IR 05000267/1986032
| ML20210D672 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 01/23/1987 |
| From: | Farrell R, Jaudon J, Michuad P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20210D627 | List: |
| References | |
| 50-267-86-32, NUDOCS 8702100140 | |
| Download: ML20210D672 (12) | |
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APPENDIX B U.S. NUCLEAR REGULATORY C0994ISSION
REGION IV
NRC Inspection Report: 50-267/86-32 License: DPR-34
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Docket: 50-267 Licensee:
Public Service Company of Colorado (PSC)
P. O. Box 840 Denver, Coloardo 80201 Facility Name: Fort St. Vrain Nuclear Generating Station Inspection At:
Fort St. Vrain (FSV) Nuclear Generating Station, Platteville, Colorado ~
Inspection Conducted: November 16 through December 31, 1986
//23/s y Inspectors:p ' R. E. Farrell, Senior Resident Inspector (SRI)
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W~. Michuad, Re ident In tor (RI)
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IN Approved:
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Q:f7.[Jau n. Chief, Project Section A Ryactor P ojects Branch Vg Inspection Sumary Inspection Conducted November 16 through December 31, 1986 (Report 50-267/86-32)
Areas Inspected: Routine, unannounced inspection of operational safety
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verification, licensee action on previously identified inspection findings, licensee action on licensee event reports (LERs), maintenance, surveillances, reserve shutdown system, quality assurance, and security.
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Results: Within the areas inspected, one violation was identified (failure to wear required identification, paragraph 2).
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Persons Contacted
- F. Borst, Manager, Support Services / Radiation Protection
- L. Brey, Manager, Nuclear Licensing and Fuels
- R. Craun, Manager, Nuclear Site Engineering
- D. Evans,' Superintendent, Operations
- M. Ferris, Manager,-QA Operations
- J. Gahm, Manager, Nuclear Production
- D. Goss, Coordinator, Nuclear Licensing and Fuels
- J. :Gramling, Supervisor, Nuclear Licensing Operations
- M. Holmes, Manager, Nuclear Licensing.
- F. Novachek, Manager, Technical / Administrative Services
- K. Owens, Commitment Analyst
- P. Tomlinson,- Manager, QA-
A. Vigil, Reactor Operator l
R. Walker, Chairman of the Board and CEO
- D. Warembourg, Manager, Nuclear Engineering
- R. Williams' Jr., Vice President, Nuclear Operations The NRC. inspectors also contacted other licensee and contractor personnel during the inspection.
- Denotes those attending the exit interview conducted January 6, 1986.
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2.
Operational Safety Verification
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-j The SRI and RI reviewed licensee activities to ascertain that the facility-is being operated safely and in conformance with regulatory requirements and that the-licensee's management control system is effectively discharging its responsibilities for continued safe operation.
The review was conducted by direct observation of activities, tours.of the facility, interviews and discussions with licensee personnel, independent verifications of safety system status and limiting conditions for-operation, and review of facility records.
Logs and records reviewed included:
Shift supervisor logs
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Reactor operator logs
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Equipment operator logs
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Auxiliary operator-logs
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Technical specification compliance logs
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Operations order book
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Operations deviations reports
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Clearance log
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-3-Temporary configuration reports
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Station service requests (SSR)
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During tours of accessible areas, particular attention was directed to the following:
Monitoring instrumentation
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Radiation controls
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Housekeeping
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Fluid leaks
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Piping vibrations
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Hanger / seismic restraints
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Clearance tags
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Fire hazards
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The NRC SRI observed that functional testing to restore systems modified during the current outage required several teams of technicians to be in the control room. The licensee was cautioned by the SRI that nonlicensed persons manipulating controls must be under the direct supervision of a licensed operator. Additionally, all activities in the control room must be under the supervision and control of the licensed reactor operators.
Activity levels and operator control of activities in the control room
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the current outage.-
On December 22, 1986, the NRC SRI, was in the control room observing reactor operator and technician activities when he observed a reactor operator not wearing a picture identification badge. The reactor operator, when questioned by the NRC SRI-explained that he did not wear his picture badge in the control room since everyone knew him and he was not going anywhere. The reactor operator additionally explained leaving the picture badge on his coat was a good way for him to remember to take his coat with him when he left the control room. Not wearing a picture identification badge within the protected area is an apparent violation.
(267/8632-01)
3.
Licensee Action on Previous Identified Inspection Findings
.(CLOSED)OpenItem(267/8507-10): Revisions to Procedures ESR 8.1.2bcd-M and SOP 62 (CAR 85-107). This open item was to ensure corrective actions associated with LER 85-04 included revising these procedures to ensure a proper lineup of demineralizers during recirculation through liquid effluent monitors by checking shut V6241 and HV6212. These actions have been completed. This item is closed.
(CLOSED) Open Item (267/8414-07):
Revision to Procedure APM Q4 (CAR 84-065).
Deficiencies in Issue 4 have been corrected and incorporated into a completely rewritten APM Q4, which is currently Issue 9.
This item is closed.
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-4-83-022)). Open Item (267/8301-01): QA Purchasing Program (CAR-83-038 and (CLOSED r
The QA purchasing program allowed the purchase of quality related material under a nonquality related purchase order. Procedure APM-Q4 has been revised to correct the discrepancies. This item is closed.
(CLOSED) Unresolved Item (267/8312-03): DCCF Preparation Not Following Procedure G2 Format (CAR-83-084). The four DCCFs used to revise Procedure G2 have been corrected.
Procedure AFM-G2 and all S0Ps are now in G2 format. This item is closed.
No violations or deviations were identified in this inspection area.
4.
Licensee Action on Licensee Event Reports (LERs)
The RI reviewed licensee event reporting activities to verify that they were in accordance with Technical Specifications Section 7, including identification details, corrective action review and evaluation of aspects relative to operations, and accuracy of reporting.
The following LERs were reviewed and remain open:
81-072 Rupture Disk Out-of-Tolerance 84-010 Continuou:: Semnler Inoperable During Greater Than 10 GPM Release 85-001 Neutron Flux Rate of Change High Scram 85-008 Reactor Scram on High Count Rate 86-024 PCRV Rupture Disc Found Out-of-Tolerance 86-025 Reactor Building Sump Release Not Continuously Sampled 86-027 Reactor Building Sump Release Not Continuously Sampled 86-028 Reactor Scram Actuations on Neutron Flux Rate of Change High LERs81-072 and 86-024 both deal with the high set PCRV rupture disc being found out of tolerance during surveillance testing. Corrective action for both occurrences, in part, consists of an evaluation of the procedure used to obtain the as-found setpoint. Since the high set rupture disc was found out-of-tolerance on both previous as-found tests, these LERs will remain open until the evaluations conunitted to in the above LERs are submitted to the NRC.
LERs84-010, 86-025, and 86-027 all deal with not continuously sampling reactor building sump discharges of greater than 10 gpm as required by ELCO8.1.3(b). These LERs will remain open and will be followed up in greater detail in a subsequent report.
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-5-i LERs85-001, 85-008, and 86-028 all report spurious high neutron flux rate of change scram actuations while shutdown. Current troubleshooting efforts are described in the maintenance section of this report. These LERs will remain open until the troubleshooting is complete.
Corrective action for the following LERs was reviewed for completeness and implementation. Based on this review, the following LERs are closed:
80-049 Rupture Disk Out-of-Tolerance 84-009 Liquid Waste Release Exceeded MPC For Unidentified Beta 84-012 Only 1/2 of Total RSD Material Was Discharged From CRD No. 21 84-013 Both Area Radiation Monitors in Group Two, Hot Service Facility Inoperable 85-004 Unplanned Liquid Waste Release While Recirculating Through Activity Monitors No violations or' deviations were identified in this inspection area.
5.
Maintenance The RI continued to monitor EQ related maintenance activities and witnessed installation of RTV foam insulation in conduit terminations at junction boxes. The RI reviewed Change Notice (CN) 2335 and Controlled Work Procedure (CWP) 86-0253, " Install Drain Holes and Sealant in Conduit and Install Drain Holes in Junction Boxes as Required." Also reviewed was the QC inspection report which verified shelf life, mixing proportions and time, sample comparison to standard chart for color and cell structure, and CERA blanket installation of the foam insulation. All portions of the process were observed for several conduits. The RI also observed installation of a RAYCHEM 3-Phase Nuclear Motor Connector Kit on Reactor Plant Cooling Water Pump P4602, and reviewed the associated paperwork; MPE1901.6, Issue 2.
The RI observed the replacement of valve HV-6302, gas waste surge tank inlet, which included rerouting of one pipe, and the reassembly of HV-2337, the emergency condensate to EES isolation valve.
New check valves and o-rings were installed in equipment storage well and fuel storage well covers in accordance with Station Service Request (SSR) 86512367-02.
Corrective Maintenance of nuclear instrumentation equipment in response to high rate of change scram signals was monitored by the RI. The RI interviewed technicians performing the troubleshooting to familiarize himself with and assess their approach to the problem. The design of the instrumentation system at FSV includes a common station ground rather than an independent ground for each system. This configuration allows
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-6-perturbations on the station ground from other sources to affect the common or reference level of the instrumentation. When new equipment is
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One of the two main objectives of the troubleshooting effort is to identify and to separate'as much as possible the common bus electrical noise on the instruments from other sources.
The other objective of the current troubleshooting is to identify and to i
. suppress noise sources such as relays, switches, and inductors. When these components operate, a small electromagnetic pulse is generated.
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this pulse is sufficiently large, it causes a noise spike on the nuclear
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instrumentation system.
The approach was, therefore, to monitor the nuclear instrument channels while exercising each item (switch, relay, alarm, etc.) which could cause a noise spike.
Any of those items which caused a noise spike was suppressed by installation of an approved shunt type suppressor.
The RI performed a walkdown of changes to the diesel generator undervoltage relay circuits done to meet single failure criteria.
CN-1622F i.s a temporary modification to supply power to the undervoltage relays from the associated DC power supply circuit used for control of the 480 VAC switchgear.
These DC control power circuits are on an automatic-throw-over (ATO) scheme which will switch control power between the two redundant'DC buses in the event one of the buses fails.
The RI reviewed the associated documentation and inspected the installed hardware.
At the end of this inspection period, the final tie-ins of the temporary DC power supplies at the 480 V switchgear cabinets had not been made. All other aspects of the installation appeared satisfactory and the RI will monitor the final tie-ins which will complete this modification.
Also associated with the emergency electrical power system meeting single failure criteria was the removal of common relays DEV-86RT and DEV-86ET1 from the emergency diesel generator (EDG) auto start circuits.
The original plant design utilized these relays to start the EDGs, then remain in standby as a precautionary measure in the event of electrical transients or loss of major equipment.
This function is now performed by the Class IE undervoltage relaying scheme on the 480 VAC essential buses.
Thus the contacts of the common relays are unnecessary and their removal will prevent unnecessary challenges to the EDG which could have the potential for degrading systems and components important to safety. The RI reviewed CN 2396 which removed these relays and found this had been accomplished satisfactorily, and did not present any unreviewed safety i
Concerns.
No violations or deviations were identified in this inspection area.
6.
Surveillances The NRC RI observed portions of the test discharge and subsequent equalizing charge on both 1A and 1B station batteries.
It was noted that
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t-7-the automatic discharge control console, which was recently. purchased, provides precise control and recording features which are an improvement over a manually controlled system.
The NRC SRI observed portions of " Functional Test of 'B' Circulator
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Instrumentation," SR-RE87-X (11-10-86), being performed. Quality control personnel were present. An alarm card FAllL-2185 failed in step 5.4.7 of the procedure causing FI-2185, the bearing water " flow high" alarm, to come in at 180 gallons per minute rather than at 200 gallons per minute.
A station service request was generated to replace the faulty card.
The NRC SRI reviewed:
MAP-2, Issue 1, " Maintenance Department Program for Calibrated
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Tools,TestEquipment,andStandards(CTTE&S)"
RP-A-04, Issue 3, " Requirements Governing the Control and
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Calibration of Test Equipment and Standards" MAP-2 is the maintenance department program for control of test and measuring equipment. RP-A-04 is the results department procedure for controlling the same type of equipment.
Procedure MAP-2 defines a +/-25%
grace period on equipment calibration intervals, but identifies that there is no minimum interval for recalibration and that the interval should be considered a maximum interval between calibrations with the grace period utilized only when the test equipment in question is being utilized in an--
on going test. Procedure RP-A-04 simply identifies a +/-25% grace period on the calibration interval. The SRI observed results department instrument checkout practices and noted that the grace period on test and measuring equipment calibration intervals is comonly utilized to extend the effective calibration of the equipment. The SRI cautioned the licensee that, while this practice is acceptable in a regulatory sense, it does leave the licensee at risk of greater rework should an instrument come back from recalibration with an unacceptable as-found condition.
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No violations or deviations were identified in this inspection area.
7.
Reserve Shutdown (RSD) System The RI reviewed surveillance procedures, operating procedures, records of tests and procedures performed, and inspected the RSD system to determine that the licensee conformed to Technical Specifications, the FSAR, and approved procedures.
The surveillance procedures were reviewed to determine whether tho procedures:
Require application of test pressure to the RSD hoppers in accordance
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with the Technical Specifications and the FSAR Identify appropriate initial conditions
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. Include acceptanceLcriteria relating to' performance of the actuating's
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valves, integrity of the rupture disc, and performance of the' orifice -
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Identify proper valve lineup to each hopper for returning'the system
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to normal and re' quire verification of the valve lineup.
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Re' quire;periodiccallibratiSn-ofthepressureindicators
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Require testing lwo hoppers and rupture discs, one hopper containing,
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20 wt-% and'one containing~40 wt-% boronated material, in accordance with requirements in Technical Specifications:and the FSAR Specify appropriate handling requirements, test conditions, and
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acceptance ~ criteria Require insp'ection of the hopper after the rupture. disc ~ ruptures to.
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determine whether all the balls were released from the hopper
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Identify appropriate specifications for the replacement rupture disc
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and requirements for inspection of the replacement rupture disc prior
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to and following installation Specify requirements for inspection of the balls before returning
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them to the hopper or for the characteristics of replacement balls.
The following currently issued surveillances were reviewed for conformance to the Technical specification requirements:
SR 4.1.8.A/8-W / 4.1.9.A/B-W, " Reserve Shutdown Helium and Nitrogen-
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Bottle Pressure Check," Issue 2,; dated March 7, 1986
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SR 4.1.8.C.1/2/3-A, f' Reserve Shutdown Hopper, ACM Disconnect, and Low -
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Pressure Alarm Test,". Issue 2, dated December 6, 1985 SR 4.1.8.D-A/4.1.9.C-A, " Channel Calibratio'n of the Reserve Shutdown
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Gas Pressure Instrumentation," Issue 1, dated July 12, 1985
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SR 4.1.9.D.1-R, "Reserv'e Shutdown Valve Operability-Test," Issue 1,
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dated July 12, 1985 h
j SR 4.1.9.D 2-R, " Channel Calibration of Reserve Shutdown Hopper
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Pressure Switch," Issue 1, dated July 12, 1985.
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The RI noted that no procedures currently exist for Interim Technical l
Specification Surveillance Requirement 4.1.9.D.3 or 4.1.9.D.4.
Specifically, SR 4.1.9.D.3 requires a visual examination of the pipe
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sections which require disassembly and reassembly within the refueling l
penetrations at each refueling outage.
SR 4.1.9.D.4 requires a functional test of two reserve shutdown assemblies out of core, one containing 20
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weight percent boronated material and the.other containing 40 weight percent baronated material,' with a visual and chemical examination of the
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material from the tested. hoppers.
The licensee's regulatory connitment, log item 1014 requires these surveillance procedures to be in place by the fourth refueling-outage.
The~1icensee was informed this will be considered an open item (267/8632-02).
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The RI inspected documentation of. surveillance tests performed during the last year for confonnance of the ' tests and test results to requirements in the Technical Specifications, FSAR, and approved procedures.
No replacement rupture ' discs or absorber balls were procured during the last year. With the exception of electrical equipment qualification work, which will be covered under a separate inspection, no modifications to the
reserve shutdown' system were made in the last year.'
The RI performed a physical inspection of all accessible portions of the reserve shutdown system to determine whether:
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The volume of the individual hopper helium bottles conforms to the
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requirements in the FSAR
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The pressure on each individual hopper helium bottle exceeds the
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lower limit set forth in the Technical Specifications The valve lineup is correct in accordance with the 50P valve lineup
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checklist and in agreement with the P&I drawings.
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The individual helium bottles are isolated in accordance with Operations Deviation Report 3436 for the remainder of the current shutdown. The pressure on each helium bottle therefore could not be checked, but will be verified by the RI when the system is returned to service.
Within the scope of this inspection no violations or deviations were identified.
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8.
Quality Assurance The NRC SRI reviewed the quality assurance program during this inspection
interval to monitor licensee progress addressing SALP concerns. The review was conducted with emphasis on staff, corrective actions, and
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audits,
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a.
Staff There are currently 60 quality assurance personnel working full-time
for-the licensee QA staff, of these 60, 48 are dedicated to FSV.
Planned staff changes will take the quality assurance department's staff to 75 full-time people of which 62 will be dedicated to FSV.
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The audit group at FSV has consisted of one supervisor, two auditors, and.one trainee. Planned staff changes will cause the audit group to consist of a supervisor and eight auditors. Quality control currently consisting of two separate groups, maintenance QC and QA/QC, will be combined-under one ; superintendent of QC reporting to the QA services manager.
In addition to auditors, planned QA staff additions include four QC inspectors as well as additional clerical staff to expedite the processing of paperwork. The licensee is also actively recruiting a vendor qualification specialist. Other changes in the QA organization are anticipated as the division is modified to accomplish its goals more effectively.
b.
Corrective Action System Historically, the corrective action system at FSV has been ineffective and untimely. The licensee has received violations in past inspections for untimely or ineffective corrective action.
Currently, the corrective action system is getting a high degree of executive and senior management attention. Open corrective action requests are discussed weekly at a senior management planning meeting with overdue corrective action requests highlighted and~ identified by the responsible manager. The number of open corrective action items is being reduced and the average age of open corrective action requests is dropping. The governing procedure, Q-16, " Corrective Action System," has been recently revised; Issue 8 became effective on December 31, 1986.
Issue 8 of Q-16 clearly delineates responsibilities for responding to corrective action requests.
It has a very specific provision for escalating unresolved or overdue corrective action requests. Additionally, Procedure Q-16, Issue 8, provides an additional mechanism for corrective actions, the quality deficiency report, which allows corrective actions requiring lower levels of management attention to be addressed formally without involving senior management. This provides a mechanism for resolving items of less significance, adding greater emphasis to corrective action requests when they are generated. The licensee already plans to again revise Q-16 making the procedure reflect its title,
" Corrective Action System." The planned revision will incorporate all mechanisms of obtaining corrective action, including station service requests and nonconformance reports. The current issue was produced on an expedited basis and is an improvement over Issue 7,.
allowing the licensee to improve the working of the corrective action system while fine tuning the procedure.
At the weekly senior management meeting the QA department presents a corrective action request (CAR) progress report which trends the
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total backlog of corrective action requests open, trends the response
times of the various nuclear divisions, and trends the average age of open corrective actions company wide and by division.
Prior to September 26, 1986, there were no weekly CAR progress reports at FSV.
Review of CAR progress reports from September 26, 1986, through
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December 10, 1986, revealed progressively improving reports starting
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c.
Audits The audit: staff, as identified earlier in this section, is being increased from one supervisor, two auditors, and one auditor trainee to'a supervisor and eight auditors which is consistent with audit group staffing -levels at other commercial nuclear power plants.
The licensee is placing emphasis on experience and expertise, endeavoring to have an audit group with a wide spectrum of engineering expertise to conduct audits which address technical adequacy as well as procedural adherence.
The NRC SRI reviewed the first issue of the 1987-88 audit schedule and interviewed the quality assurance division manager regarding the audit-schedule.
The first issue of the audit schedule is being revised to accomplish manpower loading as well as better audit coverage to emphasize audits in known problem areas, previous audit findings,.and NRC inspection findings, as well as Technical Specification required audits.
The licensee was issued two violations in the previous SALP inspection period for inadequate audits.
Senior management attention is being directed to the audit program to improve the quality of audits and utilize the audit program as an effective management tool to achieve regulatory goals.
No violations or deviations were identified in this inspection area.
9.
. Security During this inspection period, the licensee completed installation work on new perimeter detection aids and continued work on new perimeter lighting systems.
Checkout testing of the new perimeter detection aids has proceeded.
No violations of deviations were identified in this inspection area.
10.
Management and Exit Meetings During the inspection period, the NRC inspectors conducted a monthly meeting with the licensee executive management to review progress on activities to correct SALP report identified weaknesses and progress on environmental qualification of equipment.
Several meetings were held with the plant manager, the site engineering manager, and manager of quality assurance to monitor outage progress, performance enhancement program
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activities, and preparations for restart of the unit.
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An exit interview'was conducted January 6, 1987, attended by those
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indicated in paragraph 1.
At this time the inspectors reviewed the scope and findings of the inspection.
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