IR 05000327/1982003
| ML20042B708 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/05/1982 |
| From: | Butler S, Ford E, Wilcox J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20042B700 | List: |
| References | |
| 50-327-82-03, 50-327-82-3, 50-328-82-03, 50-328-82-3, NUDOCS 8203250556 | |
| Download: ML20042B708 (8) | |
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UNITED STATES
NUCLEAR REGULATORY COMMISSIEN
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REGION 11
o 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303
Report Nos. 50-327/82-03 and 50-328/82-03 Licensee: Tennessee Valley Authority 500A Chestnut Street Chattanooga, Til 37401 Facility llame: Sequoyah fluclear Plant Docket flos. 50-327 and 50-328 License flos. DPR-77 and DPR-79 Inspection at Sequoyah site near Soddy Daisy, Tennessee Inspectors:
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3W/L E. J. Ford C
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D&te' Signed
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S. D. Butler 0
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Date Signed Y[
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J. D. Wilcox, Jr
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Dat'e Signed Approved by:
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c J Ph D. R. Quick, S(ction Chief, Division of Da'te Signed Resident and Reactor Project Inspection Sultr1ARY Inspection on January 6,1982 - February 5,1982 Areas Inspected This routine, unannounced inspection involved 185 inspector-hours on site in the areas of operational safety verification, Unit 2 License Conditions, Independent Inspection Effort, Plant Incidents, Protective Coatings, and General Employee Training.
Resul ts Of the six areas inspected, no violations or deviations ~ were identified.
8203250556 820305 PDR ADOCK 05000327 O
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DETAILS 1.
Persons Contacted Licensee Employees C. C. Mason, Plant Superintendent J. W. Doty, Assistant Plant Superintendent W. T. Cottle, Assistant Plant Superintendent J. M. McGriff, Assistant Plant Superintendent D. H. Tullis, Maintenance Supervisor (H)
B. M. Patterson, thintenance Supervisor (I)
W. A. Watson, ikintenance Supervisor (E)
L. ti. Nobles, Operations Supervisor W. H. Kinsey, Results Supervisor R. J. Kitts, Health Physics Supervisor J. T. Crittenden, Public Safety Service Supervisor R. L. Hamilton, Quality Assurance Supervisor ti. R. Harding, Compliance Supervisor W. M. Halley, Preoperational Test Supervisor J. Robinson, Outage Director Other licensee employees contacted included field services craftsmen, technicians, operators, shift engineers, security force members, engineers, maintenance personnel, contractor personnel and corporate office personnel.
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2.
Exit Interview The inspection scope and findings were summarized with the Plant Superintendent and/or members of his staff on January 7,1982 and February 8, 1982. The violation concerning protective coatings was discussed and the licensee acknowledged.
During the reporting period frequent discussions are held with the Plant Superintendent and his assistants concerning inspection findings.
3.
Licensee Action on Previous Inspection Findings Not inspected.
4.
Unresolved Items One unresolved items was identified in connection with maintenance of protective coatings and is discussed in paragraph 9.
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5.
Operational Safety Verification The inspector toured various areas of the plant on a routine basis through-out the reporting period. The following activities were reviewed / verified:
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Adherence to limiting conditions for operation which were directly observable from the control room panels.
b.
Control board instrumentation and recorder traces.
c.
Proper control rcom and shift manning.
d.
The use of approved operating procedures, e.
Unit operator and shift engineer logs.
f.
General shift operating practices.
g.
Housekeeping practices.
h.
Posting of hold tags, caution tags and temporary alteration tags.
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Personnel, package, and vehicle access control for the plant protected area.
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General shift security practices on post manning, vital area access control and security force response to alarms.
k.
Surveillance, startup and preoperational testing in progress.
1.
flaintenance activities in progress.
m.
Health Physics Practices.
On January 8,1982, the inspector observed the setting of limit switches on Unit 2 valves 2-FCV-63-39 and 40 after mainternace was performed on the valves. The valves are parellel inlet valves to the boron injection tank which are required to open upon receipt of a safety injection signal. The inspector verified that an approved procedure, Maintenance Instruction MI-11.2 " Motor Operated Valve Adjustment Guidelines", was at the job site and in use. The wor:< vas discussed with the craftsmen involved and the inspector had no further questions.
Following completion of the work, the inspector observed stroke testing of the valves in accordance with Surveil-lance Instruction SI-166.1 from the main control room. The valves appeared to operate properl r,
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On January 27, 1982, the inspector verified proper alignment of major flow path valves for the Emergency Core Cooling Systems and the Containment Spray System for Unit 2.
The unit was at 30% power at the time. Motor operated valves were verified by position indication in the main control room and manually operated valves were verified by local observation.
No problems were noted.
On January 28, 1982, the inspector witnessed the stroke timing of several Unit 1 Essential Raw Cooling Water (ERCW) containment isolation valves in accordance with Surveillance Instruction SI-166.3, " Full Stroking of Category A&B Valves during Cold Shutdown". The valves appeared to operate within the maximum time allowed by technical spec.ifications when tested in accordance with the procedure. The inspector nooM on the data sheet being used that there were stroke times entered under a heading entitled
" Calculated Maximum Allowable Stroke Time" which exceeded the maximum times allowed by technical specifications. This was discussed with the lead mechanical results engineer who indicated that these times were calculated
.in accordance with the applicable ASME codes to determine the need for increased testing frequency. The inspector indicated that personnel could be misled to believe that the stroke times of the valve could exceed the tech spec times by the data sheet.
The engineer agreed and stated he would
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clarify the data sheet to prevent any misunderstanding. The inspector had no further questions.
On January 28 and 29, 1982, the inspector uitnessed time response testing of Unit 1 Engineered Safety Features Equipment required by technical specifi-ca tions. On January 28 the inspector witnessed IllI-99-RT-16.7, Appendix A, which response tested containment pressure transmitter 1-Pdt-30-43.
The inspector reviewed the procedure in use and test equipment setup, verified proper calibration of test equipment and witnessed the response time of the Hi and HiHi containment pressure setpoints.
After the test was completed the inspector verified that the response times met the acceptance criteria
of the procedure. On January 29 the inspector witnessed IllI-99-RT-N42 which
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response tested one of the Unit 1 power range neutron detector circuits.
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The inspector reviewed the procedure in use and the test setup, verified proper calibration of test equipment and witnessed the response test of the low neutron flux, high neutron flux and negative rate trip setpoints. After the test was completed the inspector verified that the response times met the acceptance criteria of the procedure. The inspector had no further
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questions in this area.
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On January 29, 1982 the inspector witnessed the eighteen month service test of the IV 125 VDC vital battery required by technical specifications. The inspector reviewed the procedure in use, Surveillance Instruction SI-105
" Vital Battery Bank and Charger Performance Tests".
The test setup was
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reviewed and the proper calibration of test equipment was verified.
During the test the inspector witnessed periodic voltage measurements made on individual battery cells. At the completion of the two hour battery discharge test a four hour capacity test was performed on the IV battery charger. The battery and charger appeared to meet the acceptance criteria of the test. The inspector had no further questions.
As a result of a Westinghouse Technical Bulletin NSD-TB-81-10, the licensee reviewed maintenance work previously performed on the rotating elements of the Unit 1 and 2 centrifugal charging pumps.
It was determined that both of the rotating elements of the Unit 2 centrifugal charging pumps were replaced with elements that did not have specific performance curves. The inspectors were informed of this on February 3,1982. The licensee's Nuclear and flechanical Engineering Groups had performed a safety evaluation to justify continued operation after consulting with the pump manufacturer and Westing-house. The pump manufacturer stated that the spare rotating elements are cast and machined to specifications of a standard manufacturing drawing which meets the necessary Westinghouse specifications. The licensee concluded that they were confident that the pumps would exceed the minimum flow requirements; however, they were not as-confident that the cumps would not exceed the maximum flow limits.
To ensure operability the licensee will insert a step in applicable Emergency Operating Instructions (E01) to monitor pump flow rate and throttle the pump discharge valves if necessary to prevent pump runout. The inspector reviewed surveillance data on the pumps before and after changeout of the rotating elements and there was no significant change of pump performance while operating on recirculation fl ow. The inspector obtained and reviewed documentation concerning the licensee's safety evaluation and reviewed changes made to E01's. The licensee stated that at the first outage of sufficient duration they would change out the rotating elements with ones of known performance character-istics.
Region II management and pump specialists were informed of the situation and had no further questions.
No violations or deviations were identified.
6.
Unit 2 license Conditions Unit 2 license, DPR-79, paragraph 2.c.(15) requires that the licensee carry out a surveillance program to monitor the corrosion of carbon steel piping in the Essential Raw Cooling Water System (ERCW).
Procedures to implement the program were to be submitted to the Nuclear Regulatory Commission by October 15, 1981 for concurrence. The licensee submitted Surveillance Instruction SI-566 "ERCW Flow Verification Test" to the Office of Nuclear Reactor Regulation (NRR) on October 15, 1981.
The inspector reviewed SI-566 and informed Region II that the procedure appeared adequate to identify system flow reduction caused by piping corrosion.
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Unit 2 license, DPR-79, paragraph 2.c.(16)m. requires *Nt by January 1, 1982, the licensee shall install a backup indication for incore ther-mocouples in the control room to cover the range of 200 F to 2000'F. The Unit 1 license DPR-77 was conditioned similarly in late December 1981; however, the licensee contacted the Office of Nuclear Reactor Regulation (llRR) dnd informed them that they would not be able to install the 2000 F indication in Unit 1 until January 18, 1982. NRR concurred witn the extended installation date for Unit 1 of January 18, 1982. The inspector reviewed work plan WP-9565 and visually inspected the Unit I control room indication for the incore thennocouples. According to the work plan the backup indication was completely installed and calibrated on January 15, 1982. Licensee condition 2.c.(23)H.(1) is closed for Unit 1.
The inspector reviewed liaintenance Request 11R 161271 und visually inspected the installed control room wide range indication for the Unit 2 incore thermocouples. The new indication was installed and calibrated on December 22, 1981. License condition 2.c.(16)m. is closed for Unit 2.
No violations or deviations were identified.
7.
Independent Inspection Effort The inspector routinely attended the morning scheduling and staff meetings during the reporting period. These meetings provide a daily status report on the operational and testing activities in progress as well as a discuss-ion of significant problems or incidents. associated with the start-up testing and operations effort.
The inspectors traveled to the Region II office on January 12, 1982 to attend the quarterly resident inspector regional meeting scheduled for January 13-15, 1982.
Due to the onset of severe weather conditions on the af ternoon of January 12, the meeting was disrupted. On January 14 the inspectors elected and received concurrence that they return to the area, if possible, in anticipation of possible plant problems.
No significant problems developed.
8.
Plant Incidents During the reporting period both Unit 1 and Unit 2 experienced numerous inadvertent or unplanned shutdowns.
In each instance the inspectors reviewed the surrounding circumstances including cause and corrective measures, notification of the NRC per 10 CFR 50.72, unit recovery and restart and related problems. On January 19, 1982, following an inadvertent reactor trip, Unit 2 experienced a low pressure safety injection. The low pressure was caused by excessive cooldown due to secondary steam leakage and a lack of decay heat from the new core. The recovery was uneventful. The
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inspector discussed the incident with plant management. The licensee stated that several steps would be added to the immediate actions of the inad-vertent reactor trip procedure A01-1 which would require the operator to imediately identify and isolate sources of secondary steam leakage to prevent plant cooldown following a trip. On January 19, Unit 1 tripped following a ground fault on the main generator isolated phase bus system.
The fault also caused a neutral transformer explosion which completely destroyed the transfomer. There were no injuries and the fire was quickly extinguished by onsite personnel. The unit commenced a scheduled surveil-lance outage early while efforts were begun to replace the neutral trans-former and damaged bus work. Following the transformer explosion the inspector verified that the unit was in a stable condition with undamaged offsite power sources. Unit I heatup following the outage was underway as of the end of the reporting period.
No violations or deviations were identified.
9.
Protective Coatings On January 6 and 7, 1981, a review was performed by the inspector of the following allegation: white spray can enamel paint had been used in touch-up painting on the manipulator crane in Reactor Building #2. This review included a direct observation of protective coating on the manipu-lator crane RB#2, numerous deck areas in upper compartment elevation 733.63 RB#2, and the dome liner RB#2, and Unit 2 main steam valve room B deck. TVA had made a previous investigation in this same allegation. TVA's investi-gation is documented in a flay 5,1981 memo from C. E. Greek, III to Sequoyah Nuclear Plant Files titled:
"Sequoyah Nuclear Plant Unit 2 - Protective Coating - Alleged use of Unauthorized Material". This investigation took samples from the apparent touch-up area and sent these samples to Singleton Material Engineering Lab with a sample of Carboline 305 topcoat to determine if all of the suspect samples were Carboline 305.
R. O. Lane of Singleton Materials Engineering Lab stated (with documentation results) that the laboratory testing of the suspect samples revealed that they were Carboline 305. This allegation was not substantiated since the areas touched-up were in fact Carboline 305.
However, during the observation of the protective coating on the manipulator crane, and numerous deck areas in upper smpart-ment elevation 733.63, it was noted that the protective coating was flaking / pealing up. The exact location of these areas was pointed out to site QA. General Construction Specification G-55, Rev. 2, for surface preparation, application, and inspection of special protective coatings for nuclear plants specifies the criteria to protect coatings.
Para-graph 1.2 of G-55 states:
Surface preparations and applications of special protective coatings to steel and concrete surfaces in a nuclear power plant must be properly done as the coated surfaces will be subjected to unusual conditions of exposure
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and usage.
In addition, the integrity of the coatings must be maintained for long periods of time during which repairs to the coatings may be impractical.
Clean, dry, properly textured surfaces and coatings free of pinholes are mandatory to ensure the proper adhesion, performance, and decontaminability of these coatings. While some plant areas will be subjected to less severe exposure than others, a uniformly high standard of surface preparation and coating application must be maintained throughout the plant.
Several areas on the manipulator crane, and numerous deck areas in upper compartment elevation 733.63 (RB#2) where the protective coating was
flaking / pealing up were not being maintained. This is identified as Unresolved Item 50-328/82-03-01, " Protective Coating Maintenance".
In addition, the large portion of the main steam valve room B had the deck protective coating pealing up.
This area is not a severe level I 6rea; however, the protective coating was applied in the same manner (ie G-55 guidelines) with documentation not required.
No violations or deviations were identified.
10.
General Employee Training (GET)
The inspector continues to attend GET classes to monitor their content quality and instructor proctoring. This topic was the subject of a Notice of Violation (327/81-42-02, 328/81-52-03) in the December 6, 1981 -
January 5, 1982 inspection report. The inspector has noted an increased concern for proctoring as evidenced by the mandatory reading of a management memorandum prior to class commencement, the implementation of spaced seating, and the presence of a designated management member as a additional proctor during the exam phase.
The inspector has not observed further breaches of exam integrity and will continue to monitor the program.
No violations or deviations were identified.
11.
Control Rod Operating Criteria In view of the unresolved nature of the issue surrounding multiple dropped control rods on Westinghouse reactors the Region II office solicited a commitment from TVA to continue the interim control rod operating criteria in effect at Sequoyah until NRR resolves ti issue with Westinghouse.
In a telephone conversation between a member of the Region II staff and the Sequoyah Plant Superintendent, on February 1,1982, TVA committed to continue the interim operating criteria, as described in Section 4.2 of the Sequoyah Safety Evaluation Report, (NUREG-0011, Supplement 1, February 1980), in effect until concurrence is obtained from this office to remove them.
No violations or deviations were identified.