IR 05000327/1986049

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Insp Repts 50-327/86-49 & 50-328/86-49 on 860906-1005. Violation Noted:Failure to Adequately Determine Cause of Repeated upper-head Injection Isolation Valve Surveillance Failures
ML20207P335
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 12/07/1986
From: Debs B, Harmon P, Hunegs G, Jenison K, David Loveless
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20207P327 List:
References
50-327-86-49, 50-328-86-49, NUDOCS 8701150426
Download: ML20207P335 (12)
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Report Nos.: 50-327/86-49, 50-328/86-49 Licensee: Tennessee Valley Authority 500A Chestnut Street Chattanooga, TN 37401 Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79 Facility Name: Sequoyah Units 1 and 2 Inspection Conducted: September 6, 1986 - October 5, 1986 Inspectors: YY M /A b3INI /

Date Signed K. N. Eeryfson, Senior Resideiit Inspector

& mv Je 1k.3. I9VL Date Signed P~./E. Ha~y fnon, Resident InI6pector 4 J+ Dae 3.11yC

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6. P.' Lpeless, ResidenF Inspector DatT! Signed

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/9/f G. K. liunegs, Rejprtor Engineer Date3, Signed Approved by:

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Date~51gned Division of Reactor Projects SUMMARY Scope: This routine, announced inspection involved inspection onsite by the Resident Inspectors in the areas of: operational safety verification (including operations performance, system lineups, radiation protection, safeguards and housekeeping inspections); maintenance observations; review of previous inspection findings; followup of events; review of licensee identified items; review of IE Information Notices; and review of inspec+.or followup items, Results: One violation (327, 328/86-49-01) - failure to take adequate correc-tive action was identified, paragraph t 8701150426 8702og

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REPORT DETAILS 1. . Licensee Employees Contacted H. L'. Abercrombie, Site Director

  • P. R. Wallace, Plant Manager
  • L. M. Nobles, Operations and Engineering Superintendent B. M. Patterson, Maintenance Superintendent
  • J. Prince, Radiological Control Superintendent
  • R. Harding, Licensing Group Manager W. E. Andrews, Site Quality Manager D. W. Wilson,. Project Engineer
  • R. W. Olson, Modifications Branch Manager J. M. Anthony, Operations Group Supervisor R. V. Pierce, Mechanical Maintenance Supervisor M. A. Scarzinski, Electrical Maintenance Supervisor H. D.. Elkins, Instrument Maintenance Group Manager J. T. Crittenden, Public Safety Service Chief R. W. Fortenberry, Technical Support Supervisor
  • G. B. Kirk, Compliance Supervisor
  • D. C. Craven, Quality Assurance Staff Supervisor
  • J. H. Sullivan, Regulatory Engineering Supervisor J. L. Hamilton, Quality Engineering Manager D. L. Cowart, Quality Engineering Supervisor
  • H. R Rogers,- Plant Operations Review Staff R. C. Burchell, Compliance Engineer
  • R. H. Buchholz, Sequoyah Site Representative
  • M. A. Cooper, Compliance Engineer
  • 0 L. Widner, ECN Closure Group
  • M. R. Sedlacik, Modifications Group A Supervisor
  • J. A. Naik, Modifications Supervisor
  • G. G. Wilson, Assistant Operations Supervisor
  • M. J. Blankenship, Information Services Manager
  • G. S. Boles, Mechanical Maintenance, 0&MM Supervisor
  • J. Robinson, Assistant Modifications Manager
  • W. S. Wilburn, Assistant to the Maintenance Superintendent
  • C. G. Hudson, Radiation Health Chief

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  • F. W. Reimann, Radiological Assessor i *R. M. Sexton, Quality Assurance Evaluator
  • T. Smith, Electrical Engineer
  • B. Schofield, Licensing Engineer
  • J. McQualls, Radwaste Section Supervisor
  • R. O. Barnett, Civil Engineering Chief Other licensee employees contacted included technicians, operators, shift engineers, security force members, engineers and maintenance personne . -. .- , , - - - - . , ..- - -..- --- - - . . - _ -

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Other NRC personnel

  • C. R. Stahle, Project Manager
  • T. F. McElhinney, Reactor Engineer
  • A. B.-Ruff, Reactor Engineer
  • R. G. Weddington, Senior Radiation Specialist
  • C. H. Bassett, Radiation Specialist
  • Attended exit interview Exit Interview The inspection scope and findings were sumnarized with the Plant Manager and members of his staff on October 5, 1986. No violations or deviations were discussed, concerning this repor The licensee did not identi fy as proprietary any of the material reviewed by the inspectors during this inspection. During the reporting period, frequent discussions were held with the Site Director, Plant Manager and other managers. At no time during the inspection was written material provided to the licensee by the inspector. Violation 327,328/86-49-01 was discussed with the plant manager by the NRC section chief via telecon on November 14, 198 . Licensee Action on Previous Inspection Findings (92702) (Closed) Unresolved Item 327,328/86-46-05. This unresolved item was closed after the inspectors determined that the the licensee's correc-tive actions described in Licensee Event Report (LER) 327-85-040 and LER 328-83-101 were adequat This determination was made after additional information concerning the loss of Recidual Heat Removal events described in the LERs was reviewe The NRC responded on October 23, 1986, to TVA's letter (00mer/ Grace)

of August 19, 1986, and agreed with TVA that the portion of Violation 86-19-01 concerning upper head injection (VHI) hydraulic lock release valves, was inappropriately written against configuration control and withdrew that portion of the violatio The hydraulic lock release valves are used to adjust the stroke time of the UHI system isolation valves. In researching TVA's response to 86-19-01, the NRC discovered that Sequoyah has had a repeated history of failed UHI isolation valve response time surveillances (20 of 24 failures between 1981 and 1985). Although physical security of the hydraulic lock release valves (as discussed in the response to Violation 86-19-01) may not be a problem, it is apparent that their throttle positions are changing due to system vibration during UHI isolation valve response time testin Subsequent to Violation 86-19-01, the NRC identified that the Sequoyah UHI hydraulic lock release valves are equipped with a set screw type mechanism that is used at other plants of the same design to prevent inadvertent stem movement. Sequoyah was unaware of the existence of these set screws until notified by the NR ,

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The cause of the repeated UHI isolation valve surveillance failures were not adequately determined by Sequoyah and corrective action was not taken to preclude further failure This constitutes a violation of 10 CFR 50 Appendix B, Criterion XVI,- the Nuclear Quality Assurance Manual, and Administrative Instruction (AI)-12, and is identified as Violation 327, 328/86-49-0 "

4. Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations. One unresolved item was identified during this inspection (paragraph 11).

5. Operational Safety Verification (71707) Plant Tours The inspectors observed control room operations, reviewed applicable logs, conducted discussions with control room operators, observed shift turnovers, and confirmed operability of instrumentatio The inspectors verified the operability of selected emergency systems, and verified compliance with Technical Specification (TS) Limiting Conditions for Operation (LCO). The inspectors verified that maintenance work orders had been submitted as required and that followup activities and prioritization of work was accomplished by the license Tours of the diesel generator, auxilia ry, cor. trol , and turbine buildings, and containment were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and plant housekeeping / cleanliness condition The inspectors walked down accessible portions of the following safety-related system on Unit 1 and Unit 2 to verify operability and proper valve alignment:

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Containment Spray System No violations or deviations were identifie Safeguards Inspection In the course of the monthly activities, the inspectors included a review of the licensee's physical security program. The performance of various shifts of the security force was observed in the conduct of daily activities including protected and vital area access controls; searching of personnel and packages; escorting of visitors; badge issuance and retrieval; and patrols and compensatory post In addition, the inspectors observed protected area lighting, and protected and vital areas barrier integrity. The inspectors also interviewed security personnel regarding their respective dutie '

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4 Radiation Protection The inspectors observed Health Physics (HP) practices and verified implementation of radiation protection control. On a regular basis, radiation work permits (RWPs) were reviewed and specific work activities were monitored to assure the activities were being conducted in accordance with applicable RWPs. Selected radiation protection instruments were verified operable and calibration frequencies were reviewe No violations or deviations were identifie . Monthly Surveillance Observations (61726) The inspectors reviewed Technical Specification (TS) required

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surveillance testing and verified that testing was performed in-accordance with adequate procedures; that test instrumentation was calibrated; that Limiting Conditions for Operation were met; that test results met acceptance criteria and were reviewed by personnel other than the individual directing the test; that deficiencies were identified, as appropriate; that any deficiencies identified during the testing were properly reviewed and resolved by management personnel; and that system restoration was adequat For complete tests, the inspector verified that- testing frequencies were met and tests were performed by qualified individual The following surveillance instructions (sis) were reviewed:

SI-162.2, Snubber Functional Testing SI-129, Emergency Core Cooling Safety Injection Pump Operability SI-130.1, Turbine Driven Auxiliary Feedwater Pumps SI-130.2, Motor Driven Auxiliary Feedwater Pumps IMI-99.CC 12.2, RCS Cold Overpressure Protection System Verification SI-682, ERCW Flow Balance Valve Position The licensee identified, during a review of an internal commitment trach ng system, that TS surveillance requirements 4.7.9.e.3 and 4.7 .f may not have been complied with during the implementation of SI-162.2. This apparently was not identified through the licensee's Surveillance Task Force review nor the present Surveillance Instruction (SI-1, Appendix F) review. The inspector reviewed this surveillance and the licensee's corrective action. Followup of this issue and other surveillance instruction deficiencies not identified during the licensee's surveillance review process will be followed as Inspector Followup Item 327,328/86-49-0 SI-130.2, SI-130.1 and SI-129 were reviewed by the inspector. The licensee identified through the SI-1, Appendix F Surveillance i Instruction Review Program that certain pump parameters may not have been maintained during pump testin This issue along with several pump testing issues will be followed as Inspector Followup Item 327,328/86-49-03, t

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5 Instrument Maintenance Instruction (IMI)-99.CC 12.2, RCS Cold'

Overpressure Protection System Verification, was reviewed by the inspector. On August 4,1986, instrument maintenance personnel were performing the above procedure. During this performance, Power Operated Relief Valve (PORV) 1-PCV-68-334 opened inadvertentl This resulted from an inadequate procedure change which was a result of certain procedural steps being omitted during the typing and administrative review process. Issues involving administrative errors in technical documents were identified in Inspection Report 327,328/86-20. The issue involving the adequacy of Plant Operations Review Committee review of procedural changes was discussed in Inspec-tion Report 327,328/86-46. The inadequacy of this portion of IMI-99 was discovered as a result of an event and not the SI-1, Appendix F, Suiteillance Instruction Review Program and will be included as one of the items to be reviewed under Inspector Followup. Item 327, 328/86-49-0 No violations or deviations were identifie . Monthly Maintenance Observations (62703)

Station maintenance activities of safety-related systems and components were reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in conformance with T The following items were considered during this review: LCOs met while components or systems were removed from service; redundant components operable; approvals obtained prior to initiating the work; activities accomplished using approved procedures and inspected as applicable; procedures adequate to control the activity; troubleshooting activities controlled and the repair record accurately reflected what actually took *

place; functional testing and/or calibrations performed prior to returning components or systems to service; quality control records maintained; activities accomplished by qualified personnel; parts and materials used properly certified; radiological controls implemented; Quality Control (QC)

hold points established and observed where required; fire prevention con-trols implemented; and housekeeping actively pursued. The following maintenance related work requests (WRs), work plans (WPs) and engineering change notices (ECNs) were reviewed:

WR B202504 WR A546226 ECN 5009 ECN 5045 WP 9653 WP 10431 WP 9652 WP 9623 No violations or deviations were identifie l

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8. Licensee Event Report (LER) Followup (92700)

The following LERs were reviewe The inspector verified that: reporting requirements had been met; causes had been identified; corrective actions appeared appropriate; generic applicability had been considered; the LER forms were complete; the licensee had reviewed the event; no unreviewed safety questions were involved; and no violations of regulations or Technical Specification conditions had been identifie LERs Unit 1 .

LER 86-036 (Closed) Plant Operation Review Committee Quorum LER 85-040 (Closed) Loss of Residual Heat Removal During Pump Swapover LER 84-048 (0 pen) Lack of Required Number of Post Accident Monitoring (PAM)

Instruments LERs Unit 2 LER 83-101 (Closed) RHR Pump Cavitation During Pump-down of Refueling Cavity 9. Event Followup (93702, 62703) During the inspection period, two separate inadvertent starts of the emergency diesel generators (EDGs) occurred:

(1) The first event on September 14 was initiated by plant personnel attempting to replace KAZ indicator fuses with MIS-5 indicator fuses in the 125 volt battery board IV panel per WP 12152. When the operator pulled fuses in a circuit labeled as a " spare", the EDGs starte Investigation by plant personnel revealed that a modification had been performed to replace the " spare" circuit with the power supply for the 2B-B EDG logic circuit. When this logic circuit is deenergized, all four EDGs are started by their respective emergency start relays. When the engineer in charge of the modification made changes to the control room prints, he made t changes only to those prints in the Unit 2 control room because

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the 28-B EDG is administratively " assigned" to Unit 2. When the operator was asked to pull the KAZ fuses and replace them with new MIS-5 type fuses, he consulted Unit 1 prints. The battery board involved is administratively " assigned" to Unit The Unit 1 prints, however, had not been marked to reflect the modificatio Sequoyah has had problems in this area since the time frame when Unit I was operating and Unit 2 was still under construction. The decision was made by the licensee to maintain separate prints for i Unit 1 and Unit 2, even when the two print series showed the same equipment. Differences between the prints began to emerge when plant modifications were reflected only on the prints for the unit to which the modified equipment was " assigns '

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Modifications, Engineering, and Operations personnel indicated a lack of consensus on how the prints are marked, corrected and identified when modifications are made. Plant procedures (AI-19)

only require marking the " applicable" control room's prints after modifications are made. This practice appears to cause confusion as to which set of prints is correct. This drawing control issue-will be tracked as IFI 327, 328/86-49-0 Approximately 20 minutes after the operable EDGs started, the IB-B EDG tripped on high jacket water temperature. This trip was caused by a configuration control error. The normal cooling water supply valve, 1-FCV-67-82, controlling the Essential Raw Cooling Water (ERCW) flow to the 1B-B EDG had been tagged out of service by Hold Order (HO) 1694. When the normal cooling water valve is not available, the alternate cooling water valve must be manually opened from the control room by the operator when the EDG start This common practice is supposed to be prompted by a caution tag

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which is hung on the applicable EDG's start switch on the control pane This caution tag should be hung whenever the normal cooling water supply valve is unavailable. In this instance, the shutdown reactor operator (SRO) who approved the tag-out of the normal supply valve forgot to hang the caution ta Of oarticular interest in this event is that a valid emergency start, coincident with a blackout would most likely have resulted in severe damage to the 18-B EDG. This postulated emergency start would have immediately loaded the EDGs, reducing the time when high temperatures would have occurred. With an emergency start, however, the high jacket water temperature trip would not trip and protect the unit. This trip and several others are bypassed under emergency start condition Instrumentation to alert the operators to a loss of cooling water does not exist in the control room. Consequently, the operators would have received no advance

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warning of this degraded condition until damage and possibly destruction of the EDG had occurred. When an EDG start occurs, the operator in the control room is required to verify that the normal cooling water valve strakes open. This was verified during the event, but in this instance a valve had been shut upstream of the supply valve. The operator could not determine if the diesel was receiving adequate cooling water. This appar.snt d~'iciency of adequate instrumentation to monitor the emergency a.uz 1 genera-tors will be tracked as IFI 327, 328/86-49-0 (2) The second inadvertent start of the EDGs occurred on October 2, 1986, while Modifications personnel were implementing WP 12090, to bypass the torque switches in the EDG's ERCW valves. While working in the valve operator's motor control section, one of the workmen accidentally shorted across contacts that caused a fuse to blow in the IA-A EDG's remote control circui Loss of this circuit caused the emergency start relay ESIAY1 to be deenergized, starting all operable EDGs. The blown fuse was a Bussman MIS-5

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(the type that had replaced a Bussman KAZ as discussed above).

While reviewing this incident with the schematic diagrams (45N767-5), it was noted that the blown fuse was of the same rating as the fuses installed in the four separate branch circuits supplied by the fuse that had blown. The Plant Manager explained that engineering design personnel had reviewed this apparent anomaly and determined that no design problem existed. It is the licensee's position that the fuses installed in the branch circuits were not designed for circuit protection, but rather for ease of branch isolation by pulling the applicable fuses. In effect, the fuses serve as disconnects for their branch circuit Additionally, when the EDGs started, control room operators attempted to open the s1 ternate ERCW supply valve, 1-FCV-67-68, to supply cooling water for the IA-A diese This was necessary because the normal supply valve, 1-FCV-67-66 was tagged out per WP 12090. The operator was unable to open the alternate cooling water supply valve because the control circuit for the valve is one of the branch circuits supplied by the blown fus An operator was immediately dispatched to the diesel building where he was able to manually open the alternate cooling water valve, 1-FCV-67-6 On October 6, 1986, the Site Director signed a stop work memo requiring all work on TVA Class A, B, and C/D pressure retaining piping components to have his direct approval. This action came following a preliminary report on an Employee Concern in the area of procuremen The perceived problem is a lack of credibility of the methods used in the Construction Heat Number Sort Program and Nuclear Power, Power Storeroom Requisition for verification of properly certified pressure boundary materials at installatio TVA is required to be able to trace the piping material to the certified mill test report that attests to its suitability for the' application for which it is installed. This issue is still under review by the plant and is being followed by the resident inspector No violations or deviations were identifie . IE Information Notices (92701)

The following IE Information Notices (IEN) were reviewed and closed. The inspector verified that: corrective actions appeared appropriate; generic applicability had been considered; the licensee had reviewed the event and that appropriate plant personnel were knowledgeable; no unreviewed safety

. questions were involved; and that violations of regulations or Technical Specification conditions did not appear to occur.

IEN 86-75 Incorrect Maintenance Procedure on Traversing Incore Probe Lines IEN 86-79 Degradation or Loss of Charging Systems I

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11. License Conditions (93702)

A review of license conditions was conducted, and three situations exist:

Certain license conditions may not have been met because of a failure by the licensee to perform a certain actio Cc-tain license conditions could not be closed-or were not met because no NRC response could be verified at the time of the inspectio Certain license conditions are still under review because of ongoing action by the license License conditions in the Appendix R and Equipment Qualification areas were not inspected during this revie License Conditions that may not have been met because of a failure by the licensee to perform a certain action:

Unit 1, license condition 2.C.(15); Unit 2, license condition 2.C.(12)

Diesel Generator Reliability - The required dew point modification and ~ ,

test were not completed u itil 198 The ECN was outstanding from January 19,1983 (when the functional test was performed) until 1985 when the ECN was closed, resulting in the licensee missing the implementation date for the modification Technical questions exist on the unresolved safety question determina-tion (USQD) for the turbocharger modification Unit 2, license condition 2.C.(16).s Primary Coolant Outside Containment - The licensee failed to meet the thirty day reporting requirement. This was discussed in inspection report 327,328/81-5 License Conditions that may not have been met or could not be closed because no NRC response could be verified at the time of the inspec-tion:

Unit 1, license condition 2.C.(6) Seismic Design Margin review -

Submittals were made by the licensee May 5,1981, and March 1,198 No NRC response could be verifie Unit 1, license condition 2.C.(18) Instrumentation - Tha downscale failure alarms were installed for the effluent monitor:19 instrumen-tation channels for radioactive gaseous and radioactive liquid effluent Operability needs to be verified. Section (b) of this license condition required the licensee to submit reactor protection system (RPS) and engineering safety feature (ESF) values for evalua-tion. No NRC response could be verifie Unit 1, license condition 2.C.(21) Control Rod Guide Thimble - The licensee submitted details of the inspection program for NRC approva No NRC response could be verifie . . . _ . .

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Unit 1,. license condition 2.C(23).h Inadequate Core Cooling Instrumentation - The licensee installed the required modifications on January 18 vice January 1, 1982. An agreement between NRC and the licensee allowed the license condition date of January 1,1982, to be

. extended. The conversation was verified in Inspection Report 327,

'328/82-0 Unit 2, license condition 2.C.(16)_.d Control Room Design - An erro was identified nn January 7, 1983 and was resolved over the phone with NRR and Region II management. The issue is described in Inspection Report 327,328/83-0 Unit 2, license condition 2.C.(16).n Voiding in the Reactor Coolant System - The Westinghouse Owners' Group x position was referenced in a December 18, 1981 letter to NRR. No NRC response could be verifie License Conditions still.under '

review because of ongoing action by the licensee:

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Unit 1, license condition 2.C.(11) Negative Pressure in the Auxiliary Building - The status of the Auxiliary Building Secondary Containment Enclosure (ABSCE) has not been determined by the inspector because of the licensee initiated plant modification A current Surveillance Instruction, S1 - 149 is being reviewed to determine if this surveil-lance ensures'that a negative pressure is maintained in the ABSC Unit ri lic.ense condition 2.C.(12) Environmentc1 Qualification Unit 1 ficense condition 2.C.(16) Appendix R, Fire Protection Unit 1 license condition 2.C.(22); Unit 2 license condition 2.C.(16).d Control Room Design - Thjs issue is prescribed by an NRC order of January 15, 1984, and i.s an ongoing issu i Unit 1 license condition 2.C.(23) Reactor Coolant System Vents - The licensee installed reactor coolant system and reactor vessel head highpoint vent However, modifications are required to prevent

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leakage problems that have been encountered.

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( t Determination' if license conditions, discussed in paragraphs and above, were satisfied is identified as Unresolved Item 327,328/86-49-07, pending further NRC revie %

12. -Inspektor Followup Items (92701) NRC letter G. Zech to S. White dated July 10, 1986, referenced Inspection Report 328/84-18. The subject report adoressed a potential safety issue involving a single failure of the steam supply transfer feature. The letter requested the licensea to " evaluate the benefit of the design feature to overall piant safety and consider its possible deletion."

Theflicensee conducted a review of the turbine driven auxiliary feedwater pump steam supply transfer feature and determined that no

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-11 modification should be performed. The inspector reviewed a memo from P. Wallace to H. Abercrombie . dated September 10, 1986 and discussed this issue with. plant engineers. The licensee's action appears to be adequate and this issue is considered closed, b. The inspector viewed vertical' cable' trays in the control building. The issue of static and seismic loading was discussed with Region II perscnnel, and will be followed as Inspector Followup Item 327, 328/86-49-06.

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