IR 05000267/1987008
| ML20215H324 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 04/06/1987 |
| From: | Farrell R, Jaudon J, Michaud P, Skow M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20215H296 | List: |
| References | |
| 50-267-87-08, IEB-86-002, NUDOCS 8704200263 | |
| Download: ML20215H324 (12) | |
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i.i, .. ' , 1_ . ?: APPENDIX B U. S. NUCLEAR-REGULATORY COMfiISSION ,,
REGION IV
NRC Inspection' Report: 50-267/87-08 License: DPR-34 Docket: -50-267 Licensee: Public Service Company of Colorado (PSC) Facility Name: Fort St. Vrain Nuclear Generating Station Inspection At: Fort St. Vrain (FSV) Nuclear Generating Station, Platteville, Colorado and PSC Offices, Denver, Colorado Inspection Conducted: March 1-31, 1987 '[[ ~ Inspectors: - 'R. E. Ft(rrell, Senior Resident Inspector (SRI) Date br 3.2/-7/ P. W. Michaud, Resident spector (RI) Date _ d%.a bO I , _.
{.(. Sow,ProjectEngineer, Project Section A Date Reac or Pro' cts Branch q h/[7 Approved: N_M J. f audon, Chief, Project Section A Date . L Rka or Projects Branch ENCLOSURE CONTAINS 13. Q 42 hh h fS@OO267 UPON SEPARATION THIS PDR PAGE IS DECONTROLLED.
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Inspection Summary ~ Inspection Conducted March 1-31, 1987 (Report 50-267/87-08) Areas inspected: Routine, unannounced inspection of operational safety verification, licensee action on previously identified inspection findings, licensee action on IE Bulletins, maintenance, surveillance, engineered safety features, design control, and security.
Results: Within the eight areas inspected, two violations were identified (improper calibration, paragraph 3; and inadequate review of plant change, paragraph 8). One unresolved item is identified in paragraph 9 and in the safeguards attachment to Appendix B.
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DETAILS 1.
Persons Contacted Principal Licensee Employees D. Alps, Supervisor, Security F. Borst, Manager, Support Services / Radiation Protection
- L. Brey, Manager, Nuclear Licensing and Fuels R. Burchfield, Superintendent, Betterment Engineering Nuclear Results R.- Craun, Manager, Nuclear Site Engineering D. Evans, Superintendent, Operations M. Ferris, Manager, QA Operations W. Franek, Superintendent, Plan / Scheduling & Stores
- C. Fuller, Station Manager D. Goss, Coordinator, Nuclear Licensing and Fuels
- J. Gramling, Supervisor, Nuclear Licensing Operations
- H. Holmes, Manager, Nuclear Licensing
- P. Moore, Supervisor QA Technical Support
- F. Novachek, Manager, Technical / Administrative Services
- T. Prenger, Manager, QA Service
- G. Reigel, Supervisor, Scheduling and Construction
- P. Tomlinson, Manager, QA 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 March 31, 1987.
2.
Licensee Action on Previous Inspection Findings . (Closed) Open Item (267/8401-02): Submittal of CRD0A Component Infomation and Revision to Procedure Q7. The licensee has documented the required CRD0A information in Memo PPC 84-0778. All CRD0A manipulations subsequent to this memo are documented by Procedure CMG12. Procedure Q7, Issue 9, includes requirements to tag and segregate nonconfoming items during receipt inspections. This item is closed.
(Closed)OpenItem(267/8410-01): Control of Sealed Valves Associated with ODRs. Administrative Procedure P-2, " Equipment Clearances and Operation Deviations," Issue 13, references Station Manager Administrative Procedure (SMAP) 19, " Processing Equipment Clearances and Operation Deviations."
Issue 6 of Procedure SMAP19 includes the appropriate instructions, and forms to control sealed and critical valves. This item is closed.
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(Closed) Open Item (267/8434-01): Control of Operation and Maintenance (0&M) Manuals. This item concerned the use of uncontrolled O&M manuals while performing maintenance or troubleshooting activities.
Procedure G-15 Issue 2, dated July 29, 1986, Control of Vendor Manuals, controls the use of vendor O&M manuals. G-15 states that vendor manuals ' - may be used for troubleshooting when procedural guidance is not available.
' If excerpts from a vendor manual are required to perform field work, a i temporary controlled copy shall be obtained form Engineering Services.
' This item is closed.
' l (Closed) Open Item (267/8434-02): Procurement Committee Recommendations.
l This item concerned the evaluation by the licensee of recommendations by a procurement committee. The NRC inspector reviewed Procedures Q-4, Issue 9, dated October 24, 1986, Procurement System, and ENG-16, Issue 2, dated January 5,1987, NED Procurement Document Evaluation.
Procedure Q-4 was revised significantly since the item was noted. Procedure ENG-16 is a new procedure issued in response to the evaluation of the procurement i comittee recomendations. The procedures are considered to be adequate.
< This item-is closed.
Closed Violation 267/8414-02): General Housekeeping
Closed Open Item 267/8415-03): General Housekeeping
Closed Violation 267/8507-08): General Housekeeping i The three. items above all related to inadequate housekeeping. The i licensee has implemented a program for plant tours and maintenance of general housekeeping under Procedure SMAP13, Issue 3, dated December 2,
1985, " Plant Tours and General Housekeeping Program." This procedure assigns managers and supervisors routine tour responsibilities and
provides a means of documenting housekeeping discrepancies. Housekeeping ' . conditions within the plant appeared to be satisfactory. Any housekeeping '
items noted by the NRC inspectors and discussed with the licensee during this inspection were promptly corrected and were considered to be isolated cases. These items are considered closed.
(Closed) Violation (267/8507-02): Failure to Follow Procedure i (Closed) Violation (267/8514-01): Failure to Follow Procedure
The two violations listed above concerned failure to follow procedures l resulting in the backwards installation of bearings during control rod drive mechanism refurbishment.
It was noted in NRC Inspection ' i Report 50-267/85-14 that the response by the licensee appeared adequate. The NRC inspector reviewed surveillance test data for the two drives taken during the last period of reactor criticality. The surveillances reviewed were
i SR 4.1.1.D-X, Issue 2, performed January 18, 1986, " Full Stroke Scram ,
Test;" and SR 4.1.1.C-X, Issue 3, perfonned April 27, 1986, " Precritical
10 inch Scram Test." The documentation and data were adequate. These items are closed.
l (Closed) Violation (267/8632-01): Failure to Wear Picture Badge l Identification Within the Protected Area. This violation involved a
j reactor operator not wearing his picture badge identification in the l control room. The operator had used his badge to get into the control , j ENCLOSURE CONTAINS !, , i UPON SEPARATION THIS
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5 room and had his badge in the control room but not displayed on his person. The licensee has conducted retraining and had taken disciplinary action. All of the control room staff have been counseled regarding the importance of wearing their security badges at all times. The NRC-inspectors have not detected recurrences of this problem. This violation is closed.
No violations or deviations were identified in this inspection area.
3.
Operational Safety Verification The NRC inspectors 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 systems status and limiting conditions for operation, and review of facility records.
Logs and records reviewed included: Shift supervisor logs . Reactor operator logs . Equipment operator logs . Auxiliary operator logs . Technical Specification compliance logs . Operations order book . ' Operations deviations reports . Clearance log . Temporary configuration reports . Station service requests (SSR) . During tours of accessible areas, particular attention was directed to the following: Monitoring instrumentation . Radiation controls . Housekeeping . Fluid leaks . Piping vibrations . Hanger / seismic restraints . Clearance tags . Fire hazards . Control room manning . Annunciators . During a tour of the control room, the NRC inspectors observed a significant (>20 psi) difference among the various reactor coolant system pressure indications. When questioned, the control room operators were ENCLOSURE CONTAINS. UPON SEPARATION THIS PAGE IS DECONTROLLED.
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< not familiar with the accuracy or tolerances of the different instruments.
The NRC inspectors then requested the licensee to provide information on the design requirements, calibration, accuracy, and tolerance of all , reactor coolant pressure indications.
Using this infomation, the NRC inspectors discovered that XI93508, a 0-100 psia digital instrument on
Control Room Board 109,~ had been calibrated on March 7,1987, to psig, not-psia..This instrument is one which is used to verify compliance with
Technical Specification LC0 4.2.7, which requires actions to be taken
prior to. exceeding 100 psia in the PCRV. The instrument was recalibrated using correct input and acceptance values on March 14, 1987.
! The summary of infomation prov<ded by the licensee also showed the ' reactor coolant pressure instruments with a 0-1000 psia range (PI1108, -1109,-1110) are calibrated only over the 400-800 psia range. The ' '. licensee explained that this was the nomal operating range and that by demonstrating linearity over this range, one could extrapolate accuracy over the' entire range of the instrument. The NRC inspectors concluded that since these instruments are also used to verify compliance with LC04.2.7 (100 psia), their calibration should also be checked in that range.. The licensee agreed to revise the calibration procedures to - include a 100-.900 psi calibration range. These instruments will be recalibrated using the new procedures prior to plant startup.
. Criterion XI of Appendix B to 10 CFR 50, as amplified by the licensee's J QA Program, and Section B.5.11.2 of the licensee's FSAR, require that ~ testing be performed in accordance with written test procedures which include the requirements and acceptance limits in the applicable design i or procurement document. The calibration of XI 93508 using incorrect-0-1000 psia instruments (PI1108, -1109, and -1110)ge of calibration of the iaput and acceptance values and the incomplete ran is an apparent ' i violation of NRC requirements.
(267/8708-01) ! During the inspection period, the licensee reported an incident very similar to one reported in February 1987. The incident involved a safety [ system with an inoperable plant protection system logic channel not in a i tripped condition.
In each case, the licensee had identified an inoperable signal channel not required to be operational at the time that ! the inoperable condition was identified.
In each case, the licensee generated an SSR requesting that the inoperable equipment be repaired.
, l Subsequent to such request and prior to repairs being completed, the i inoperable equipment was restored to service with the signal channel still inoperable.
In each case, the inoperable signal channel reduced the '
degree of redundancy in the plant protection system actuation for the effected system to below that required by Technical Specifications.
Each case was isolated and independent of the other.
In each case the problem i was licensee identified and reported in accordance with 10 CFR 50.72 with a licensee event report to follow.
In each case the reactor was shut down in cold shutdown with no safety significance because of the inoperable
channel. The NRC inspectors reviewing these incidents have verified that
standard operating procedures for the plant include the directive to place inoperable channel in;the trip condition if the channel is required to be , l ENCLOSURE CONTAINS.
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operational. The plant procedures are silent regarding inoperable channels that are not required to be operational. The NRC inspectors have discussed this problem at length with the operation superintendent, and - the operation superintendent has assured them that the plant operation procedures will be revised to include direction that inoperable channels will be placed in a tripped condition whenever such channel is discovered regardless of whether or not the channel is required to be operational at the time it is discovered to be inoperative. Such a change to plant procedures should preclude such rep (ortable incidence in the future.
This will be tracked as an open item.
267/8708-02).
4.
Licensee Action on IE Bulletins . The purpose of this portion of the inspection was to ascertain whether the infomation submitted by the licensee in response to IE Bulletins was technically adequate, satisfied the requirements established in the , bulletins, and represented the actions taken by the licensee. The NRC inspector reviewed the following IE Bulletins, including the licensee .' review and response documentation: (Closed) IE Bulletin'86-02: Static "0" Ring Differential Pressure Switches.
The licensee met the reporting requirements of IE Bulletin 86-02 by letters serial P-86490, dated July 28, 1986, and serial P-87069, dated February 20, 1987. The first letter stated that the licensee did not have SOR Model 102 or 103 differential pressure switches installed at-Fort St. Vrain in equipment important to safety or in systems which are subject to Limiting
Conditions for Operations of the FSV Technical Specifications. The second letter went on to commit that they would submit additional information regarding the future use of these differential pressure switches. This item is considered closed.
No violations or deviations were identified in this inspection area.
5.
Maintenance The NRC inspectors witnessed repairs on the "1A" gas waste compressor.
Since this machine compresses radioactive gas, it is enclosed in a tank to collect any leakage. Radiological controls and a radiation work pemit were required to enter this tank and perform the maintenance. The NRC inspector observed the replacement of the third stage valves and clearance checks on this compressor. The inspector also reviewed the associated ' paperwork (SSR 87502323) and found it to be in order. No discrepancies were noted.
The NRC inspectors observed governor adjustments and engine load balancing on Emergency Generator "18."
This generator is driven by diesel engines "1C" and "1D."
The governor from engine "1C" had been removed and sent, along with a spare governor, to the vendor for test and repairs. The
repaired governors were reinstalled, with the fomer spare governor now on engine "1D."
Neither engine started after repeated attempts, and the governor vendor's field representative was called in to troubleshoot the ENCLOSURE CONTAINS. UPON SEPARATION THIS PAGE IS DECONTROLLED.
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! problem.
It was disclosed during the troubleshooting that, at the-vendor's facility, the governor from the "1C" engine was found to have two pins.on an amphenol connector wired incorrectly. The wires were in a DC circuit.and were rewired correctly in accordance with the vendor's specifications. However, the wiring between the control panel and the goverr. ors was found to be reversed on the same DC circuit. Prior to this . maintenance, this entire DC control loop had apparently been wired with opposite polarities of what it should have. Since the entire DC circuit ' was installed in reverse, it had virtually no impact on governor ' operation. Now, with correctly wired governors and oppositely wired c feeders, the control circuitry would not function. The two wires for each governor were reversed _in a junction box and both governors and engines .then operated satisfactorily.
The NRC inspectors also observed work on Controlled Work Procedure CWP-86478 as modified by Deviation Report D86-478c. The work was on the "B" diesel generator set, engines K9205X and K9206X. The work was on the declutching limit switches, and the deviation report changed the configuration from normally open to normally closed on the limit switches. The switches assure the engine is put in the manual mode if it declutches when attempting to start.
The NRC inspectors also observed work performed under Station Service ' Request (SSR) 86513796, which covered the annual inspection on the "B" diesel generator set.
The NRC inspectors also followed the troubleshooting effort to eliminate i , noise problems on instrument channels, especially neutron monitoring , system channels, being performed under SSR 86513001. This SSR was written
on November 4,1986, and as originally written estimated an expenditure of i 4 manhours to troubleshoot the noise problem _on the wide range flux-rate-
of-change channels. The SSR was reviewed by the NRC inspectors on ' March 10, 1987, and they observed that a crew-was working full time on this SSR. The operations superintendent was interviewed regarding the ' - apparent expansion of the job without a change in the documentation. The operations superintendent advised that this was normal but was being tracked and the time expended would be documented on the SSR when it was - closed. The operations superintendent did share the NRC inspectors' concern regarding the expanding scope of work on instrumentation important
< to safety and added a full time QC inspector to the wurk crew to document i the crew's activities and to stop work if it proceeded to a point requiring engineering and or management review.
- The NRC. inspectors observed troubleshooting and repair work to the plant protection system (PPS) done under.SSR 87503732. The reactor operators
, had identified a problem with the handswitch controlling " steam turbine bypass block valve-loop -1."
The operators had identified a mismatch in ! the "A" and "B" trains of PPS logic regarding this handswitch. The r technician performed troubleshooting efforts utilizing Drawing E1203, r Page 403, which describes the logic of the handswtich and the associated l relay logic and GA Drawing ELJ169-3101. The technician discovered a l , l ENCLOSURE CONTAINS l. I UPON SEPARATION THIS PAGE IS DECONTROLLED.
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faulty logic module, which was inputting an erroneous signal that told the PPS that the steam turbine on the "B" helium circulator was tripped.
Module CT1-BR2, " Circulator Trip Module on 'B' Logic, second circulator," was replaced with a spare module which had been previously repaired under SSR 87503555, utilizing Procedure RP79, Issue 5, " Shop Testing and Maintenance of CT1 Circulator Trip End Sections." The faulty module identification and replacement was performed under Procedure RP-S01, ' Issue 1, " Plant Protective System Troubleshooting Guidelines and Module ! Replacement." The module replacement was documented on Form RP-S-01, . Attachment C, "PPS Module Replacement Fom." The work was completely witnessed by QC and independently verified by a results engineer. The
replacement module was verified to be performing properly after installation by checking the input voltages using Surveillance Procedure SR-RE30-B, Issue 3, " Testing Logic Module Input Voitages."
No violations or deviations were identified in this inspection area.
' 6.
Surveillance The NRC inspectors witnessed calibration of the primary coolant dewpoint moisture monitoring equipment in accordance with Surveillance Procedure SR 5.4.1.1.6.c-R.
This surveillance is normally performed during a refueling outage, but, because of the extended time of the present outage ' and the considering the time between the last perfomance of this surveillance and next projected refueling outage, the licensee decided to run this surveillance prior to startup.
The inspectors reviewed the surveillance procedure for conformance with , Technical Specification requirements. Performance of the surveillance was witnessed in order to verify: Required administrative approvals and clearances were obtained prior ' . to test initiation.
, Test instrumentation was calibrated.
. The system being tested was removed from and restored to service . properly.
Conduct of the surveillance test was performed in accordance with the + . procedure.
Test results were within acceptance criteria and met Technical . Specification requirements. Discrepancies documented and rectified.
The inspector had several concerns with the procedure being difficult to follow or understand. Specifically, on certain steps, when as found values were out of tolerance, there was no explanation or direction to generate an SdR to correct the condition. Some steps have a note to do so, but others do not. Although every person interviewed by the inspector knew what to do when as-found values were out of tolerance and no ' ENCLOSURE CONTAINS. UPCN SEPARATION THIS PAGE IS DECONTROLLED.
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indications of' improperly performed surveillances were found, the SSR' number should be indicated on the surveillance procedure when an out of-tolerance value is encountered. The licensee agreed to incorporate this and additionally to revise the way in which acceptance criteria were expressed as part of the ongoing procedure rewrite task.
The NRC inspectors also observed portions of the annual maintenance and inspection of emergency diesel generator engines "1C" and "10."
Surveillance Procedure Sr-MalB-A, " Preventive Maintenance Inspection of Emergency Diesel Generator Units K9205X and K9206X," was used and reviewed .by the inspectors. This surveillance was performed in part to investigate the intermittent failure of engine "1C" to start and governor problems associated with keeping the engines in balance. Additional observations on work perfonned are included in paragraph 4 of this report. This surveillance was completed satisfactorily, and no discrepancies were noted.
Because the root cause of the intermittent failures of engine "1C" to start was never identified, a requalification program was implemented to provide assurance of the operability of the "1B" emergency diesel generator set. This program involved starting and loading the "1B" diesel generator to greater than or equal to 1200 KW for a minimum of 60 minutes once per day for 7 consecutive days. This requalification program was successfully completed on March 28, 1987.
The NRC inspectors observed results technicians preparing to perform Surveillance Proedure SR 5.3.10ab-AX, Issue 5. " Reheater Safe Shutdown Cooling Instrument Tests." The tests were being performed to verify electrical maintenance done on PT2267 and PT-2268, which were found out of calibration on a previous surveillance and repaired on SSR 87503528. The NRC inspectors noted that several of the steps to be performed by the technicians were stamped with a QC stamp, requiring QC witnessing of these steps in the procedure.
No QC inspector was present; the work had not yet begun. The technicians ceased their activities and went to obtain a QC inspector prior to proceeding. The NRC inspectors continuing their plant tour later in the day returned to the area of work and found that a QC inspector was present and prepared to witness the work.
No violations or deviations were fcund in this inspection area.
7.
Engineered Safety Features The NRC inspectors perfonned a walkdown of all accessible portions of the dewpoint moisture monitoring system. This system provides identification of a leaking loop, alarm loop shutdown, steam / water dump, and reactor scram functions. The inspector reviewed applicable sections of the Reference Design Manual, FSAR, and Technical Specifications to become familiar with the system and its requirements.
The following items were examined as part of this inspection: ENCLOSURE CONTAINS. UPON SEPARATION THIS PAGE IS DECONTROLLE. . 11~ Agreement between plant drawings and the as-found configuration . Equipment conditions"and. items which might degrade performance . Interior of PCRV penetrations . Valves in proper position . . r . Instruments functioning and within appropriate calibration dates . These instruments were being calibrated during this inspection period.
Observations of this by the NRC are included in Section 5 of this report.. No violations or deviations were identified in this inspection area.
8.
Design Control The licensee contacted the NRC-inspectors regarding a solution to a problem identified in testing. Specifically, the licensee had replaced reactor pressure vessel thermocouples during the outage and during the functional tests on the thermocouples' discovered that they could no longer test the thennocouples in accordance with Technical Specifications.
Specifically, the original thermocouples were wired in a grounded configuration.. The thennocouples have a signal transmitter and an internal test unit all wired in a grounded configuration. During the replacement, some of the thermocouples were intentionally replaced by thermocouples in an ungrounded configuration. This required modification of the signal transmitter units to an ungrounded configuration to correspond with the ungrounded condition of the thermocouples involved.
The licensee did not change the internal test signal apparatus to an ungrounded configuration. Consequently, the ungrounded thermocouples could not be tested with an internal test signal from the grounded test signal equipment. Technical Specification SR 5.4.1 specifically requires that these thermocouples be tested with an internal test signal.
Consequently, the new ungrounded thennocouple configuration could not meet the existing Technical Specifications.
' The NRC inspectors did not identify a significant safety concern with this specific case (thermocouples being grounded or ungrounded). licwever, the NRC inspectors did question the licensee design control and design review process, which allowed the equipment to be modified in the field and j installed in a configuration that could not meet existing Technical Specifications. The NRC inspectors reviewed Change Notice (CN)-2144, " Replace Safe Shutdown Foxboro Thermocouples With Weed Instruments j Qualified Thermocouples," and CN 2246 and CN 2144A which replaced ' additional thennocouples. The NRC inspectors also reviewed the independent design verifications and the 10 CFR 50.59 safety evaluations associated with these change notices. The change notices and safety evaluations all specifically acknowledge the change from grounded to ungrounded of those thermocouples that were changed. The change was intentional to solve a noise problem in the signal channels. The required
modifications to the signal transmitters were identified. The required - Technical Specification surveillance that must be met was identified.
i Nowhere in the documentation was it recognized or identified that the new configuration could not meet the existing Technical Specifications without i ' ENCLOSURE CONTAINS, UPON SEPARATION THIS i PAGE IS DECONTROLLED.
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additional modification. The licensee designed, approved, reviewed, and installed a change to the plant which precluded meeting existing Technical Specifications. This is an apparent violation of NRC regulations.
(267/8708-03) The NRC inspectors have been following changes to the plant configuration effected through the temporary change process as described in Station Manager's Administrative Procedure SMAP-18 Issue 3, " Processing of Temporary Configuration Reports." The Procedure SMAP-18 does not contain an explicit definition of " temporary. However, an implicit definition is contained in Procedure Step 4.5-63C, " Keep a record of those TCR's that have been in affect for more than three months and are still required."
The licensee recognizes that temporary changes in effect for more than 90 days are unusual and require additional tracking and identification.
The licensee has ongoing program among operations, engineering, and ' licensing to solve the issue of backlog of temporary change requests. The temporary changes, most of which have a permanent change request in process, are being processed and permanetized by nuclear engineering.
This, however, is an old problem as evidenced by the dates on the temporary change reports and-is one that the licensee has been attempting to address for some time. The NRC inspectors will continue to monitor both the temporary change backlog and the licensee's effort to reduce this backlog. This will be tracked has an open item.
(267/8708-04) _ 9.
Security One unresolved item was-identified in this inspection area and is documented in the_ attachment to this report.
No violations or deviations were identified in this inspection area.
10. Management and Exit Meetings During the inspection period, the NRC inspectors conducted several meetings with the licensee senior management to review progress towards outage completion and preparation for plant startup.
An exit interview was conducted on March 31, 1987, attended by those indicated in paragraph 1.
At this time, the NRC inspectors reviewed the scope and findings of the inspection.
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