Action Item I.C.6.
(CLOSED) Unresolved Item 325/81-12-06, Review of Technical Specificatien on Shift Staffing. This item appears unnumbered in Report No. 81-12. The inspector reviewed the TS on minimum shift compliment, the procedures that implement the TS, and discussed current practices with the licensee.
(CLOSED) Unresolved Item 325/85-22-03 and 324/85-22-03, Drywell Average Temperature Calculations. The inspector reviewed the licensee's response to an UNR 85-22-03, the United Engineers report on primary containment average air temperature, and two reports on the effects of high drywell tempera-ture. The inspector discussed the derivation of the new coefficients with members of CP&L's technical staff. The coefficients are based solely on mathematical models. While the inspector feels it may have been prudent to field verify the coefficents calculated, the coefficients calculated are adequate for their intended purpose.
No violations or deviations were identified.
12. Onsite Followup of Events (93702)
a.
Unit 2 Scram On March 11, 1987, at 3:07 a.m.,
Unit 2 reactor scrammed while operating at 100% of full power due to reactor vessel water level i
I low. The High Pressure Coolant Injection (HPCI) and RCIC systems automatically started and injected into the vessel to recover level.
The MSIVs closed when level decreased to the group 1 isolation low level setpoint. At approximately 3:13 a.m., HPCI turbine tripped and the RCIC steam admission valve closed on high vessel level.
Between 3:30 and 3:35 a.m.,
three safety relief valve manual lifts were used for pressure control.
At 3:35 a.m.,
the MSIVs were opened for pressure control and feedpump A was started at 3:39 a.m., for level control.
Except for HPCI, all engineered safety features performed as designed.
This included starting of all four diesel generators, groups 1, 2, 6 and 8 containment isolations and starting of both trains of the standby gas treatment system. Approximately one hour after the scram, the control operator noted that the control board position indication for the outboard HPCI injection valve, 2-E41-F007, was out. The licensee found that the valve motor was shorted and severely damaged. The motor was replaced and the valve verified
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as operable prior to unit restart on March 13. The licensee has conducted an investigation into the circumstances surrounding the motor failure. The inspector has not completed the review of all the data and documentation concerning the failure. This item will be
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tracked as part of the unresolved item described below. pending further inspection.
The scram was caused by personnel error and possible inadequate training and procedures.
While attempting to troubleshoot the Uninterruptible Power Supply (UPS), the UPS bus power was momentarily interrupted.
Loss of power to portions of the feedwater control circuit resulted in both feedpumps running back to minimum speed. At minimum speed little or no flow was delivered to the vessel. The loss of feedwater initiated the low level scram. The licensee will submit an LER concerning the event. The inspector plans to review the LER for adequacy of corrective action to preclude repetition.
The adequacy of the corrective action and the E41-F007 motor failure review is considered an Unresolved Item: Review of March 11, 1987 Scram LER and E41-F007 Motor Failure Determination, (324/87-06-04).
b.
Chem-Nuclear Shipping Cask Bolt Torque Problems On March 17, 1987, the licensee discovered a problem with bolt torquing on a Chem-Nuclear shipping cask 8-120A-1 used for the
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transportation of Reactor Water Clean Up (RWCU) resin. After onsite movement, but prior to leaving the site, the torque settings were rechecked on the cask to cask lid bolts. The licensee found several bolts loose and other bolts below the required setting. The licensee re-created the condition on March 20, 1987, when the 32 cask bolts were torqued to 320 ft-lbs. and the flatbed tractor-trailer with the cask was driven on site. The cask was then returned to the protected area and the bolt torque re-checked. The licensee found 20 bolts at 320 ft-lbs., 10 bolts at various torque values between 120 and 300 ft-lbs., and 2 bolts loose.
Further investigation revealed that Chem-Nuclear had added thimbles to the liner / lifting device (wire rope). The licensee concluded that the thimbles acted like a spring, preventing the lid from seating properly.
Once the cask was moved, the thimbles would shift slightly, loosening the lid. The cask was returned empty to Chem-Nuclear for further investigation.
Chem-Nuclear agreed with the licensee and concluded that the problems encountered were caused by the liner cable thimbles interfering with the cask lid and preventing it from seating properly.
Chem-Nuclear has supplied the licensee with a new liner lifting device without thimbles.
The inspectors have concluded that the licensee's performance regarding this event was satisfactory.
No violations or deviations were identified.
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13.
Environmental Qualification (EQ) of Electrical Equipment (25576)
a.
Unit 1 RCIC Steam Admission Valve, 1-E51-F045 With the unit defueled, en March 10, 1987, the licensee reported that a terminal block in the Limitorque actuator for the above valve had not been qualified per main 00R guidelines since November 30, 1985.
The main DC power leads and motor leads were terminated on the block, manufactured by Kulka. Based on discussions with the manufacturer, the licensee believes that the part can be qualified.
The licensee (
has Qualification Data Packages (QDP) for similar parts made of similar materials by other manufacturers.
10 CFR 50.49 (j) requires that a record of qualification must be maintained for the entire period during which the covered item is installed in the nuclear power plant.
Since the qualification records did not exist from November 30, 1985, to the present, the requirements of 50.49 (j) were not met.
This item is the first example of an Unresolved Item that will remain unresolved pending overall NRC resolution of the EQ enforcement issue: Item Not EQ - U1 A
RCIC Steam Admission Valve Terminal Block, (325/87-06-05 and 324/87-06-05).
b.
Unit 1 SBGT System Temperature Switch Leads With the unit defueled, the licensee found on March 11, 1987, that leads connected to the IB SBGT system deluge system temperature switches (1-VA-TS-5296 to 5299) used unidentified wire. The licensee replaced the wire with qualified wire by March 14 under work requests 87-AICT1, U1, Y1 and Z1. The licensee has sent the wire to the Harris Environmental and Energy Center for identification. This item is example two of the above Unresolved Item:
IB SBGT Temperature Switch Lead Identification, (325/87-06-05 and 324/87-06-05).
c.
Torus Electrical Penetrations The licensee reported, on March 31, 1987, that torus penetrations IX232A,1X2320, 2X232A, and 2X2320, manufactured by CONAX, had no applicable QDP.
The error was discovered while the licensee was researching cable data for the torus and drywell. The penetrations are used for cabling associated with the Unit 1 and Unit 2 Suppres-sion Pool Temperature Monitoring System (SPTMS) detectors,1-CAC-TE-4426-02 to 13,1-CAC-TE-4426-15 to 26, 2-CAC-TE-4426-02 to 13, and 2-CAC-TE-4426-15 to 26, respectively.
The licensee does have a design qualification report from CONAX that qualified the penetra-tions but that information was never incorporated into a QDP. The licensee plans to document the qualification of these penetrations in a QDP.
The inspectors have no further questions regarding this particular item.
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d.
Primary Containment Hydrogen-Oxygen Analyzer Solenoid Valves The licensee discovered on March 25, 1987, that four Unit 2 primary containment hydrogen-oxygen analyzer solenoid valves for both analyzers (2-CAC-SV-4409-25 & 26 and 2-CAC-SV-4410-25 & 26) had not been replaced per the interval specified in QDP 51, Revision 0.
The valves had been rebuilt in April 1986, but instructions provided in the work requests (85-AFRE1, 85-AFRII, 85-AFRW1, 85-AFRY1) which invoked portions of MI-16-050, ASCO Solenoid Valves Rebuilding and Testing, failed to specify solenoid replacement as required by the QDP.
The licensee performed an analysis to show that even with the solenoid valves failed in the open position, the analyzers would have continued to function. The numbers 25 and 26 valves, associated with each analyzer, aid in condensate removal from the sample condensate drain trap.
Every 17 minutes, the numbers 25 and 26 valves, shut, isolating the sample dryer and precooler from the trap. The trap drains then open, providing a flowpath for the condensate back to the torus. The drain cycle lasts 20 seconds, after which the numbers 25 and 26 valves normally reopen and the trap drain valves shut. The analysis has shown that with the numbers 25 and 26 valves failed open, the condensation rate or fill rate of the trap is smaller than the amount that can be drained during the 20 second drain cycle.
Thus, the inlet piping of the analyzers will not back up with condensation in the event of a numbers 25 and 26 valve failure.
The licensee's failure to replace the numbers 25 and 26 valves per the QDP, was caused by inadequate implementing procedures.
The violation was identified by the licensee and all the requirements of 10 CFR 2 Appendix C, Section V. A were met, permitting the NRC to not issue a Notice of Violation. However, the inspectors will review the licensee's corrective action regarding this event. This is an Inspector Followup Item: Failure to Replace Hydrogen-0xygen Analyzer Valve Solenoids Per QDP, (324/87-06-06).
No violations or deviations were identified.
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