IR 05000327/1981030
| ML20033D273 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/22/1981 |
| From: | Butler S, Ford E, Quick D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20033D256 | List: |
| References | |
| 50-327-81-30, 50-328-81-38, NUDOCS 8112070477 | |
| Download: ML20033D273 (8) | |
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NUCLEAR REGULATORY COMMISSION
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I REGION 11 h
101 MARIETTA ST., N.W., Sti:TE 3100
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ATLANTA, GEORGIA 30303
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' Report flo. 50-327/81-30 50-328/81-38
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Licensee: Tennessee Valley Authority 500A Chesnut Street
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ChattanuJga, Tfl 37401
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u Facility Name:
Sequoyah Nuclear Plant Docket Nos. 50-327 and 50-328 License'lloc DPR-77 DPR-79 Inspection at Sequoy Nuclear Plant near Soddy Daisy, Tennessee Inspectors:
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'S. D. Butler j/
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Approved By:
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D.,R., Quick - Section Chief, Resident Date Signe1 e
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and Reactor Project Inspection Division
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Inspection on ' July 6,1981 Thru. August 5,1981
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[AreasInspected Jhis routine, unannounced inspection involved 168 inspector-hours onsite in tne
' areas of Operational Safety Verification, Unit 2 Preoperational and Safety
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Tasting, Plant I~cidents, Open Items and License Conditions, Independent n
' Inspection Efforts, and Radiological Emergency Preparedness drill observation.
Results
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Of the six areas inspected, no violations or deviations were identified in one
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area; five violations were -found.in five areas.
A. Failure to follow procedures
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for refueling operations (328/81-38-01), B.
Failure to establish procedures for testing UHI check valves (328/81-38-02), C.
Failure to follow fuel handling
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procedures (328/81-38-03) D. Failure to retrieve QA records (328/81-38-04) and E.
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Failure to maintain cleanliness of internals (328/81-38-05).
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DETAIL 5 1.
Persons Contacted Licensee Employees J. M. Ballentine, Plant Superintendent J. W. Doty, Assistant Plant Superintendent (Acting)
W. T. Cottle, Assistant Plant Superintendent J. M. McGriff, Assistant Plant Superintendent D. H. Tullis, Maintenance Supervisor (M) (Acting)
B. M. Patterson, Maintenance Supervisor (I)
J. A. Watson, Maintenance Supervisor (E)
D. J. Record, Operations Supervisor W. H. Kinsey, Results Supervisor R. J. Kitts, Health Physics Supervisor R. S. Kaplan, Public Safety Service Supervisor D. O. McCloud, Quality Assurance Supervisor M. R. Harding, Compliance Supervisor W. M. Halley, Preoperational Test Supervisor C. R. Brimer, Outage Director
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Other licensee employees contacted included construction craftsmen, technicians, operators, shift engineers, security force membe.. s, engineers, maintenance personnel, contractor personnel and corporate office personnel.
Other Organizations One engineer and two contractor personnel-0ffice of Nuclear Material Safety and Safeguards, various inspectors and contractor personnel-0ffice of Inspection and Enforcement.
2.
Exit Interview The inspection scope and findings were summarized with the Plant Super-intendent and/or members of his staff on July 17-21,- 1981.
The Plant Superintendent acknowledged the findings.
3.
Licensee Action on Previous Inspection Findings -
(Closed) Unresolved item 327/81-19-03, failure to lock high radiation area.
This licensee identified violation was discussed with Region II radiation specialists to determine if a Notice of Violation was required in accordance with the Revised Enforcement Policy.
It was detennined that the violation would constitute a licensee identified violation as described in the enforcement policy.
This unresolved item is close _-
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Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve noncompliance or deviations. New unresolved items identified during this inspection are j
discussed in paragraph 9.c.(2), 9.c.(3).
5.
Operational Safety Verification The inspector toured various areas of the plant on a routine basis throughout the reporting period.
The following activities were j
reviewed / verified:
a.
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 room and shift manning.
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The use of approved operating procedures, e.
Unit operator and shift engineer logs, f.
General shift operating practices.
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Housekeeping practices.
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Fire protection measures for hot work, i.
' Posting of hold tags, caution tags and temporary alteration tags.
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fleasures to exclude foreign materials from entry into clean systems.
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Personnel, package, and vehicle access control for the plant protected area.
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General shift security practice's on post manning, vital area access l
control and security force response to alarms.
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Surveillance, start-up and preoperational testing in progress.
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Maintenance activities in progress.
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Health Physics practices.
On July 7,1981, the inspector reviewed Surveillance Instruction SI-400.1,
" Radioactive Liquid Waste Effluent - Batch Release," for a release of. the cask decontamination tank that was in progress. The procedure was reviewed for implementation, completeness and adherence to license requirements. The inspector had no further questions on the procedure.
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On July 7-9, 1981, the inspectors participated with a team of Region II and Headquarters Nuclear Regulatory Commission inspectors in witnessing the licensee's annual Radiological Emergency Plan drill.
The inspectors found certain portions of the onsite activities to be inadequate; further details are provided in inspection report 50-327/81-26, 50-328/81-33, dated August 13, 1981.
On August 3,1981, the inspector entered the Unit 1 containment with a team
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of licensee personnel. The entry was made in an attempt to identify the source of increased secondary leakage and evaluate its impact on continued plant operation. A significant steam leak was found in Fan Room #1 from a steam generator blowdown isolation valve which sppeared to be the major contributor to the elevated temperature and humidity in the lower contain-ment. Although there was considerable condensation from cooling water piping there was no direct steam impingement or significant wetting of vital equipment or instrumentation.
Inside the polar crane wall there was no observable indication of steam leakage but there was indication of conden-sation. Again the condensation did not appear to be causing any detrimental equipment wetting.
The licensee repaired the leaking steam generator blowdown valve whicn reduced the leak rate but did not eliminate it. The unit was subsequently shut down to locate the remaining leakage which was found to be coming from a steam generator manway cover.
6.
Unit 2 Preoperational and Startup Testing The inspectors continued to n,aintain cognizance of Unit 2 fuel load preparations, fuel loading and preparations for startup subsequent to the receipt of a low power license on June 25, 1981. On July 13, 1981, the inspector discussed the Unit 2 core load verification with the licensee's lead reactor engineer.
In addition, the inspector reviewed completed Technical Instruction TI-1, "Special Nuclear liaterial Control and Account-ability System," and TI-45, " Physical Verification of Core Load Prior to Vessel Closure." The appropriate data sheets in TI-1 had been completed, including dual signatures to indicate two person verification of correct core had, however, due to equipment problems with the vidio recorder, a video tape of the core load was not obtained. According to Part II, Section 7.2, paragraph 3.0 of the Operational Quality Assurance fianual a video tape
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record of core loading is preferable but not mandatory. The inspector had no further questions in this area.
During reassembly of the Unit 2 reactor vessel, the threads of one of the head studs apparently galled and could not be completely inserted. The
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licensee was unable to extract the stud and subsequently was forced to drill the stud out and replace it with a space stud. The inspector discussed the problem and the remedial action with several cognizant licensee personnel and the Westinghouse onsite representative.
Region II management and technical personnel were informed of the problem and kept apprised of the licensee's actions by the inspector.
The removal of the damaged stud was uneventful and the reactor vessel was subsequently reassembled.
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On July 14, 1981, during a tour of Unit 2 containment, the inspector reviewed the official copy of Maintenance Instruction 111-1.2, " Removal and Replacement of Reactor Pressre Vessel Head and Attachments." The inspector noted 'that there had oeer. ito procedural sign offs since the placement of the upper internals package sc.eral days earlier.
In addition, HI-1.2, Appendix B, " Clearance Requirements for Vessel Disassembly," was not in the official procedure.
The inspector discussed these apparent discrepancies with outage
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and operations personnel. The inspector learned that the Appendix B
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clearance requirements had been establish after the completion of hot functional testing to support vessel disassembly but was removed afterwards to support preoperational test W-6.1F, " Integrated Engineered Safety Features Test." After completion of the test, the operations group'retagged
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l what they considered necessary to support vessel reassembly, however, Appendix B was not reestablished in its' entirety. The Appendix B clearance was reestablished after the discrepancy was identified by the inspector.
l The fa!iure to sign off procedural steps as they were completed was l
discussed with the outage refueling supervisor.
He claind he had personal I
knowledge that the work was being done as described by the procedure but due to the pace of activities he had not yet signed the steps.
He indicated he would sign the steps off at that time. The inspector cautioned the individual about not signin.) off work as being done unless he did have personal knowledge of it being completed, he acknowledged.
The inspector found no instance of work not being performed as described in the procedure.
The failure to properly sign off procedural steps for work being performed and failure to properly implement the Appendix B clearance requirements for vessel reassembly constitutes a violation of Sectica 6.8.1.b of Unit 2 Technical Specifications which states that written procedures shall be established 19 emented and maintained for refueling operations (328/81-
-38-01).
On July 15, 1981, during a tour of the Unit 2 containment, the inspector l
noted a test pump connected to the Upper Head Injection piping. The bent l
hose from the test rig was submerged in a bucket of water which had a film l
of oil on the surface of the water.
Further investigation revealed that the outage group had been performing leakage checks on the 12" Upper Head Injection-(UHI) check valves to detemine if maintenance was required. The inspector discussed his observations with the cognizant outage engineer and requested he review the leak test procedures to detemine what the clean-liness requirements were. The cognizant engineer stated that the leak
testing was being done by maintenance requests and this was their normal l
method for performing such evaluations. The inspector obtained copies of the maintenance requests (Nos. 75146,75147) for leak testing the 12" UHI check valves which stated that the valves were to be leak checked at a
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pressure specified by the cognizant engineer and the leak rate recorded.
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addition, the requirements were to use demineralized water, maintain piping cleanliness, prevent entry of foreign material and verify system cleanliness
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l during vessel reassembly per flaintenance Instruction HI-1.2.
The inspector
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asked if the maintenance requests had been reviewed by the-Quality Assurance Staff prior to performing work as required by Part II, Section 2.1, paragraph 4.1 of the Operational Quality Assurance Nanual (0QAM).-. He said
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it had not because they conside.ed them emergency maintenance requests which receive post-perfonnance review. The maintenance requests did not appear to qualify as emergency in nature as defined in the above mentioned paragraph of the 0QAM.
Failure to properly maintain system cleanliness and. failure to have approved written procedures for leak testing the Upper Head Injection 12" check valve constitutes a violation of Unit 2 Technical Specifications, Section 6.8.1.c, which states that written procedures shall be established, implemented and maintained for surveillance and test activities of safety-related equipment (328/81-38-02).
The licensee comenced Unit 2 fuel loading on July 3,1981. The inspector witnessed the conduct of fuel loading on July 4 and at various times through completion of fuel load.
The following procedures were in use:
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SU-6.1, " Initial Core Loading"
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SU-6.3, '" Reactor Systems Sampling prior to and During Core Loading" (c)
FHI-4B, " Movement of Fuel Assemblies with the flanipulator Crane
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The inspector reviewed the procedures and activities to ensure pre-requisites, communications, proper precautions and procedure compliance were in effect. There were no discrepancies discovered with the exception of workers in the refueling area not using a safety line as required by reference (c), paragraph IV.A. 'The descrepancy was first noted on July 4,1981, and was brought to management's attention.
While the inspector did not feel that nuclear or radiatics safety was involved it was nevertheless a procedural requirement and needed correction. The licensee agreed and expressed concern that safety rules were not being effected.
Internal measures were taken to correct the situation. On a subsequent inspection of the same area during the fueling operation, on July 8,1981, the inspector again noted the same reglect to implement the requirement.
This is a violation (328/
El-38-03).
7.
Plant Incidents On July 9, 1981, Unit 1 tripped from 100% power due to low water level in #3 steam generator.
The low level was caused when the air line to #3 tiain Feed Regulating Valve failed, causing the valve-to shut. The inspector verified that the Nuclear Regulatory Commission had been notified per 10 CFR 50.72.
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The circumstances of the trip and plant response were discussed with opera-tions personnel and no problems were noted. The air supply lines for the main feed regulating valves are fabricated of copper tubing with lead
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soldered joints and appear to be subject to vibration induced failure. The
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licensee has already modified the air lines on Unit 2 by providing expansion loops in the tubing and replacing the solder joints with compression type fittings. The modification is intended for Unit 1 at the earliest oppor-tunity. The broken air line.was repaired and the unit restarted on July 12,
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1981. On July 11, 1981, the Unit I reactor tripped twice on low steam
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generator level while attempting to restart.
The low level trips were due to difficulties in controlling steam generator level in the manual mode. No
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problems with plant response were noted and the Nuclear Regulatory Commission was properly notified per 10 CFR 50.72.
On August 5,1981, the auxiliary building was evacuated by the shift engineer due to increasing activity levels as read on the continuous area monitors. The inspector observed the licensee's response from the main control room. There was no indication on the control room monitors of any increased release of activity in the auxiliary building or shield building vent stack. Sample results from health physics indicated that particulate activity remained below maximum permissible concentrations (MPC) and noble gas activity reached a peak of 2.2 times HPC in the auxiliary building.
There was no indication of significant personnel contamination or exposure.
The leak was traced down to the "A" boric acid evaporator room. Activity levels began ta drop shortly after the room was sealed off.
Subsequent investigation by the licensee revealed the normal ventilation exhaust damper for the room had failed shut causing a buildup activity in the room and forcing it out into the auxiliary building through the door fire damper.
The action taken by the licensee to correct the problem and protect personnel appeared both-timely and adequate. The Nuclear Reguletory Commis-sion was informed per 10 CFR 50.72.
No violations or deviations were identified.
8.
Open Items and License Conditions
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(Closed) Open items 328/81-01-03 (NCR27P), 327/80-39-03 (LER 327-8040)
The inspector reviewed work plan WP-8922 (Enqineering Change Notice FCH-L5114) which modified the control circuitry for the breakers on the 1A, 1B, 2A and 2B 6.9 KV Shutdown Boards to provide control roaa indication when the breakers lockout and allow the control room operator to reset and attempt to shut the breaker manually from the control room when they do lockout. This modification was corrective action for the deficiency des-cribed in NCR 27P and LER 327-8040. The work plan was reviewed for com-pleteness and accuracy including the safety evaluation for the modification as required by 10 CFR 50.59.
These items are closed. The violation (327/
81-23-02) and deviation (328/81-28-02) associated with this discrepancey will remain open pending receipt and review of the licensee's responses.
(Closed) Open item 328/81-02-21 (NCR 28P)
The inspector reviewed the following documentation for charging pump 2A-A:
(a MR-107271 b
MI-6.4 c
HI-6.10 d
MI-6.15
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.and the following.do:umentation for charging pump 2B-B:
(a MR-71402 b
letter of March 30, 1981 c
letter of December 9, 1980 d)
IE Notice 80-38 e)
MI-6.4, Appendix A The inspector verified that the serial numbers of the pumps in place matched those in the above documentation.
This item is closed.
While reviewing a work package from a previous inspection (IE report no. 50-328/80-21) of the hydrostatic test performed on November 4,1980, for the reactor coolant system _ the inspector noted that the control room continuous tracing of system pressure was missing.
The licensee has made repeated attempts to locate and retrieve this document and, as of June 13, 1981, has been unable to produce the tracing.
This is a violation of 10 CFR 50, Appendix B, Criterion XVII, which states in part "... records shall be maintained to furnish evidence of activities affecting quality and that these records shall be retrievable." This item is designated 328/81-38-04, 9.
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 discussion of significant problems or incidents associated with the start-up testing and operations efforts.
a.
On July 15, 1981, the inspector accompained a member of the Office of'
Nuclear Material Safety and Safeguards (NMSS) on.a tour of the proposed low-level waste storage facility at the Sequoyah site. The license application for operation of the facility'is presently under review by NMSS.
b.
On July 22-24, 1981, the inspectors attended the quarterly resident-inspectors meeting in the Region II office in Atlanta,-Georgia.-
c.
On July 9,1981, subsequent to the completion of the core load and with the upper internals-in position in the reactor vessel the inspector made an independent inspection of the reactor head, refueling cavity, internals set down stands,_ transfer canal and general. vicinity; (1) The ' inspection of _ the upper internals set down _ stand revealed a
~1arge amount of organic material inside _of the circular I-twm ring of the stand.
This discovery 'took-place after the refueling
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cavity flood-up, which was perfonned on June 26, 1981, with the
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reactor vessel open; thus,- there was the possibility of foreign.
material entry into the reactor coolant system.
Discussions with the licensee were held to express concerns over the possibility-of_-
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autoclaved a representative sample, traced possible migration paths and sent a diver into the RWST to determine if there has been samples deposited there. An engineering analysis has concluded that the material presents no operational problems and will be dissolved at temperature. Due to the failure to implement cleanliness requirements of 10 CFR 50, Appendix B, Criterion II as detailed in MI-12, " Removal and Replacement of RPV Head and Attachments," this is a violation (328/81-38-05).
(2) The inspection of the general vicinity included the RCCA (Rod Cluster Control Assembly) change fixture.
The fixture had a carbon bolt as evidenced by its rusting and magnetic qualities.
Additional carbon bolts were found on the lower internals set down stand. These two items are designated unresolved pending further inspection to determine whether carbon steel is appropriate for use in an otherwise stainless steel environment. This item is unresolved (328/81-38-06).
(3) The inspection of the general vicinity revealed a high scale (in excess of 250,000 lb.) Dillon Load Scale. The scale did not have a calibration sticker on it.
In a discussion with an outage supervisor it was noted that the cell was used in conjunction with the polar crane readout, as a backup, to determine if there are load changes during heavy lifts.
This would indicate, for example, reactor vessel head binding as it is lowered on the guide studs.
This item is designated unresolved pending further inspection efforts to determine whether the use of an uncalibrated instrument was appropriate for this application (328/81-38-07).
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