IR 05000250/1988040
| ML17347A974 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/10/1989 |
| From: | Butcher R, Crlenjak R, Mcelhinney T, Schnebli G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17347A973 | List: |
| References | |
| 50-250-88-40, 50-251-88-40, NUDOCS 8902270494 | |
| Download: ML17347A974 (17) | |
Text
UNITED STATES NUCLEAR REGULATORY, COMMISSION
REGION II
101 MARIETTASTREET, N.VP.
ATLANTA,GEORGIA 30323 qt ~ R~CO Ci+
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Report Nos.: '0-250/88-40 and 50-251/88-40 Licensee:
Flori.da Power'nd Light Company'250'West Flagler Street
" Miami, FL 33102 Docket Nos.:
50-250 and 50-251 License Nos.:
DPR-31 and DPR-41 Facility Name:
Turkey Point 3 and
- Inspection Conducted:
December 23, 1988 through January,27, 1989 Inspectors: M~
K..~e. ~.
R.
G. Butcher, Senior Resident n'spector Da e
igned C-T.
F. McElhinney, Resident Ins ector G. A. Sc e
i eside t Inspec or Approved by: '.
. Cr enjak, ction ief Division of Reactor Projects Dat S gned Dat S gne
~i~8>
Date igned SUMMARY Scope This routi.ne resident inspector inspection entailed direct inspection at the site in the areas of monthly survei.llance observations, monthly maintenance observations, engineered safety features walkdowns, operational safety, plant events, management meeting and meeting with local officials.
~t e
Results Three Inspector Followup Items were identified:
Licensee's corrective actions regarding draining Unit 4 to the point of RHR pump cavitation; provide permanent labeling for RHR system reach rod mechanisms; and followup lic'ensee's determination of the cause for leakage through seal table conduit at locations
..J7 and J12.
One concern was expressed to licensee management regarding the Unit 4 operators relying solely on visual observation of water level with other level indicators availabl REPORT DETAILS Persons Contacted Li
- J J.
L.
- J
- R T.
- R.
- K.
- D
- S R.
J.
V.
J.
R.
J.
- L
- G
- F
- R J.
D.
M.
J.
W.
Supervisor Maintenance censee Employees W Anderson, guality Assurance Supervisor Arias, Regulation and Compliance Supervisor W. Bladow, equality Assurance Superintendent E. Cross, Plant Manager-Nuclear J. Earl, guality Control Supervisor A. Finn, Training Supervisor J. Gianfrancesco, Nai,ntenance Supervisor W. Gross, Regulation and Compliance W. Haase, Safety Engineering Group T. Hale, Engineering Project Supervisor D. Hart, Regulation and Compliance Engineer W. Kappes, Maintenance Superintendent A. Kaminskas, Reactor Engineering Supervisor A. Labarraque, Senior Technical Advisor G. Mende, Operations Supervisor S.
Odom, Site Vice President W. Pearce, Operations Superintendent M. Smith, Services Manager-nuclear H. Southworth, Technical Department Supervisor J.
Stevens, Manager, Plant Licensing C. Strong, Mechanical Department Supervisor Tomaszewski, Instrument and Control Department B. Wayland, Electrical Department Supervisor D. Webb, Operations
- Maintenance Coordinator R. Williams, Assistant Superintendent Planned Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, mechanics, and electricians.
~Attended exit interview on January 27, 1989.
Followup on Items of Noncompliance (92702)
A review was conducted of the following noncompliances to assure that corrective actions were adequately implementedand 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 that the responses were timely and that corrective actions were implemented within the. time periods specified in the reply.
(Closed)
Violation 50-250,251/88-02-01.
This violation involved the failure to meet the requirements of Technical Specification (TS) 6.8. 1, three examples.
The inspectors reviewed the licensee's corrective actions and found them to be complete.
Finding A. 1 involved a violation of the caution tag clearance procedure.
Inspection report 50-250,251/88-39
identified numerous examples of clearance proc'edure violations.
Violation 50-250,251/88-02-01 is closed.
However the licensee's corrective actions for clearance problems will be. followed by the inspectors and tracked via violation 50-250,251/88-39-01.
Followup on Inspector Followup Items (IFIs),
Inspection and Enforcement Information Notices ( IENs),
IE Bulletins ( IEBs) (information only),
IE Circulars '(IECs),
and NRC Requests (92701).
(Closed)
Inspector Followup Item 50-250,251/88-39-03, licensee's resolution of the root cause for the cracked flan'ges found in the Intake Cooling Water ( ICW) system.
The licensee performed a visual inspection of.
72 ca'st iron threaded/screwed flanges on the Unit 3 ICW system and six flanges were determined to contain cracks.
Five of the six flanges contained cracks in the hub and were replaced.
Four were replaced with carbon steel flanges in accordance with PC/Yi 88-574, which included the appropriate
CFR 50.59 evaluation.
The fifth flange was replaced with a cast iron flange of the same class and material and -was inspected to verify the absence of cracks.
The sixth flange was evaluated and determined to be acceptable in its as-found condition per NCR-C-0640-88.
This NCR was reviewed by both resident and regional inspectors and was found to be adequate.
The remaining 66 flanges in the system showed no evidence of cracking, The root cause of the cracking, based on fracture mechanics analysis, was determined to be stored interference fit stress in combination with hydraulic transients experienced by the system.
The licensee considers that the remaining 66 flanges have already experienced these stresses'hroughout the life of the plant and that they are not large enough to initiate new cracking.
This item is closed.
Onsite Followup and In-Office Review of Written Reports of Nonroutine Events (92700/90712)
The Licensee Event Reports (LERs)
discussed below were reviewed and closed.
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 that 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 2, Appendix C, were applied.
(Closed)
Three Intake Cooling Water ( ICW) Pumps Inoperable Upon
'ICW Pump A and Emergency Diesel Generator B Being Out of Service Concurrently.
The inspector's investigation into this event resulted in violation 50-250,251/88-14-01.
The licensee's corrective actions will be tracked by the inspectors via the violation.
LER 250/88-07 is close.
Monthly Surveillance Observations (61726)
The inspect'ors observed TS required surveillance testing and verified:
That the test procedure conformed to the requirements of the TS, that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation (LCO) were met, that test results met acceptance criteria requirements and 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 inspectors verified that testing frequencies were met and tests were performed by qualified individuals.
The inspectors witnessed/reviewed portions of the following te'st activities:
~,
3-OSP-055.
Emergency Containment Cooler Operability Test 0-OSP-23.
Emergency Diesel Generator Operability Test Emergency Diesel "Generator Eight Hour Full Load Test and Load Rejection No violations or deviations were identified within the areas inspected.
Monthly Maintenance Observations (62703)
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 TS.
The following items were considered during this review, as appropriate:
that LCOs were met while components or systems were removed from service; that approvals were obtained pr>ol 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 troubleshooting 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 gC 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 gC hold points were established and observed where required; that fire prevention controls were implemented; that outside contractor activities were controlled in accordance with the approved gA program; and that housekeeping was actively pursued.
The inspectors witnessed/reviewed portions of the following maintenance activities in progress:
Replacement of Unit 3 ICW cast iron flanges due to crackin.
, Replacement of Unit, 4 CCW Heat Exchangers.
-
- 'eplacement of 3B RHR Pump Seal.
Overhaul of "B" Emergency Diesel Generator Upper End.
Troubleshooting 3A ICW Pump Failure.'o violations or deviations were identified in the areas inspected.
Operational Safety Verification (71707')
The inspectors observed control room operations, reviewed applicable logs,
.
conducted discussions with control room operators, observed shift turnovers and confirmed 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 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.
Plant housekeeping/cleanliness conditions arid implementation of radiological controls were 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.
The inspectors walked down accessible portions of the following safety related systems to verify operability and proper valve/switch alignment:
'
and B Emergency Diesel Generators Control Room Vertical Panels and Safeguards Racks Intake Cooling Water Structure 4160 Volt Buses and 480 Volt Load and Motor Control Centers Unit 3 and 4 Feedwater Platforms Unit 3 and 4 Condensate Storage Tank Area Auxiliary Feedwater Area Unit 3 and 4 Main Steam Platforms No violations or deviations were identified in the areas inspected.
8.
Plant Events (93702)
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'ppropriate safety systems and the J
Cl
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'ollowing 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 actions.
- On January 4,
1989, the licensee reported a significant event to the NRC in accordance with 10 CFR 50.72(b)(2)(iii).
This was due to the feedwater bypass control valve (3-. 479)
fuses b'eing removed on a
clearance in November 1987 and not being reinstalled when the clearance was released on November 5,
1987.
On December 2,
1988, during the Unit 3 integrated safeguards test, this valve did not close as designed and subsequent investigation revealed that the fuses were not installed.
This event was cited as a violation (50-250,251/88-39-01)
of the licensee's clearance procedure.
The licensee reviewed this event to determine potential reportabi lity and concluded that no specific requirements applied.
The inspectors reviewed this'vent and determined that a conditio'n existed where, -with the Standby Steam Generator Feedwater Pumps (SBSG FW pumps)
operating and valve 3-479 unable to isolate in response to a Feedwater Isolation-Signal (FWIS), "a situation existed where a single condition alone would have prevented the fulfillment of a safety function.
The licensee investigated this scenario and initially determined that the plant had operated in 'a condition outside of its design basis when the SBSG FW pumps were used.
The Turkey Point Safety Analysis takes credit for redundant feedwater isolation by closing the feedwater control valves, tripping the Main Feedwater (MFW)
pumps and closing the MFW pumps discharge motor operated valve (MOV), in response to a
FWIS.
The SBSG FW pumps were installed in July 1985, under Plant Change/Modification (PC/M)82-296.
The purpose of these pumps is to provide feedwater for normal plant startup, cooldown, and hot standby operation.
These pumps allow the Auxiliary Feedwater (AFW) pumps to be used only for emergency conditions.
The safety evaluation for the PC/M concluded that operation of the SBSG FW pumps does not affect any accidents analyzed in the Final Safety Analysis Report (FSAR).
However, the design of the system did not address response to a
FWIS, therefore, redundant feedwater isolation was not provided while operating these pumps.
The licensee also concluded that during the period that the fuses for valve 3-479 were removed, the SBSG FW pumps were used for Unit 3 startup.
Therefore, during operation of the pumps, the FWIS would not have performed its safety function in that feedwater flow would not have been terminated.
The licensee evaluated the concern and the results showed that both core and containment responses were within their acceptance limits.
Based on this Safety Evaluation Report, dated January 9,
1989, operation at low power using the SBSG FW pumps does not represent an unreviewed safety questio On January 7,
1989, at 8:10 a.m'.,
while in'he process. of partially releasing clearance 3-89-01-001, the Unit 3. Reactor Control Operator (RCO)
'noticed that pressurizer level was decreasing.
Also, fire alarm point 21',
~
, Unit 3 Residual Heat Removal (RHR),heat exchanger, and pump rooms, had alarmed.
The RCO had just completed valving in the 3B RHR pump.
The Nuclear Operators (NOs) then reported steam in the RHR pump room.
The RCO instructed the NO to close the discharge valve he had just opened and also
.
to close the discharge drain valve 3-767B.-
The NO noticed there was no reach rod labeled 3-767B but found two reach rods labeled 3-898N (one of
.which had just been opened).
The NO closed the 3-898N.that had just been
.opened and the leakage stopped.
The Assistant Plant Supervisor Nuclear (APSN)
and the NO yerifi,ed that the 3-898N reach rod that was opened was actually. connected to valve 3-767B."
This verification w'as made by moving the reach rod and, visually verifying which valve actually moved.
The 3-767B and 3-.898N valves were correctly labeled at the valve.
The APSN corrected the labeling deficiency at the reach rod.
Valve 3-898N -is on the interdisc relief path for valve MOV-860B and valve 3-767B is a
3B RHR
.pump discharge line drain valve.
The licensee has verified correct labeling for all RHR reach rods for Units 3 and 4.
No other mislabeling problems were found.
Although the licensee
.labeled the valve at the reach rod, the reach rod mechanism is on the auxiliary building floor and the labels would not suffice for permanent use.
The licensee discussed various methods that could be used to permanently identify the reach rod mechanism.
Misidentification of valve 3-767B was the root cause for this event.
The licensee was recently issued a violation for labeling deficiencies in Inspection Report 50-250,251/88-3 The licensee's corrective actions for this event will be followed as part.of their corrective actions in response to violation 50-250,251/88-34-01.
The permanent labeling of the reach rod mechanisms will be identified as Inspector Followup Item 50-250,251/88-40-02.
The Pressurizer level decreased from about 22 percent to approximately
percent during this event.
The level decrease occurred over approximately a
seven minute time period.
The NO had secured let down and started a
second charging pump during this time.
This event met the licensee's Emergency Procedure 20101, Duties of Emergency Coordinator, Table
(Emergency Classification Table),
for declaring an Alert (reactor coolant system leakage greater'han 50 gpm);
Based on discussions with the Senior Resident Inspector, the licensee determined that it would not be prudent to declare an alert and activate the emergency plan since the event was terminated before they could act.
The event was determined reportable per
CFR 50.72(a)(3)
at a later time and reporting was delayed until 1:28 a.m.
on January 8,
1989, due to miscommunication.
The failure to promptly report this event is being followed up by a Region II inspection and will be discussed in Inspecti.on Report 50-250,251/89-02.
On January 10, 1989, at 1;40 a.m
, with Unit 3 in Mode 5.and Unit 4 defueled, the licensee reported a significant event in that the Emergency Notification System (ENS)
phone was out of service.
The licensee contacted the telephone company to effect repairs and the phone was returned to servic On January 10, 1989, at 2:50 a.m., with Unit 3 in Mode
and Unit 4 defueled, the licensee notified the NRC of a significant event in accord-ance with
CFR 50.72(b)(2)(iii)(B).
While the:
B Emergency Diesel Generator was being tested prior to returning it to service after over-haul, the 3A ICW'ump was taken out of service due to apparent excessive noise and vibration.
Due to the electrical distribution at the plant the emergency backup power was not available for the 3B and 3C ICW pumps as they are powered from the B diesel.
When the 3A ICW pump was taken out of service, both loops of RHR would have been inoperable, the A.loop due to the 3A ICW pump and the B loop due to the B diesel.
The licensee considered the event reportable because the event or condition could have prevented the fulfillment of the safety function of a
system needed to remove residual heat.
Unit 4 was not affected because the unit was defueled.
-The inspectors witnessed subsequent testing of the 3A ICW pump during troubleshooting which'ncluded vibration analysis, and observation of motor bearing temperatures, motor amps, and pump discharge pressure using various pump combinations.
The testing did not identify any problems with the pump or motor, and no abnormal noises or vibrations were noted by the inspectors during pump operation.
The licensee returned the 3A ICW pump to service at 6:30 p.m.
on January 10, 1989, and has PWOs and procedures ready to repair the pump should failure occur in the future.
In addition, the licensee is monitoring the pump on an increased basis to determine if a problem exists by observing the pump during operation and'monitoring the bearing temperatures on a shift basis and performing vibration analysis on a daily basis.
On January 16, 1989, at 4:45 p.m.,
the licensee found a slight leak on Unit 3 (approximately 18 drops per minute)
through the J12 incore flux mapping system seal table conduit.
This leak was non-isolable.
The reactor was in Mode 3 at 547 degrees F and 2335 psig, which is 100 psig above normal pressure, for an overpressure test.
The licensee reported this event under
CFR 50.72(b)( l)(i)(A).
The licensee commenced a
cooldown at 7: 15 p.m.
on January 16, 1989, to effect a repair on the J12 stub tube.
Cooldown to cold shutdown is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per TS 3.1.3.c.
Inspection by the licensee determined that two seal table conduits (J7 and J12)
were Leaking and the licensee is cutting out the leaking conduit.
Followup on the licensee's determination of the cause of the conduit leakage problem will be identified as Inspector Followup Item 50-250,251/88-40-03.
On January 19, 1989, at approximately 3:39 a.m.,
the operators commenced draining the Unit 4 upper refueling cavity in accordance with 4-0P-201, Filling/Draining the Refueling Cavity and the SFP Transfer Canal, revision dated January 3,
1989.
Section 6.1 of this procedure provides instructions for draining the cavity via the Residual Heat Removal (RHR)
system to the Refueling Water Storage Tank (RWST).
Prior to performing this evolution an On The Spot Change (OTSC)
was made to the procedure.
This OTSC (6814)
allowed draining the cavity using only the 4A RHR pump, due to the 4B pump being out of service.
The other pump was not needed for decay heat removal since the core was off-loaded and stored in the
Spent Fuel Pit.
During the evolution, operators were stationed in the following locations:
At the 58 foot level inside containment for visual level indication; at. the RHR pump to monitor differential pressure (dp);
at valve 4-887, RHR pump discharge to RWST; and in the control room to monitor RHR pump motor amps for indication of cavitation.
At approximately 5:48 a.m.
the Reactor Control Operator (RCO) stopped the 4A RHR pump due to swinging amps.
The Nuclear Operator (NO) inside containment indicated that the cavity level was approximately one foot above the reactor vessel flange.
Based on this information', verifying
=
sufficient water level, the RCO restarted the 4A RHR pump at 5:53 a.m.. At 5:57 a.m.,
the pump amps started,to swing, and the RCO secured the pump.
The NO inside containment indicated that the cavity water level was still 6 to 8 inches above the reactor vessel flange.. The operators then checked the level hose inside containment and the level indicator (LI 6421)
located in the control room.
Both of these indicated that the water level was in the mid-nozzle area, where pump cavitation is likely to occur.
The evolution was stopped and the licensee conducted an investigation into this event.
The inspectors expressed a concern to the licensee regarding the operators relying solely on visual level observation with other level indicators available.
(i.e. level hose, LI 6421, pressurizer level).
The NRC is very concerned with loss of decay heat removal as evidenced by Generic Letter 88-17, dated October 17, 1988, which requested licensees respond to specific recommendations concerning operation during shutdown cooling.
These recommendations apply whenever there is irradiated fuel in the reactor vessel.
Although there was no fuel in the reactor vessel, the licensee treated this 'event very seriously.
The results of the licensee's investigation along with the implementation of corrective actions will be reviewed by the inspectors and tracked via Inspector Followup Item 50-250,251/88-40-01.
On Jan'uary 20, 1989, at 4:00 a.m.,
a security guard was observed sleeping at her post.
The licensee notified the NRC-at 8:50 a.m. that same day.
The licensee is conducting an investigation of this event and it will be followed up by Region II security inspectors.
On January 24, 1989, at 12:53 a.m.,
the licensee notified the NRC of a control room ventilation system actuation from normal to recirculation mode.
While performing a
check of the control room normal air intake radiation monitor channel B (RAI-6642), the Reactor Operator (RO) placed the selector switch to check, received the expected upward deflection of the meter and amber trip light.
The RO then began 'to slowly release the selector switch, in accordance with a posted caution tag, back to operate when the control room ventilation system automatically swapped to recirculation mode.
A significant event was declared per
CFR 50.72(b)(2)(ii).
RAI-6642 was placed out of service for troubleshooting.
Unit
was in Mode 5 (cold shutdown)
and Unit 4 was defueled.
The licensee has reported a similar inadvertent actuation previously by LER 250/88-20.
The resident inspectors will followup on the licensee's LERs.
No violations or deviations were identified within the areas inspected.
Management Meeti'ng (94702)
e 10.
On January 12, 1989, the monthly NRC/FPL Management Meeting was conducted at the site.
The meeting was attended by NRC Regional and Headquarters Management and FPL Site and Corporate Management.
'he topics of
'iscussion included:
Plant Status; Drawing Update Program Status; Procurement and Spare Parts Deficiencies; Revised TS Status; Enhancements to Improve Configuration Control; PEP Program Status; Security Initiatives; New Employee Concerns Program and the Licensee's Fitness for Duty Program.
A Maintenance Inspection Team exit was also conducted and will be discussed in Inspection Report 50-'250,251/88-32.
Information Meeting with Local Officials (94600)
On January 23, 1989, the Resident -Inspectors, Region II Projects Section 2B Section Chief, and'
Region II State and Governmental Affairs Representative met with Homestead and Florida City local officials.
Also present were representatives from the Dade County Emergency Management Office, Monroe County Civil Defense Office and FPL representatives including the Plant Manager and Site Vice President.
Attendees were presented a handout illustrating the oral presentation which consisted of the following:
Introduction of NRC personnel; NRC Mission and Statutory Authority; NRC Organization; NRC Functions and telephone numbers for contacting NRC Region II, Headquarters or Resident Inspectors.
A short question and answer period followed.
.The USNRC Local Public Document Room at Florida International University was toured and found to be in satisfactory condition.
Exit Interview (30703)
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 January 27, 1989.
The areas requiring managemen.
attention were reviewed.
No proprietary information was provi'ded to the inspectors during the reporting period.
The inspectors had the following findings:
Inspector Followup Item 50-250,251/88-40-01.
Licensee's corrective actions regarding draining Unit 4 to the point of RHR pump cavitation.
(paragraph 8)
Inspector Followup Item 50-250,251/88-40-02.
Provide permanent labeling for RHR system reach rod mechanisms.
(paragraph 8)
Inspector Followup Item 50-250,251/88-40-03.
Followup on licensee's determination of the cause for leakage thru seal table conduit at J7 and J12.
(paragraph 8)