IR 05000250/1986030
| ML20203C135 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 07/02/1986 |
| From: | Brewer D, Brooks C, Elrod S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20203C108 | List: |
| References | |
| 50-250-86-30, 50-251-86-30, NUDOCS 8607180360 | |
| Download: ML20203C135 (9) | |
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UNITED STATES
[pa Ofcoq'o, NUCLEAR REGULATORY, COMMISSION
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REGloN 11 j
k.s 101 MARIETTA STREET, N.W.
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"2 ATL ANTA. GEORGI A 30323
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i Report Nos.:
50-250/86-30 and 50-251/86-30
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Licensee:
Florida Power and Light Company
9250 West Flagler Street
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Miami, Florida 33102
Docket Nos.: 50-250 and 50-251 License.Nos.: DPR-31 and DPR-M
l Facility Name: Turkey Point 3 and 4 Inspection Conducted: May 12 - June 9, 1986 Inspectors:
- &uk M/ 2, /f[d j
D. R. Brewer, Senior Resident Inspector (/'DatySigned
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i bMfA _ /M b J17. /$$5 C.R. Brooks, Reside]tIgspector,BrownsFerry
[ Dye Signed d L,1 %
Approved by:
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Stephen A. Elrod, Section Chief Datd Signed Division of Reactor. Projects
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SUMMARY Scope: This routine, unannounced inspection entailed direct inspection at the site, including backshift inspection, in the areas of licensee action on previous i
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inspection findings, annual and monthly surveillance observations, maintenance
observations and reviews, operational safety verification, engineered safety l
features walkdown, independent inspection, and follow-up of plant events.
l Results: Violation - Failure to meet the requirements of Technical Specification (TS) 6.8.1 (paragraph 5).
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L 8607180360 860707 PDR ADOCK 00000250 G
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REPORT DETAILS
1.
Licensee Employees Contacted
- C. M. Wethy, Vice President - Turkey Point
- C. J. Baker, Plant Manager-Nuclear - Turkey Point
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F. H. Southworth, Administration Department E. Preast, Site Engineering Manager
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D. W. Hasse, Safety Engineering Group Chairman D. D. Grandage, Operations Superintendent - Nuclear
1 T. A. Finn, Operations Supervisor i
- V. A. Kaminskas, Acting Operations Supervisor J. Crockford, Assistant Operations Supervisor
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l J. Webb, Operations / Maintenance Coordinator O. A. Chaney, Corporate Licensing i
- J. Arias, Regulation and Compliance Supervisor
- R. Hart, Regulation and Compliance Engineer
- J. W. Kappes, Maintenance Superintendent-Nuclear J. C. Strong, Electrical Maintenance Supervisor
R. A. Longtemps, Mechanical Maintenance Supervisor
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D. Tomasewski, Instrument and Control (I&C) Maintenance Supervisor i
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R. G. Mende, Reactor Engineering Supervisor R. E. Garrett, Plant Security Supervisor
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P. W. Hughes, Health Physics Supervisor i
W. C. Miller, Training Supervisor
J. M. Donis, Site Engineering Supervisor J. M. Mowbray, Site Mechanical Engineer R. H. Reinhardt, Quality Control (QC) Inspector
- L. W. Bladow, Quality Assurance (QA) Superintendent
- J. A. Labarraque, Technical Department Supervisor
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- E. L. Anderson, Inservice Test (IST) Specialist
- B. A. Abrishami, System Performance Engineer
- H. E. Hartman, IST Engineer NRC Personnel
- W. Kleinsorge, NRC Region II Inspector
- J. E. Menning, NRC Region II Inspector Other licensee employees contacted included construction craftsmen,
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engineers, technicians, operators, mechanics, electricians and security force members.
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- Attended exit interview l
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i 2.
Exit Interview
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j The inspection scope and findings were summarized during management interviews held throughout the reporting period with the Plant Manager -
Nuclear and selected members of his staff.
An exit meeting was conducted on June 6,1986. The areas requiring n.:nage-
ment attention were reviewed.
One violation was identified:
Failure to meet the requirements of TS 6.8.1, in that Operating Procedure (0P) 14004.1 was improperly implemented (250/86-30-01)
(paragraph 5).
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Two Inspector Followup Items (IFIs) were identified:
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Inspect the interim control of measuring and test equipment (M&TE)
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restricted use labels and the procurement of new decade boxes (250, 251/86-30-02) (paragraph 5).
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j Review the licensee's engineering evaluation of noise in the steam flow
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and pressure signals (250,251/86-30-03) (paragraph 5).
The licensee did not identify as proprietary any of the materials provided
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to or reviewed by the inspectors during this inspection.
The licensee j
acknowledged the findings without dissenting comments.
j 3.
Licensee Action on Previous Inspection Findings (92702)
j a.
Performance Enhancement Program (PEP) Summary i
j A meeting between NRC Region II. management and the licensee was held on i
May 29-30 at the Turkey Point Site to discuss, in detail, the progress
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of the PEP program. After the meeting, NRC Region II management toured
the nuclear plants.
l b.
Previously Identified Items
(Closed) Unresolved Item 250,251/86-10-05 - This unresolved item involved an NRC concern that degraded pipe supports adjacent to the
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Unit 3 intake cooling water (ICW) discharge headers could have allowed the circulating water pump (CWP) lube water system piping to shear at
,the lube water /ICW pipe joint, resulting in a reduction of ICW flow.
This issue was discussed in Inspection Report Nos. 250,251/86-13. The degraded pipe supports were determined not to meet the requirements of.
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i paragraph (a)(1) of 10 CFR 50.55a, and ~ Violation 250,251/86-13-02 was
issued. The licensee performed an evaluation' which verified that the degraded supports did not meet the seismic requirements specified in l
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I the Final Safety Analysis Report (FSAR) but did meet functionality criteria and were capable of withstanding.at least one seismic event.
Therefore, no potential for pipe shear existed assuming that the pipe in question would have been replaced following a single, design basis
seismic event. The degraded pipe supports were expeditiously replaced
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with new stesmically qualified supports. Additional concerns relative to the degradation of pipe supports will receive followup review as part of the licensee's corrective action for Violation 250,251/86-13-02, which remains open.
4.
Unresolved Items An unresolved item is a matter about which more information is required to determine whether it is acceptable or'may involve a violation or deviation.
No unresolved items were identified during this inspection. One unresolved
item is closed in this report as discussed in paragraph 3.
5.
Monthly and Annual Surveillance Observation (61776/61700)
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The inspectors observed TS required surveillance testing and verified the following: that the test procedure conformed to the requirements of the TS,
that testing was performed in accordance with adequate procedures, that test j
instrumentation was calibrated, that limiting conditions for operation (LCOs) were met, that test results met acceptance criteria requirements and
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were reviewed by personnel other than the individual directing the test, that deficiencies were identified, as appropriate, and were properly reviewed and resolved by management personnel and that system restoration was adequate.
For completed tests, the. inspector verified that testing frequencies were met and tests were performed by qualified individuals.
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The inspectors witnessed / reviewed portions of the following test activities:
Reactor Protection System (RPS) Logic Test, 3-OSP-049.1
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Differential Temperature Calibration, 4-PMI-041.8 Steam Generator Protection Channels - Periodic Test, OP-14004.1
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a.
RPS Logic Test The inspector observed the performance of 3-0SP-049.1 on June 3,1986, l
during the second use of the new, upgraded procedure. Two On-The-Spot Changes (OTSCs) were initiated to correct minor procedural inaccuracies detected during the previous performance.
The test was completed
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smoothly with no discrepancies, i
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Average Temperature - Delta T Calibration
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The inspector observed the performance of 4-PMI-041.8 on June 4, 1986,
during the first use of the new, upgraded procedure. Several errors in the procedure had been detected by the I&C specialists who performed j
the work and 12 OTSCs had been processed since the work started.
The
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procedure also contained new acceptance criteria for AC ripple. Four neutron monitoring and temperature monitoring modules failed the new criteria and were replaced.
The failures were reportedly caused by end-of-life failure of electrolytic capacitors.
The inspector identified a potential for inadequate control of measuring and test equipment (M&TE) used during this calibration. Two General Radio 1K decade boxes used for the calibration had " Restricted Use" stickers attached to them.
The sticker on serial number PTN-I-301-5 stated that the actual resistance was +.006 ohms greater than the dial setting, and the sticker on serial number PTN-I-301-3 stated that the actual resistance was +.0067 ohms greater than the dial setting. The sticker on this decade box also contained the following data:
.01 ohm steps Tolerance Actual 1 x.01
.0002
.0006 2 x.01
.0004
.0007 3 x.01
.0006
.0008 Administrative Procedure 0190.9, Control of Measuring and Test Equipment, Section 8.4.9, states that when a scale or function of a multiscale or multifunction instrument is not calibrated, a restricted use label that indicates the restriction and/or that the scale or function is not calibrated shall be applied. The inspector questioned whether the information contained on the restricted use labels provided adequate guidance to the M&TE user.
Conflicting statements were obtained from several licensee representatives as to whether the data provided on the sticker was to be used to correct the decade box indicated resistance values or whether the user was to evaluate the accuracy of the decade box prior to use to assure sufficient accuracy for his intended use.
The inspector located eight additional decade boxes (six in the calibration lab and two outside the control room) all of which had failed to meet the acceptance criteria on one or more ranges during their last calibration.
All of these decade boxes contained restricted use labels similar to those described above.
Licensee representatives agreed that a potential problem existed with the control of the decade boxes and initiated immediate corrective action. New decade boxes were to be procured and all M&TE users were to be trained on restricted use labels in the interim. This item will be tracked as an Inspector Followup Item (250,251/86-30-02) pending replacement of the decade boxes and an evaluation of the interim controls.
AP 0190.9 also requires that any M&TE restrictions shall be noted on
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the calibration data sheet and on the M&TE use log. This requirement was not satisfied for the two decade boxes in use during the conduct of 4-PMI-041.8.
Licensee representatives stated that they were in the process of transferring to a new computer based M&TE tracking and I
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I control system which will eliminate this concern. With the new system,
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M&TE users will check-out equipment with a bar code identifier. The computer will provide a traveler (printout) which will contain equipment calibration data and any restrictions. The traveler' will be attached to work orders for retention as a QA record.
Licensee representatives stated that they were the first utility to use this new M&TE control system and expected immediate benefits from its use.
c.
Steam Generator Protection Channels - Periodic Test
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The inspector observed the performance of Operating Procedure 14004.1 on Unit 3 on June 5, 1986.
The procedure was performed by two specialists and an I&C supervisor. One specialist read the procedure while the other performed the switch and test equipment manipulations.
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Prior to performing critical switch manipulations, the specialist repeated the switch identification back to the reader and the super-i visor verified that the proper switch was selected. This policy for the conduct of reactor protection system testing was instituted several years ago by the licensee and was claimed to have been effective in
eliminating all inadvertent trips due to personnel error.
The test
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progressed smoothly until step 8.3.82 was reached when a " clean" action of bistable PC-3-485A could not' be achieved.-
The - bistable proving light was observed to flash intermittently at the trip point and to
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chatter rapidly at the reset point, making it impossible to determine the exact trip and reset points. The test methodology specified in the
procedure was then deviated from in an attempt to prove bistable operation. A power supply was used to input a test signal directly into the PC-3-485A module in lieu of the procedurally required transmitter simulator connected to test jack TJ-3-485. The digital voltmeter test leads were also changed from test point TP-3-485-6 to the power supply output test leads. When this method was used, a clean bistable operation was observed which was within the acceptance
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criteria provided in the procedure. The inspector was informed by the I&C supervisor that since this problem has existed for some time and
was well known, the deviation from the procedure would not be documented on the data sheets or work order and no procedure change would be initiated. This failure to adhere to procedures is a viola-tion of TS 6.8.1 (250/86-30-01).
Licensee management indicated in the exit meeting that the alternate test method would be evaluated and the procedure changed as appropriate.
The licensee re-emphasized its policy of verbatim compliance with procedures.
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Subsequent communication with licensee representatives indicated that the problem was a result of noise inherent to the steam flow and steam pressure signals. This noise may or may not be noticed on the control i
board indicators as an oscillation in the indication.
Bistable chattering results when the process signal is at or near the setpoint
and the noise strength exceeds 40 millivolts peak to peak.
The licensee maintains that the naise creates a tendency for the bistable
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to trip prematurely in the conservative direction.
This problem will
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be tracked as Inspector Followup Item 250,251/86-30-03 to review the engineering evaluation of noise in the steam flow and pressure signals.
6.
Maintenance Observations (62703/62700)
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Station maintenance activities on safety-related systems and components were observed and reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards and in conformance with the TS.
The following items were considered during this review, as appropriate:
that LCOs were met while components or systems were removed from service; i
that approvals were obtained prior to initiating work; that activities were accomplished using approved procedures and were inspected as applicable; that procedures used were adequate to control the activity; that trouble-shooting activities were controlled and repair records accurately reflected the maintenance performed; that functional testing and/or calibrations were performed prior to returning components or systems to service; that QC records were maintained; that activities were accomplished by qualified personnel; that parts and materials used were properly certified; that radiological controls were properly implemented; that QC hold points were established and observed where required; that fire prevention controls were implemented; that outside contractor force activities were controlled in accordance with the approved QA program; and that housekeeping was actively pursued.
The following maintenance activities were observed and/or reviewed:
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Installation of protected area fencing
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Installation of protected area fence vibration sensor system
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Replacement of protected area gate 43 lock mechanism
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Partial rebuild of Unit 4 power operated relief valves
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Plant Work Order (PWO) #058145 for pipe support installation No violations were identified in this area, 7.
Operational Safety Verification (71707)
The inspectors observed control room operations, reviewed applicable logs, conducted discussions with control room operators, observed shif t turnovers and confirmed the operability of instrumentation. The inspectors verified the operability of selected emergency systems, verified that maintenance work orders had been submitted as required and that followup and prioriti-zation of work was accomplished.
The inspectors reviewed tagout records, verified compliance with TS LCOs and verified the return to service of affected components.
By observation and direct interviews, verification was made that the Physical Security Plan was being implemente.
Plant housekeeping / cleanliness conditions and implementation of radiological controls were observed.
Tours of the intake structure and the diesel, auxiliary, control and turbine buildings and the Unit 4 containment were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks and excessive vibrations.
The inspectors walked down accessible portions of the following safety-related systems on Unit 3 and, as applicable, on Unit 4 to verify oper-
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ability and proper valve / switch alignment:
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4160 Volt and 480 Volt Switchgear
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Control Room Vertical Panels and Safeguards Racks During a general plant tour conducted during the week of June 2,1986, the following deficiencies were noted and turned over to the appropriate licensee representatives for resolution:
a.
Wooden shims were found wedged between pipes and the wall penetration for the safety injection system (SIS) cold leg and hot leg piping on Unit 3 as they passed from the SIS pump room to the component cooling water (CCW) pump room, b.
The pipe clamp for the instrument piping connected to the low side of FT-3-943 was loose.
This is an SIS flow transmitter sensing line.
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c.
A copy of the remote air sampling procedure was found taped to the wall just outside the Unit 3 containment spray (CS) pump room.
It was not possible to ascertain whether the procedure was controlled.
8.
Engineered Safety Features Walkdown (71710)
The inspector performed a partial verification of the operability of the Unit 3 containment spray system by inspecting the accessible portions of the system. The verification will be completed in June 1986 as part of the next monthly resident inpection.
The following specifics were reviewed and/or observed as appropriate for selected portions of the system:
a.
that the licensee's system lineup procedures matched plant drawings and the as-built configuration;
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b.
that the equipment conditions were satisfactory and items that might degrade performance were identified and evaluated (e.g., hangers and supports were operable, housekeeping was adequate, etc.);
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that instrumentation was properly valved in and functioning and that calibration dates were not exceeded;
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that valves were in proper position, breaker alignment was correct, power was available, and valves were locked /lockwired as required; e.
local and remote position indication was compared and remote instru-
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mentation was functional; and
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breakers and instrumentation cabinets were inspected to verify that they were free of damage and interference.
No violations or deviations were identified.
9.
Independent Inspection
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I During the. report period, the inspectors routinely attended meetings with licensee management and monitored shift turnovers between shift supervisors, shift foremen and licensed operators.
These meetings included daily discussions of plant operating and testing activities as well as discussions
of significant problems or incidents.
The inspectors reviewed potential
problem areas to independently assess their importance to safety, the j
adequacy of proposed solutions, improvement and progress, and the adequacy of corrective actions. The inspector's reviews of these matters were not i
limited to the defined inspection program.
Independent inspection effort l
was continued in the area of emergency diesel generator (EDG) loading and security plan implementation.
Concerns identified in the security area
prompted a special inspection by an NRC Region II security specialist and will be documented in Inspection Report Nos. 250,251/86-32.
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10.
Plant Events (93702)
An independent review was conducted of the following events.
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a.
On May 16 and again on May 27, the emergency notification system (ENS)
telephone was briefly out-of-service. The ENS telephone failure lasted
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30 minutes on May 16 and 40. minutes on May 27.
On both occasions
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commercial telephone service was available between the plant and the NRC Operations Center. The trouble with the circuit did not originate at the site on either occasion.
b.
On May 21, 1986, the licensee's continuing analysis of EDG loading l
limitations determined that a single failure of a vital-to-nonvital
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electrical tie breaker could cause the EDGs to overload. Compensatory-
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tion JPE-L-86-59.
The potential for the EDGs to overload under i
accident conditions is the subject of continuing NRC Region II and Office of Nuclear Reactor Regulation evaluation and will be addressed in Inspection Report Nos. 250,251/86-29.
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