IR 05000250/1993029
| ML17352A433 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 01/20/1994 |
| From: | Binoy Desai, Johnson T, Landis K, Trocine L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17352A432 | List: |
| References | |
| 50-250-93-29, 50-251-93-29, NUDOCS 9402080106 | |
| Download: ML17352A433 (47) | |
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Report Nos.:
50-250/93-29 and 50-251/93-29 (
Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33102 UNITED STATES NUCLEAR REGULATORY COMMlSSION
REGION II
10l MARIETTASTREET, N.W., SUITE 2900 ATLANTA,GEORGIA 303234I99 Inspection Conducted:
Novemb 28 through December 31, 1993 Inspectors:
T.
P. Johnson, Senio Resident Inspector D
e igned Docket Nos.:
50-250 and 50-251 License Nos.:
DPR-31 and DPR-41 Facility Name:
Turkey Point Units 3 and
B.
B. Desai, R
ide Inspector C'a S'gned L. Trocine, Resident nspector-Date Signed Approved by:
K. D. Land
, Chief Reactor Projects Section 2B Division of Reactor Projects Date igned SUHHARY Scope:
This resident inspection to assure public health and safety involved direct inspection at the site in the areas of operational safety, plant events, maintenance observations, surveillance observations, followup of previous items, review of written reports, self-assessment, and design changes and modifications.
Backshift inspections were performed in accordance with Nuclear Regulatory Commission policy.
Results:
Within the scope of this inspection, the inspectors determined that the licensee continued to demonstrate safe plant operations.
The following two non-cited violation were identified:
Non-Cited Violation 50-250,251/93-29-01, combustible chemistry and health physics materials stored without transient combustible permits (section 4.2.3).
9402080106 940121 PDR ADOCK 05000Z50
Non-Cited Violation 50-250,251/93-29-02, exceeding overtime limits specified in the Technical Specifications (section 4.2.6).
During this inspection period, the inspectors had comments in the following Systematic As'sessment of Licensee Performance functional areas:
0 erations Two operators exceeded overtime requirements that resulted in a non-cited violation (section 4.2.6).
Periodic reports were appropriately written and submitted (section 9.2).
The Plant Nuclear Safety Committee
'demonstrated a good safety conscious approach to resolving issues that occurred during the period (section 10.2. 1).
A quality assurance review of recent personnel errors was thoroughly conducted and appeared effective in problem identification and corrective actions (section 10.2.2).
Unit 3 and 4 turbine volumetric flow tests were well planned and conservatively conducted, and strong management involvement and oversight was evident (section 11.2.2).
Maintenance and Surveillance
Witnessed surveillance and maintenance activities were appropriately and conservatively conducted (sections 6.2.1 and 7.2.1).
Safety-related equipment troubleshooting activities for the reactor protection system and emergency diesel generator demonstrated a safety conscious attitude and approach (sections 6.2.2 and 5.2.2).
Continued licensee conservatism was demonstrated during followup activities for the safety injection pump motor rotor bar cracking issue (section 8.2.2).
This included maintenance assessments and diagnostic testing.
This issue remains open pending completion of an accelerated schedule for motor replacements.
En ineerin and Technical Su ort Engineering and technical support during the troubleshooting and repair activities relating to a failure of the 48 emergency diesel generator were conducted in an efficient and conservative manner (section 5.2.2).
Strong engineering support was noted during the Unit 3 residual heat removal system flushes (section 4.2.5)
and the Unit 3 and 4 turbine volumetric tests (section 11.2.2).
A safety evaluation performed by the licensee to address potential violations of design criteria for electrical systems represented a substantial effort by engineering to support plant operations in recent months.
The safety evaluation was well defined, the requirements and criteria were clearly stated, and the conclusion was adequately supported.
In addition, questions on the safety evaluation were fully answered, and the inspector followup item was closed (section 8.2.1):
Modifications related to the relaxed axial offset control methodology technical specification change were effectively implemented (section 11.2.1).
A potential
CFR Part
issue related to wire and cable products was also conservatively addressed (section 11.2.3).
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Plant Su ort Radiation Controls Emer enc Pre ar edness Securit Chemistr Fire Protection Fitness For Dut and Housekee in Controls Implementation of the hand geometry biometric system for site access control was smooth and efficient (section 4.2.1).
The annual emergency preparedness exercise was successfully conducted (section 4.2.4).
Management displayed a safety conscious attitude in requiring strict radiological controls and assurance of system operability during residual heat removal system flushing activities (section 4.2.5).
Very good coordination and team work were also noted among operators, system engineers, and health physics personnel during these flushing activities.
Hechanical maintenance, health physics, and fire brigade personnel responded to a fire in the Rad Waste Building in a timely and efficient manner, and licensee corrective actions were prompt and effective (section 5.2. 1).
However, post-fire followup determined that transient-combustible material was stored in three locations in the Rad Waste Building without the appropriate permits.
This resulted in a non-cited violation (section 4.2.3).
TABLE OF CONTENTS 1.0 Persons Contacted 1.1 Licensee Employees 1.2 NRC Resident Inspectors 2.0 Other NRC Inspections Performed During This Period 3.0 Plant Status 3.1 Unit 3 3.2 Unit 4 4;0 Operations 4.1 Inspection Scope 4.2 Inspection Findings 5.0 Plant Events 5.1 Inspection Scope 5.2 Inspection Findings 6.0 maintenance Observations 6.1 Inspection Scope 6.2 Inspection Findings 7.0 Surveillance Observation 7. 1 Inspection Scope 7.2 Inspection Findings 8.0 Followup on Previous Items and Noncompliances 8.1 Inspection Scope 8.2 Inspection Findings 9.0 Onsite Followup and In-Office Review of Written Reports 9.1 Inspection Scope 9.2 Inspection Findings 10.0 Evaluation of Licensee Sel f-Assessment Capability
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15 11.0 10.1 Inspection Scope 10.2 Inspection Findings Design, Design Changes, and Modifications
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'Table of Contents ll.1 Inspection Scope 11.2 Inspection Findings 12.0 Exit Interviews 13.0 Acronyms and Abbreviation
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REPORT DETAILS 1.0 Persons Contacted l. 1 Licensee Employees Abbatiello, Site guality Manager Bowskill, Reactor Engineering Supervisor Franzone, Instrumentation and Controls Maintenance Supervisor Gianfrancesco, Maintenance Support Services Supervisor Heisterman, Mechanical Maintenance Supervisor Higgins, Outage Manager Hollinger, Training Manager Jernigan, Operations Manager Johnson, Operations Supervisor Kaminskas, Services Manager Kirkpatrick, Fire Protection/Safety Supervisor Knorr, Regulatory Compliance Analyst Kundalkar, Engineering Manager Lindsay, Health Physics Supervisor rchese, Site Construction Manager Harcussen, Security Supervisor Pearce, Plant General Manager Pearce, Electrical Haintenance Supervisor Plunkett, Site Vice President Powell, Technical Manager Rose, Nuclear Haterials Manager Rossi, guality Assurance Supervisor Steinke,.Chemistry Supervisor Wayland, Maintenance Manager Weinkam, Licensing Manager T. V.
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Other licensee employees contacted included construction craftsman, engineers, technicians, operators, mechanics, and electricians.
1.2 NRC Resident Inspectors B. B. Desai, Resident Inspector T.
P. Johnson, Senior Resident Inspector L. Trocine, Resident Inspector
Attended exit interview on January 5,
1994 Note:
An alphabetical tabulation of acronyms used in this report is'isted in the last paragraph in this repor.0 Other NRC Inspections Performed During This Period Re ort No.
Dates Area Ins ected 50-350,251/93-27 December 13-17, 1993 Annual Emergency Exercise 50-250,251/93-28 3.0 Plant Status November 29-December 3, 1993 Health Physics 3.1 3.2 Unit 3 Unit 3 operated at 100% power throughout this reporting period and had been on line since October 20, 1993.
Unit 4 Unit 4 operated at 100% power throughout this reporting period and had been on line since August 17, 1993.
4.0 Operational Safety Verification (71707)
4.1 Inspection Scope The inspectors observed control room operations, reviewed applicable logs, conducted discussions with control room operators, observed shift turnovers, and monitored instrumentation.
The inspectors verified proper valve/switch alignment of selected safety systems, verified that maintenance work orders had been submitted as required, and verified that followup and prioritization 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 implemented.
The implementation of radiological controls, fire protection, fitness for duty, chemistry, emergency preparedness, and plant housekeeping/cleanliness conditions were also observed.
Tours of the intake structure and diesel, auxiliary, control, and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations.
4.2 4.2.1 Inspection Findings Implementation of the Hand Geometry Biometric System In order to implement a hand geometry biometric system for site access control such that photographic identification badges could
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4.2.2 be taken off site, the licensee submitted an application for exemption from certain requirements of 10 CFR 73.55, Requirements for Physical Protection of Licensed'Activities in Nuclear Power Plant Reactors Against Radiological Sabotage.
The applicable
CFR 73.55 requirements related to the issuance, storage, and retrieval of badges for personnel who have been granted unescorted access to the site's protected areas.
The licensee's applications for exemption were submitted on October 13 and November 2, 1993; and in order to test the new system while NRC approval was pending, the licensee began to utilize the hand geometry biometric system in lieu of personal identification numbers on November 12, 1993.
During this time frame, the licensee did not allow the photographic identification badges to be removed from the site in order to continue satisfying the requirements of the existing Physical Security Plan.
The NRC approved the exemption on November 29, 1993, and the licensee
,subsequently proces'sed the appropriate changes to its Physical Security Plan and permitted the removal of the photographic identification badges from the site on December 6,
1993.
The inspectors observed the implementation of this. system during their daily entrances and exits. from the plant site.
System
'mplementation was smooth and efficient.
Management Change On December 2,
1993, Hr. F.
E. Harcussen replaced Hr. F.
R.
Timmons as the Security Supervisor after Mr. F.
R. Timmons resigned.
Mr'. Harcussen was formerly the Security Systems Coordinator.'.2.3 Lack of Transient Combustible Permits On December 6,
1993, a dust mop in the Rad Waste Building Filling Room North (a locked high radiation area)
was ignited by a piece of slag blown through a penetration in the wall while mechanical maintenance personnel were performing arc gouging on a pipe.
(Refer to section 5.2.1 for additional information.)
In addition to the mop,'his room also contained 14 drums of activated carbon/water processing media (dry).
Licensee fire protection personnel and the NRC inspector on the scene noticed that these combustible chemistry supplies were stored in this room, without a transient combustible permit as required by procedure O-ADH-016.1, Transient Combustible and Flammable Substances Program.
As a result, the licensee generated condition report No.93-995 and inventoried all locked high radiation areas.
The licensee identified two additional areas within the Rad Waste Building containing transient combustibles which required permits in accordance with procedure 0-ADH-016. 1.
The transient combustibles within these areas belonged to HP, and they included a large
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,number of plastic and nylon bags for hot particle trash containing paper, plastic, muslin, etc.
The licensee subsequently issued transient combustible permits for all three locations.,
Longer-term corrective actions as documented in the condition report included retraining HP and chemistry personnel on procedure 0-ADM-
- 016.1 by January 15, 1994, and placement of materials associated with the HP permits in metal drum containers with lids and assignment of these areas as designated storage areas by January 31, 1994.
TS 6.8. l.h required that written procedures be established, implemented, and maintained covering the activities referenced in the Facility Fire Protection Program.
For the control of transient combustibles
.and flammable substances to help lessen the possibility of a fire or hazardous material release, paragraph 5. 1. 1. 1 of the Facility Fire Protection Program (procedure 0-ADH-016, Fire Protection Program) required that jobs involving transient combustible and flammable substances shall comply with the requirements'f procedure 0-ADH-016. I.
Enclosures 1 and 2 of procedure 0-ADM-016. 1 designated the Rad Waste Building as a
transient combustible control area, and paragraph 4. 1. 1 of this procedure defines Class A materials as ordinary combustible materials such as wood, cloth, paper, rubber, plastics, etc.
In addition, applicable portions of paragraph 5. 1.6 delineated that transient combustible permits shall be required and visually posted at the work location when there is an excess of 100 pounds of Class A materials.
The licensee identified that contrary to these requirements, on December 6 and 7, 1993, Class A transient combustible, chemistry and HP materials in quantities large enough to require permits were stored in three areas of the Rad Waste Building without transient combustible permits.
The inspectors reviewed condition report No.93-995 and verified that three transient combustibles permits (No.930056, 930057, and 930058)
had been generated.
The licensee promptly responded to the lack of a transient combustible permit, and the licensee's, corrective actions resulted in the identification of two additional problem areas.
The inspectors concluded the licensee's initial corrective actions were adequate.
This storage of transient combustible material without a required permit is a violation.
However, this violation will not be subject to enforcement action because the licensee's efforts in identifying and correcting the violation meet the criteria
'pecified in Section VII.B of the NRC Enforcement Policy.
This item will be tracked as NCV 50-250,251/
93-29-01, transient combustible chemistry and HP materials stored without transient combustible permits.
This item is close. 2.4 4.2.5 4.2.6 EP Annual Exercise The annual NRC evaluated EP exercise was conducted on December 15, 1993.
The inspectors participated in exercise as players and team evaluators.
Exercise results are documented in NRC Inspection Report No. 50-250,251/93-27.
Unit 3 RHR System Flush As discussed in section 7.2;1 of NRC Inspection Report No. 50-250, 251/93-26, high radiation levels occurred in the Unit 3 RHR system rooms during a shutdown initiated crud burst.
The licensee developed and approved a flushing procedure TP-1015, Unit 3 RHR System Flush From RWST Procedure.
The licensee then proceeded to perform sequential RHR system flushes on December 3, 17,"21, and 29, 1993.
Each flush was successful in reducing RHR system contamination levels and thus lowering RHR room radiation levels.
The licensee noted that radiation levels were reduced to normal expected levels.
The inspectors reviewed procedure TP-1015, attended the PNSC meeting that reviewed and approved the TP, witnessed flushing activities in the field, and reviewed selected before and after radiation survey information.
The inspectors noted that the PNSC and management personnel displayed a safety conscious attitude in requiring strict radiological controls during the flush and in assuring that RHR system operability was maintained.
Further, the inspectors noted very good coordination and team work during the flush among operators, system engineers, and HP personnel.
Procedure O-ADN-217, Conduct of Infrequently Performed Tests and Evolutions, was implemented including the use of a test director and pre-evolution checklist,and briefing.
Overtime Violation Two incidents involving operations personnel exceeding overtime requirements without management approval occurred during the month of December 1993.
The first incident involved a licensed operator who miscalculated his hours and worked greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48-hour period-.
The operator had worked two peak shifts, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> each, from ll:00 a.m.= to ll:00 p.m.
Following an 8-hour break, he traded shifts with a day-shift operator and worked a 12-hour day shift.
This resulted in the operator violating TS 6.2.2.g.2 which prohibits an individual from working greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48-hour period.
The other incident involved a non-licensed operator who worked a
mid shift and a day shift on'ne day and a day shift and peak shift the following day.
This resulted in his working 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> in a 48-hour perio As corrective action, a night order concerning overtime limits was issued.
Additionally, the licensee developed an, overtime documentation sheet which shows an, individual's work history for the past 7 days and the, future 7-day schedule.
Accepting overtime is the individual's responsibility.
The overtime documentation sheet should, in the inspector's opinion, help individuals to better calculate their worked hours prior to accepting offered overtime.
An overtime-related NCV (50-250,251/93-22-02)
due to a HP technician was identified on October 20, 1993.
Licensee's corrective action include'd disciplinary action and discussion of the event with the entire HP staff.
Based on low safety significance of the event, the fact that the two incidents were licensee identified and involved operations staff,- and the corrective actions including development of the overtime documentation sheet; this failure to meet the requirements of TS 6.2.2.g.2 will be classified as an 'NCV and will be tracked as NCV 50-250,251/93-29-02, exceeding overtime limits specified in TSs.
The criteria in Section VII.B of the Enforcement Pol,icy were satisfied.
This item is closed.
The inspectors will continue to monitor this issue to determine if corrective actions are adequate.to reasonably prevent future similar occurrences.
4.2.7 General Results As a result of routine plant tours and various operational observations, the inspectors determined that the general plant and system material 'conditions were satisfactorily maintained, the plant security program was effective, and the overall performance of plant operations was good.
5.0 Plant Events (93702)
5.1 Inspection Scope The following plant events were reviewed to determine facility status and the need for further followup action.
Plant parameters were evaluated during transient response.
The significance of the event was evaluated along with the performance of the appropriate safety systems and the actions taken by the licensee.
The inspectors verified that required notifications were made to the NRC.
Evaluations were performed relative to the need for additional NRC response to the event.
Additionally, the following issues were examined, as appropriate:
details regarding the cause of the event; event chronology; safety system performance; licensee compliance with approved procedures; radiological consequences, if any; and proposed corrective action.2 5.2.1 Inspection Findings Fire in Rad Waste Building On December 6,
1993, mechanical maintenance personnel were performing arc gouging on a pipe located in the Rad Waste Building Hopper area.
This pipe penetrated an opening in a wall and entered the Rad Waste Filling Room North which is a locked high radiation area.
During this arc gouging activity, a piece of slag was blown through the penetration, travelled approximately. 40 to 50 feet across the Rad Waste Filling Room North, and landed on a dust mop located at the opposite side of the room.
The mop ignited, and the mechanical maintenance personnel performing the
,work subsequently detected the fire.
The control room was notified at 10: 10 a.m.
The fire, which was confined to the mop, was extinguished by an HP technician using a hand held extinguisher.
The Fire Brigade responded but was not needed.
A contributing factor to this event was the licensee's mis-interpretation of step 5.3.2.3 of procedure O-ADH-016.5, Hot Work Program.
This procedural step referred to a designated fire watch whose only duty would be to observe work being performed, and it stated that a designated. fire watch guarding against fire shall be required when work is being performed in an area, on a barrier (e.g., walls, floors, tanks)
where direct penetration of sparks, slag, or heat transfer may introduce a fire hazard to the opposite side.
Designated fire watches had been provided in the Rad Waste Building Hopper area where the work was being performed and in the tool room below the work area.
However, on December 6,
1993, a designated fire watch had not been provided on the opposite side of the barrier in which'he hot work was being performed.
As a result of this event, all hot work was ceased until the licensee determined whether or not proper. safeguards were in place and whether or not proper housekeeping was maintained.
A fire incident report and condition report No.93-994 were also generated.
Prior to continuing with this particular hot work activity, the licensee installed shielding in the penetration where the piping was being cut, and a designated fire watch was placed on the opposite side of the barrier.
Additional corrective actions completed by the licensee included the retraining of mechanical maintenance personnel on fire watch requirements of procedure O-ADH-016.5, a walk down of current hot work activities in the plant to ensure procedural compliance with procedure 0-ADH-016.5, and the revision of procedure O-ADM-016.5 to clarify the locations for placement of designated fire watches under similar conditions.
The inspectors responded to the site of the fire as well as to the control room to monitor licensee actions.
Further, the inspectors reviewed the condition report, fire incident report, and
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5.2.2 procedural revisions.
It was noted that mechanical maintenance, HP, and Fire Brigade personnel responded to this event in a timely and efficient manner and that licensee corrective actions were prompt an'd effective.
However, in addition to the mop, the Rad Waste Filling Room North contained 14 drums of activated carbon/water processing media (dry).
Licensee fire protection personnel and the NRC inspector on the scene noticed that these combustible chemistry supplies were stored in the Rad Waste Filling Room North without a transient combustible permit.
(Refer to section 4.2.3 for additional information.)
4B EDG Failure At 1:09 a.m on December 23, 1993, the 4B EDG was removed from service for quarterly preventive maintenance prior to a scheduled surveillance test.
This placed the Units in 72-hour LCOs in accordance with action statements b and d of TS 3.8. 1. l.b due to a required EDG being inoperable and action statement f of TS 3.5.2.a due to the 4B HHSI pump not capable of being powered by its associated operable EDG (4B).
At 9: 15 a.m., during the performance of the required surveillance, operators observed the following phenomena approximately 6 times, each lasting approximately 5-10 seconds:
generator amps pegged high ()600 amps);
generator voltage increased by approximately 0-3 KV, generator VARs out increased from approximately 50 KVARs to 500 KVARs, exciter
=DC volts decreased from 60 VDC to 50 VDC, and exciter amps increased from 135 amps to 150 amps.
The licensee shut down the 4B EDG, maintained it out-of-service due to the valid failure, began troubleshooting activities, and made preparations to run the other required EDGs in accordance with the TSs and the surveillance procedure.
The licensee also consulted Division (the load dispatcher)
and verified that possible system grid disturbances were not occurring.
Action statement b of TS 3.8.1. I.b required the demonstration of the operability of the required startup transformers and their associated circuits per TS 4.8. l. 1. l.a within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, the demonstration of the operability of the remaining required EDGs per TS 4.8.1. 1.2.a.4 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the restoration of the inoperable EDG to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The startup transformers and their associated circuits were verified to be operable at 8:35 a.m.,
4:30 p.m.,
and 11:55 p.m.
Operability testing of the 4A EDG was completed at 2:00 p.m.,
and operability testing of the 3B EDG was completed at 7:00 p.m.
Troubleshooting revealed a problem with a remote gate firing circuit.
This circuit.was replaced, and the 4B EDG was tested and returned to service at 2:20 a.m.
on December 24, 1993.
The licensee generated condition report No. 93-1026 to followup on this failure and also placed the 4B EDG on a mandatory increased surveillance frequency per TS Table 4.8-1.
This EDG was successfully tested again at 4:45 a.m.
on December 29, 1993, and is currently scheduled to be tested weekly for the next 5 week The licensee is also currently in the process of generating a
Special Report to address this EDG valid failure.
The inspectors followed up on the,licensee's troubleshooting process and the status of the repairs and also reviewed the control room logs and condition report.
Troubleshooting and repair activities were conducted in an efficient manner, and the 4B EDG 'was returned. to service in a timely manner.
6.0 Maintenance Observations (62703)
6.1 Inspection Scope Station mainte'nance activities of safety-related systems and components were observed and reviewed to ascertain they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in conformance with the TSs.
6.2 6.2.1 The following items were considered during this review, as appropriate:
LCOs were met while components or systems were removed from service; approvals were obtained prior to initiating work; activities were accomplished using'approved procedures and were inspected as applicable; procedures used were adequate to control the activity; troubleshooting activities were controlled and repair records accurately reflected the maintenance performed; functional testing and/or calibrations were performed prior to returning components or systems to service; gC records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were properly implemented; gC hold points were established and observed where required; fire prevention controls were implemented; outside contr actor force activities were controlled in accordance with the approved gA program; and housekeeping was actively pursued.
Inspection Findings Maintenance Witnessed The inspectors witnessed/reviewed portions of the following maintenance activities in progress:
4A HHSI pump repair (procedure O-CMM-6.2.2),
Unit 4 OTBT and channel B troubleshooting (Refer to section 6.2.2 for additional information.),
and 3A and 3B HHSI motor vibration monitoring (Refer to section 8.2.2 for additional information).
6.2.2 For those maintenance activities observed, the inspectors determined that the activities were conducted in a satisfactory manner and that the work was properly performed in accordance with approved maintenance work orders.
Unit 4 OTET Channel Troubleshooting During the conduct of procedure TP-1017, Unit 4 Turbine Volumetric Flow Test (Refer to sections 7.2.1 and 11.2.2 for additional information.),
OTBT channel setpoint spiked low.
This resulted in a momentary OTBT RPS bistable trip.
This was confirmed by an annunciator alarm and by DDPS 'and ERDADS computer printouts.
The licensee stopped the test and entered procedure 4-0NOP-49.1, Deviation or Failure of Safety-Related or Reactor Protection System Channels.'he Channel 8 temperature-related bistables were placed in the tripped condition as required by TS Table 3.3-1.
I&C maintenance personnel initiated troubleshooting per work request No.
WR93020794, inspected the RPS temperature channel (Eagle-21 rack)
and all related OTBT inputs (e.g.,
pressure, NI',
hT), reviewed associated electrical drawings, and monitored channel operation using a multi-pen recorder.
These activities noted abnormal noise from one of the NI channels.
Detailed inspections in the NI cabinet found a loose connector.
The licensee repaired the connector, tested the channel, and placed the channel back into oper ation.
Subsequent performance of the channel has been normal.
The inspectors monitored troubleshooting activities in the control room and at the Eagle-21 rack.
The inspectors reviewed drawings, the ONOP, the work request, and observed operations and IKC activities.
The inspectors verified that TS requirements were met.
Further, the inspectors concluded that licensee actions were appropriate.
7.0 Surveillance Observations (61726)
7.1 7.2 Inspection Scope The inspectors observed TS required surveillance testing and verified that the test procedures conformed to the requirements of the TSs; testing was performed in accordance with adequate procedures; test instrumentation was calibrated; limiting conditions for operation were met; test results met acceptance criteria requirements and were reviewed by personnel other than the individual directing the test; deficiencies were identified, as appropriate, and were properly reviewed and resolved by management personnel; and system restoration was adequate.
For completed tests, the inspectors verified testing frequencies were met and tests were performed by qualified individuals.
Inspection Findings
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7.2. 1 Tests Observed The inspectors witnessed/reviewed portions of the following test activities:
'rocedure 3-0P-59.5, Power Range Nuclear Instrumentation Shift Checks and Daily Calibration; procedure 4-0P-59.5, Power Range Nuclear Instrumentation Shift Checks and Daily Calibration; procedure O-OSP-62.2, HHSI pump inservice testing (Refer to section 8.2.2 for additional information.);
procedure TP-1016, Unit 3 Turbine Volumetric Flow Test (Refer to section 11.2.2 for additional information.);
and procedure TP-1017, Unit 4 Turbine Volumetric Flow Test (Refer to section 11.2.2 for additional information.).
The inspectors determined that the above testing 'activities were performed in a satisfactory manner and met the requirements of the TSs.
8.0 Followup on Previous Items and Noncompliances (92702 and 92701)
8. 1 Inspection Scope A review was conducted of the following noncompliances and open items to assure that corrective actions were adequately implemented and resulted in conformance with regulatory requirements.
Verification of corrective action was achieved through record reviews, observation, and discussions with licensee personnel.
Licensee correspondence was evaluated to ensure the responses were"timely and corrective actions were implemented within the time periods specified in the reply.
8.2 Inspection Findings 8.2. 1 (Closed)
IFI 50-250,251/91-45-05, Miscellaneous Relay Racks Circuit Relays Separation from RPS Power Supply
'A report submitted by Westinghouse Electric Corporation pursuant to 10 'CFR Part 21 presented some possible generic problems with miscellaneous relay racks.
The report, which was issued to the site on July 18, 1991, stated that equipment usually referred to as miscellaneous relay racks furnished by Westinghouse was not originally intended to contain safety-related circuits.
However, it had come to the attention of Westinghouse that a few"plants were using a small percentage of the relays in these racks for safety-related functions.
The concern was that licensees may have erroneously assumed the miscellaneous relay racks met the same
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12 design criteria as safety-related racks furnished by Westinghouse.
The NSSS vendor advised licensees to review all relays in the miscellaneous racks and the associated circuits against their plant specific electrical isolation and separation criteria.
Within a short period of time after having received the Part
report, the Turkey Point plant was able to demonstrate that its miscellaneous relay racks including the relays were seismically qualified.
Therefore, it is possible to use the racks for safety-related functions as long as other relevant criteria are met.
In November 1991,, the date of,the last NRC inspection on this topic, the licensee had reviewed about 651 of the relays and circuits against its electrical isolation and separation criteria and had not reported any problems.
The engineering evaluation addressing the concerns of the Part
report was issued on Harch 17, 1992.
The title of that report is Safety Evaluation for Safety Functions in the Miscellaneous Relay Racks, JPN-PTN-SENP-91-006, Revision 3.
A total of 4 relay racks were evaluated (QR46 and QR47 for Units 3 and 4).
The
'ispositioning of 564 relays was reduced to 7 general cases as listed below:
(1)
The complete circuit is non-safety-related except for the fact that it has a safety-related power source.
The circuit power supply is derived from a vital AC power supply via the output from a channelized protection loop.
The comparator outputs are separately fused to provide isolation.
(2)
The complete circuit is non-safety-related except for the fact 'that it has a safety-related power source.
The circuit power supply is derived from a vital AC power supply via the secondary side of an ungrounded isolation transformer.
(3)
The complete circuit is non-safety-related except for the fact that it has a safety-related power source which is derived directly from a vital AC power panel.
Isolation is provided by a circuit breaker or fuse or both.
(4)
Similar to case (3) above except the panel branch circuit feeds safety-related circuits and circuits having no active safety function.
All the relays are classified as safety-related or quality-related.
Routing of cables outside the relay racks was reviewed to determine that wiring belonging to redundant safety-related systems or components was routed separately.
(5)
(6)
The complete circuit is safety-related.
Routing of cables was reviewed to determine that wiring belonging to redundant safety-related systems or components was routed separately.
Contacts from a non-safety related relay are used or wired into a safety-related circuit.
Failure modes of the non-
safety portion or the circuit were analyzed to determine that they would not defeat the safety function of the circuit.
Cable routing was reviewed to determine that wiring belonging to redundant safety-related systems or components was routed separately.'7)
The complete circuit is non-safety-related including the power supply.
The purpose of IFI 50-250,251/91-45-05 was that the inspector would review the licensee evaluation and dispositioning of the Part 21 report.
The inspector reviewed the licensee's Safety-Evaluation for Safety Functions in the Miscellaneous Relay Racks (JPN-PTN-SENP-91-006, Revision 3).
Based on this review, the inspector concluded that the criteria and methodology of the licensee's safety evaluation were adequate.
The conclusion of the safety evaluation is that the circuits related to relay 'racks 3QR46, 3QR47, 4QR48, and 4QR47 meet the site specific requirements for electrical isolation and separation.
The inspector found the safety evaluation to be adequate.
During the review process, the inspector requested and received clarification on certain details of the safety evaluation.
The evaluation itself does not define the safety classification QR as
.
applied to components.
Therefore, the licensee provided a
definition and explanation.
QR stands for quality related.
The total population of QR includes components that must meet certain design criteria due to the fact that they are electrically connected to a safety-related circuit (although they serve no active safety function)
and components that serve important non-safety-related functions such as in the fire protection system.
The total equipment data base indicates the basis for the safety classification.
Examples of classification basis information include relays connected in a safety-related circuit; seismic requirements, etc.
Thus, the total equipment data base defines the relevant design criteria for all the components.
Attachment B, pages 8 through 12, of the safety evaluation gave the analysis for the control circuits for motor operated valves 750 and 751.
A total of ten relays are involved, but the conclusion discussed only five relays.
The conclusion for HOV-750 and 751 circuits was revised (Revision 4, dated December 20, 1993)
to include all ten relays.
The inspector reviewed the additional information provided in Revision 4 and agreed that the design criteria were met for these circuits.
With regard to general case (4), it was not made clear within the safety evaluation whether cable routing had been reviewed.
The licensee stated they were uncertain as to whether the cable routings for circuits falling into the case (4) category had been reviewed.
The licensee performed the review of cable routing, concluded that the design criteria were met, and revised (Revision
4) the safety evaluation to clarify this matter.
This item is closed.
8.2.2 (Open)
IFI 50-250,251/93-26-03, SI Pump Motor Rotor Bar Cracking During this inspection period, the licensee continued with its testing and assessment of the HHSI pump motor rotor bar cracking issue.
This included special motor vibration and signature analysis testing during inservice testing conducted on December 7,
1993.
Two vendors performed this diagnostic testing using motor current readings versus frequency techniques and velocity signature analysis.
Both vendors independently concluded that there were no motor rotor bar cracks.
The licensee updated the applicable condition rep'ort (No.93-953),
and the PNSC subsequently reviewed the condition report.
This update included the results of the above mentioned diagnostic testing and a schedule and priority for motor replacements as follows:
HHSI Motor Date 4B 3B 3A January 31, 1994 March 31, 1994 July 31, 1994 The inspectors reviewed the revised condition report, attended the PNSC meeting, and observed the special testing and maintenance diagnostics.
(Refer to sections 6.2. 1 and 7.2. 1 for additional information.)
This IFI remains open pending completion of the HHSI pump motor replacement activities.
9.0 Onsite Followup and In-Office Review of Written Reports 9.1 Inspection Scope The reports discussed below were reviewed.
The inspectors verified that reporting requirements had been met, root cause analysis was performed, corrective actions appeared appropriate, and generic applicability had been considered.
Additionally, the inspectors verified the licensee had reviewed each event, corrective actions were implemented, responsibility for corrective actions not fully completed was clearly assigned, safety questions had been evaluated and resolved, and violations of regulations or TS conditions had been identified; When applicable, the criteria of 10 CFR Part 2, Appendix C, were applie '.2 9.2.1 Inspection Findings Monthly Operating Report The inspectors reviewed the November 1993 Monthly Operating Report and determined it to be complete and accurate.
9.2.2 Annual
CFR 50.59 Report The inspectors reviewed the annual
CFR 50.59 report dated November 18, 1993.
This report documents all safety evaluations performed during the period July 1, 1992, through June 1,
1993, as required by 10 CFR 50.59.b(2).
These safety evaluations are required for changes to procedures and to the facility as described in the FSAR.
In addition, the licensee documented PORV actuations per TS 6.9. 1.5 and steam generator tube inspections per TS 4.4.5.5 in this report.
The inspectors also discussed the report with selected licensee personnel, and concluded that the report was appropriate and met NRC requirements.
10.0 Self Assessment Capability (40500)
10.1 10.2 Inspection Scope The inspectors performed a review of the licensee's self assessment capability including PNSC and CNRB activities, gA/gC audits and reviews, line management self-assessments, individual self checking techniques, and performance indicators.
Inspection Findings 10.2. 1 PNSC Activities The inspectors attended several PNSC meetings during the period.
The inspectors verified that PNSC meetings met the requirements of applicable TSs and procedural guidance.
The inspectors noted a
good questioning attitude and safety-conscious perspective on the part of the PNSC members.
This was particularly noted during PNSC review of the temporary procedures for the Units 3 and 4 turbine
, volumetric flow tests (Refer to section 11.2.2 for additional information.)
and subsequent implementation, and during the review of the Unit 3 RHR system flush temporary procedure (Refer to section 4.2.5 'for additional information.).
10.2.2 gA Performance Monitoring of the Personnel Errors As a result of recent human errors,,plant management requested gA to perform a special PMON to evaluate condition reports involving personnel error,.
gA selected 23 condition reports that discussed N
~
~
events which occurred during 1993 and were partially attributed to personnel error.
The PHON (gAO-PTN-93-028)
was issued on November 23, 1993.
The
.licensee concluded that the condition report process, including root cause determination and corrective action implementation, was satisfactory.
However, some deficiencies were noted.
This included root cause training weaknesses that resulted in not identifying all of the caused factors, inconsistent use of the HPES process and infrequent self-evaluation by supervision and management involved in an event.
The PHON report recommended corrective actions'including condition report process revision and root cause training program enhancements.
Management is in the process of implementing these corrective actions.
The inspectors reviewed the PHON report and selected condition" reports.
The, inspectors concluded the gA audit was thorough and appeared effective in problem identification and corrective action recommendations.
The inspectors will continue to follow this area.
11.0 Design, Design Changes, and Modifications (37700, 37828)
.Inspection Scope The inspectors reviewed selected PC/Ms including the applicable
.safety evaluation, in-field walkdowns, as-built drawings, associated procedure changes and training, modification testing, and changes to maintenance programs.
11.2 11.2.1 Inspection Findings Relaxed Axial Offset Control Methodology TS Change.
- On November 12, 1993, the NRC issued license amendments Nos.
156 and 150 for Turkey Point Units 3 and 4, respectively.
These amendments consisted of changes to the TSs involving both implementation of a relaxed axial offset control methodology for axial flux difference control and relocation of cycle-specific parameter limits from the TSs to a Core Operating Limits Report.
The new TS amendments also required that all changes in cycle specific parameter limits be submitted to the NRC prior to operation with the new parameter limits.
As a result, the licensee provided the NRC with the new Core Operating Limits Report by letter dated December 28, 1993.
The curves in this report are applicable to,the current operating cycles of Unit 3 Cycle 13 and Unit 4 Cycle 14.
The existing constant axial offset control methodology maintained axial power distribution within a band of 15% axial flux
difference around a measured target value during normal plant operations (including power change maneuvers).
By controlling the axial flux difference, the constant axial offset control methodology limited the possible skewing of the axial xenon distribution and thereby minimized xenon oscillations and their effects on the power distribution.
On the other hand, the newly approved relaxed axial offset control methodology was developed to provide wider axial flux difference control bands at reduced power by utilizing core margin effectively, to provide more operating flexibilityat the plant by increasing the axial flux difference operating envelope, and to increase plant availability.
This flexibility allowed the operator to minimize and/or smooth the boron system duty relative to the'constant axial offset control
- methodology,.
It reduced control rod motion (corrections)
and hence operator action required to maintain, conformance with the power. distribution control TSs.
It also increased the ability to return the unit to power after a plant trip.
Although the TSs have been changed to allow more flexibility, Plant Management has retained the i5% axial flux difference around a measured target value as an administrative limit.
In order to address the modifications required to implement this new power distribution control strategy, the licensee developed and approved PC/M No.93-053, Annunciator Panel Modifications to Support the Relaxed Axial Offset Control (RAOC) Implementation.
The necessary changes included the replacement of annunciator plaques and the updating of the emergency response data acquisition system process computer program which provides alarms if the axial flux difference is outside the prescribed bands.
Annunciator panel G-5/1, DELTA FLUX >5% HAX POWER 90%,
has been changed to read AXIAL FLUX TS LIMIT EXCEEDED; and annunciator panel G-5/2 DELTA FLUX >5% HAX POWER 50%
has been changed to read AXIAL FLUX ADMIN LIMIT EXCEEDED.
The licensee effectively implemented these changes on December 8, 1993, and the engineering'taff briefed. the operating crews.
The licensee also issued Training Brief No.
438 and mentioned this TS change during the operators'equalification training.
In addition to these actions, the licensee revised various procedures to provide guidance if the administrative limit is exceeded and to provide operability checks of the new axial flux difference alarm.
The inspector reviewed the newly approved TS changes, the training brief, portions of the PC/M, and'he implementation of the changes in the control room.
The inspector concluded that these changes were effectively implemented.
11.2.2 Turbine Volumetric Flow Test FPL is pursuing a facility design change to increase the power output of each unit from 2200 to 2300 megawatts thermal.
In preparation for this, the licensee performed turbine volumetric flow tests on both units.
These tests provided data on steam flow
and turbine control valve position as a function of reduced primary temperature (Tave)
and secondary steam pressure.
Thus, the licensee determined the steam flow and pressure necessary to open the final (fourth) control valve.
The licensee, in conjunction with Westinghouse, performed a safety evaluation (JPN-PTN-SEEP-93-037)
which addressed the effects of reducing Tave by 16'F'for a duration of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
The safety evaluation concluded that the conduct of this test on both units did not constitute an unreviewed safety question, and did not require any TS changes.
The safety evaluation addressed the design basis accidents and transient conditions and assumptions; the effects.on plant systems, structures and components; and the radiological and dose analyses.
The PNSC reviewed and approved the safety evaluation and the applicable temporary procedures (procedure TP-1016, Unit 3 Turbine Volumetric Flow Test, and procedure TP-1017, Unit 4 Turbine Volumetric Flow Test).
(Refer to section 7.2.1 for additional information.)
Further, the licensee. considered the conduct of these tests as a special evolution and therefore implemented procedure O-ADH-217, Conduct of Infrequently Performed Tests and Evolutions.
This required appointment of a test director and test specialist, conduct of a pre-evolution briefing, and completion of the procedure checklist.
The license performed these above requirements and successfully conducted the test for Unit 4 on December 13 and for Unit 3 on December 14, 1993.
During the Unit 4 test, an RPS channel setpoint spike resulted in temporarily stopping test performance.
(Refer to section 6.2.2 for additional information.)
The inspectors reviewed the safety evaluation, procedures TP-1016 and TP-1017, the procedure 0-ADM-217 checklist,- and other related documentation.
The inspectors also discussed these tests with engineering personnel, attended the PNSC meeting which reviewed and approved the testing, and observed test conduct in the field and in the control room.
This included observing the pre-evolution briefings, the test conduct, the PNSC meeting for the test hold points, and the final test review.
The inspectors noted that the test involved personnel from IKC, engineering, system engineering, reactor engineering, on-shift operators, and the STA.
Further, good involvement by a test director and a test specialist, as well as plant, site, and engineering management oversight were noted.
The inspectors concluded that the safety evaluation and test procedures addressed the issues, including appropriate precautions.
Further, test.preparation, conduct, and review were performed conservatively and deliberately.
The test was coordinated well, with excellent management involvement and oversight.
management demonstrated conservative and cautious test approach as evidenced by stopping the tests at a 8'F Tave decrease
due to concerns with secondary plant stability.
The inspector also noted excellent pre-test briefings by the NPS and very good control communications and conduct among test participants.
11.2.3 Vendor Testing Deviations in Certain Mire and Cable Products Supplied to FPL Including Turkey Point On October 22, 1993, Teledyne Thermatics notified the NRC of the existence of possible testing deviations in certain wire and cable products supplied to licensees including Turkey Point.
The notification was made pursuant to
CFR 21.21(c)(3)(1),
and it emphasized 'that at that time, it was not clear=whether the testing discrepancies had resulted in defects within the meaning of the
.applicable regulations.
Additionally, Teledyne Thermatics had no reason to believe that'there.had been any adverse effects on the quality or reliability of the wire and cable products supplied.
On November 23, 1993, Teledyne Thermatics requested FPL assistance in determining where and how the cable and wire products were used at FPL.
Consequently, the licensee initiated a condition report to resolve the concern and directed engineering to perform an operability assessment associated with the issue.
The licensee determined that the subject wire and cable products supplied by Teledyne Thermatics are used extensively at Turkey Point as well as St.
Lucie in 600V AC and DC applications.
The cables'ost critical application is as jumper/extension wire in Eg qualified NOVs.
Turkey Point engineering evaluated testing information pertaining to the cable and wire products supplied by Teledyne Thermatics.
This testing included thermal aging,'radiation exposure, LOCA simulation and post-LOCA cooldown, and post-LOCA tests which included insulation resistance measurements, water immersion, and high pot testing.
Engineering concluded that all applicable testing required to be performed on cable and wire products supplied to Turkey Point had been performed.
Additionally, engineering concluded that the testing discrepancy associated with certain cables supplied by Teledyne Thermatics was not applicable to cable purchased for use at Turkey Point.
Coincident-with the potential
CFR Part 21 notification, FPL had performed.a gA audit at Teledyne Thermatics.
The gA audit concluded that Teledyne Thermatics did not have an adequate gA program.
Consequently, Teledyne Thermatics had been removed from the FPL vendor list.
However, the FPL gA findings did not apply to the fore-mentioned cable and wire testing issue.
This issue was discussed with the NRC regional staff.
The inspector concluded that the licensee's actions associated with the issue were conservativ Exit Interview The inspection scope and findings were summarized during management interviews held throughout the reporting period with the Plant General Manager and selected members of his staff.
An exit meeting was conducted on January 5,
1994.
The areas requiring management attention were reviewed.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
Dissenting comments were not received from the licensee.
The inspectors had the following findings:
Item Number Descri tion and Reference 50-250,251/93-29-01 NCV - combustible chemistry and HP materials stored without transient combustible permits (section 4.2.3).
50-250,251/93-29-02 NCV - exceeding overtime limits specified in the TSs (section 4.2.6).
Acronyms AC ADM amp CFR CMM CNRB 4F bT DDPS DC EDG EP ERDADS EQ FPL FSAR HHSI
'HP HPES IKC IFI JPN KV KYAR LCO LOCA MOV NCV NI NPS and Abbreviations Alternating Current Administrative Ampere Code of Federal Regulations Corrective Maintenance
- Mechanical Company Nuclear Review Board Degrees Fahrenheit Delta Temperature Digital Data Process System Direct Current Emergency Diesel Generator Emergency Preparedness Emergency Response Data Acquisition Display System Environmental Qualification Florida Power and Light Final Safety Analysis Report High Head Safety Injection Health Physics Human Performance Evaluation System Instrumentation and Control Inspector Followup Item Juno Project Nuclear Kilovolt Kilovolts Amperes Reactive Limiting Condition for Operation Loss-of-Coolant Accident Motor-Operated Valve Non-Cited Violation Nuclear Instrumentation Nuclear Plant Supervisor
NRC NSSS ONOP OP OSP OTBT PC/H PHON PNSC PORV PTN QA QAO QC QR RHR ROAC RPS RMST SI STA Tave TP TS V
VAR VDC MR
Nuclear Regulatory Commission Nuclear Steam Supply System Off Normal Operating Procedure Operating Procedure, Operations Surveillance Procedure Over Temperature Delta Temperature Plant Change/Hodification Performance Honitoring Plant Nuclear Safety Committee Power Operated Relief Valve Project Turkey Nuclear Quality Assurance Quality Assurance Organization Quality Control Quality Related Residual 'Heat Removal Relaxed Axial Offset Control Reactor Protective System Refueling Mater Storage Tank Safety Injection Shift Technical Advisor Temperature Average Temporary Procedure Technical Specification Volt Volts Amperes Reactive Volts Direct Current Mork Request
P