IR 05000250/1981002
| ML17340B313 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/16/1981 |
| From: | Dance H, Ignatonis A, Marsh W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17340B297 | List: |
| References | |
| 50-250-81-02, 50-250-81-2, 50-251-81-02, 50-251-81-2, NUDOCS 8106080449 | |
| Download: ML17340B313 (17) | |
Text
~7
~S RE0y P
~
A.0O I
O~
YJ
+ip qP
+**++
UNITED STATES NUCLEAR REGULATORY COMMISSION ROON II 101 MARIETTAST., N.W., SUITE 3100 ATLANTA,GEORGIA 30303 Report Nos. 50-250/81-02 and 50-251/81-02 Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33101 Facility Name:
Turkey Point Docket Nos. 50-250 and 50-251 License Nos.
DPR-31 and DPR-41 Inspection at Turkey Point site near Homestead, Florida Inspectors:'.
J.
gn toni W. C.
Ma s
ra r
Date Signed
~lie Date igne Approved by:
H.
D e, Chief SUMMARY'nch Da e Si ned Inspection on. January 1-28,, 1981:-
Areas Inspected This'outine inspection involved 120 resident inspector-hours on site in the areas-of (1) followup on previous inspection findings;,,(2) fol,lowup on licensee.
event reports; (3) surveillance test observations; (4) inspection applied to IE Bulletin 80-24 and followup on licensee's response to IE Bulletin 79-21; (5)
review of Emergency Procedures for adequacy in dealing with an ATWS event; (6)
followup of implementation of post-TMI requirements in accordance with NUREG-0737; (7) plant operations; and (8) plant tours.
Results Of the eight areas inspected, no apparent'iolations or devi'ations were identi-fied in six areas; one apparent violations was found in one area (Violation-inadequate procedures which caused inadvertent safety injection system actuation
- paragraph 10; one deviation was found in one area (Deviation - failure to perform surveillance test on control switches within the commited time interval in response to IE Bulletin 80-20 -paragraph 3).
s 10S08 0'W%
DETAILS Persons Contacted Licensee Employees
- H. E. Yaeger, Site Manager'J.
K. Hays, Plant Manager - Nuclear J.
E. Moore, Operations Superintendent - Nuclear V. B. Wager, Operations. Supervisor D. W. Haase, Technical Department Supervisor J. Wade, Chemistry Supervisor P.
W. Hughes, Health Physics Supervisor J.
P. Mendietta, Maintenance Superintendent J.
Lowman, 1&C Department Supervisor
R. E. Garrett, Plant Security Supervisor W. R. Williams, Assistant Superintendent Electric Maintenance
- D. W: Jones, QC Supervi.sor Other licensee employees contacted included technicians, operators, security force members, and office personnel.
NRC Resident Inspectors
- A'. J. Ignatonis
- W. C. Marsh
- Attended,exit interview 2.
Exit. Interview The inspection scope and findings were summarized on February 9, 1981 with-those persons indicated in Paragraph 1 above.
The site and plant managers acknowledge the stated violation and deviation.
3.
Licensee Action on Previous Inspection Findings (Cl osed)
Unreso1 ved Item SO-251/80-34-01, Fai lure to perform timely inspection of'ype W-2 Spring return Switches in response to IEB 80-20.
Based on further review and discussions with the regional office inspection personnel, this Unresolved.
Item has been changed to a Deviation.
The licensee missed their own commitment date in performing continuity testing on the safety related control switches of. Unit 4 in response to Item 2 of IEB 80-20.
The licensee had committed to perform quarterly inspections (every 92 days)
on the subject control switches, contrary to the NRC require-
,ment of every 31 days after the initial test and after each manipulation of the switch per Item 2 of IEB 80-20.
More than ninety-two days elapsed before a continuity test was repeated on the control switch for Unit 4.
This constitutes a deviation from a commitment made to the NRC.
Per 'recent discussion with the licensee Assistant Superintendent of Electrical
5.
Maintenance, the inspector has been informed that the continuity tests on
'the control switches have been repeated on both Units 3 and 4 on January 7,
1981.
The licensee is preparing a supplemental response to this bulletin to address a more frequent surveillance.
Unresolved Items Unresolved items were not identified during. this inspection.
New Unresolved Items No new unresolved items were identified during this inspection period.
6.
Licensee Event Report (LER) Followup During this inspection the following Licensee Event Reports were reviewed to assure the accuracy and completeness of. the report, that regulatory require-ments had been met, and that appropriate corrective actions were being taken.
The LER,'s listed below were-reviewed with comments as appropriate.
250-80-28 Heater Drain Pump Breaker Malfunction 250-80-26
"4B" 125 VDC Battery Out.of Service for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 250-80-24 Feedwater Flow Control Valve Failures 1n reviewing the licensee corrective actions taken for the 3B Heater Drain Pump breaker malfunction report',
the inspector noted that the rerouted circuit wiring and conduit from the electrical box near the Heater Drain Pumps provides only a short-term fix.
Per discussions with the Electrical Department Supervisor the circuit wiring appeared to have shorted out at an electrical junction block inside the. electrical box.
Further investigation revealed another short. circuit in the wiring somewhere between the electrical box and the 4160 Volt breaker 3AB10.
This circuit wiring has been disconnected and no longer used.
The apparent cause of the shorting out of the circuit, possibly at both places, was due to steam entering the electrical box generated from a leaking seal on the 3B Heater Drain Pump.
The electrical supervisor has stated that a work order will be written to seal the. electrical boxes affecting both Units 3 and 4; The LER 250-80-28 will remain open pending implementation of corrective actions and an follow-up inspection.
The review of-LER number 250-80-24, pertaining to the Unit 3 Feedwater Flow Control valve failures and the rupture of the two inch non-safety related bypass line was initiated during this inspection reporting period.
The inspector expects to complete the evaluation and report in a
subsequent inspection repor P
7.
Surveillance Test Observations Portions of the containment Integrated Leak Rate Testing (ILRT) and low power core physics testing on Unit 4 were observed by the inspector.
The inspector ascertained that the licensee activities were conducted in accordance with the" license requirements.
No violations or deviations were identified within the areas inspected.
On January 4,
1981'he inspector witnessed in its entirety the annual engineered safeguards initiation test performed on Unit 4 per O.P.
4104.2,
"Engineered Safeguards and Emergency Power Systems-Integrated.
Test".
Attention was given to the following aspects of of the surveillance test:
a.
Review: of the surveillance procedure for conformance to technical specification requirements and proper licensee review.
b.
Verification,that test instrumentation was calibrated.
c.
Observation of the removal of the system from service.
d.
Observation of the conduct of the surveillance test.
e.
Observation of the restoration of the system to service.
f.
Review= of the test data for accuracy and completeness'
.
Independent calculation of selected test results data. to verify its accuracy.
g..'onfirmation that surveillance test documentation was reviewed.and test discrepancies are rectified.
h.
Verification that test results technical specification requirements.
.i.
Verification that. testing was. done by qualified personnel.
j.
Verification that surveillance schedule for this test was met.
No violations or deviations were identified for the areas inspected above.
8.
Inspection Followup on IE.Bulletins The inspector reviewed licensee actions in response to IE Bulletin 80-24, Prevention of Damage Due to water Leakage inside Containment and IE Bulletin 79-21, Temperature Effects on Level Instruments.
Mith regard to IE Bulletin 80-24, the inspector complied and submitted information per TI 2515 on the leakage history of all systems and components inside the containment.
Neither Unit 3 or
have open systems inside containment, hence, tabul'ated information is related to only the closed systems.
The inspection on the IE Bulletin was performed in accordance with IE Procedure The inspector reviewed the licensee responses to IE Bul 1etin 79-21 dated September 18, 1979.
During discussions with the licensee Technical Department personnel the-inspector has determined that there is no margin of safety in the established low-low steam generator level trip setpoint for safety injections and auxiliary feedwater initiation.
This is the NRR required event of a 3% minimum level reference to be used for all NTOL applications prior to adding allowance for all other errors such as transmitter error, channel draft, and reference leg heatup errors.
The plant safety analyses require that the trip be actuated when the actual water level is at or above 0% of span.
Both Turkey Point Units-3 and 4 have operating level trip setpoints at 15% of span.
9.
Review of Licensee Emergency Procedures for Provisions to Adequately Dealing-with an ATWS Events The following licensee procedures were reviewed against the inspection and acceptance criteria to determine.if procedures address and direct operator action in case of anticipated. transient without a scram (ATWS).
Emergency. Procedure 20000;. Immediate actions and;-Diagnostics..**
Emergency Procedure 20006, Loss of Normal feed water flow or steam generator level.
Operating Procedure 0103.2, Duties and Responsibilities of Operators on shift.and maintenance of operating, logs. and. Records;..
Off-Normal Operating; Procedure.-1008.',3, Loss'of R'eactor Coolant Flow.
Off-Normal Operating Procedure 1608.1, Full Length RCC,-Nalfunction.
Off-Normal Operating Procedure 0208, 1, Shutdown Resulting from reactor trip or turbine trip.
Off-Normal Operating Procedure-2608. 1,,
Chemical and Volume Control System Emergency Boration.
("Note:
EP 20000 covers immediate action for spurious safety injection, Loss of Reactor Coolant, Loss of Secondary Coolant, and Steam Generator Tube Rupture.)
The Inspection criteria were addressed in the above procedures as follows:
a
~
Failure to scram when re uired -
EP 20000, EP 20006, 0/N OP 1008.3, and 0/N OP 0208. 1 Directs the operator to manually scram if automatic initiation should have occurred but did not.
0/N
. 1608. 1 refers the operator to 0/N OP 0208. 1 for action if the reactor has tripped.
b.
Failure to corn lete scram when initiated automaticall or
~manuall 0/N OP 2608. 1 requires emergency boration to be commenced and maintained for at least 5 minutes in the event that more than one rod control cluster (RCC) fails to fully insert on a trip.
The other
J I
~
procedures under discussion are not consistent in requiring a verifi-cation that all rods are fully ins0rted after a trip.
0/N OP 0208. 1 and EP 20006 do require the verification; EP 20000 and 0/N OP 1008.3 do not.
c.
Inabilit to Move or Drive Control Rods
-.
0/N OP 1608. 1 Addresses inability to move rods or to control rod motion. It does not address failure of rods to fully insert on a trip.
d.
e.
111 A<
the-licensee did not meet the implementation date, schedule. The new schedule-completion for these items are March 1, 1981 and July 1, 1981', respectively. Item (g) does not apply to Turkey Point'nits 3 and 4 because they do not have proportional integral deviation (PID) controllers in the primary system. The inspectors will followup on items with new implementation-dates. No further questions were: asked. Plant Operations The inspector kept informed on a daily basis of the overall plant status and any significant safety matters related to plant operations. Discussions were held with plant management and various members of the operations staff'n a regular basis. Selected portions of daily operating logs and operating data sheets were reviewed on at least a weekly basis during the report period. The inspector conducted various plant tours and made frequent visits to the control room. Observations included witnessing work activities in progress, status of operating and standby safety systems, confirming valve positions, instrument readings and recordings, annunicator alarms, housekeeping, radiation area controls, and vital area controls. Informal discussions were held with operators and other personnel on work activities in progress and status of safety-related equipment or system Ouring the refueling outage maintenance for unit 4, two separate inadvertent-Engineered Safety Features (ESF) actuations occurred on January and January 6, 1981; These two inadvertent actuations resulted from maintenance on safety-related equipment using inadequate procedures which resulted from the failur'e to follow administrative procedure AP0190. 19, Control of Maintenance on Nuclear Safety-Related Systems. The Licensee Administrative Procedure APO 190. 19 control of Maintenance on Nuclear Safety Related Systems requires in paragraph 8. 1 that: 1'. A decision by< supervision be made as to whether or not a maintenance procedure is required for the work under consideration, and if not, 2'. The plant work order (PWO) be sufficiently detailed to accomplish the work safely, and 3. Any required maintenance procedures. be listed on-the PWO. In paragraph 8.4, the: procedure requires adequate consideration of con-current or sequential maintenance, testing, or operating activities. On January 3, 1981 maintenance personnel were working under PWO 7922 on Unit 4 Feedwater Regulating Valves. The repairs had been completed and it was desired to stroke the valves over their full travel to verify proper operation-. The plant was in a cold, refueling, shutdown condition. Work on a partial'ly completed surveillance'n steam flow had been suspended for several days-with two steam flow channels left in a. tripped (high flow). condition. The existing conditions of the shutdown (Reactor trip with low TAVG) had made up the necessary inputs to the protective system to generator feedwater isolation signals which prevented stroking the valves. To clear the feedwater isolation signals, the maintenance personnel inserted a test signal into the TAVG circuitry greater than the TAVG setpoint. That action accompl,ished the desired result of clearing the feedwater isolation signals but also unblocked ESF actuation so that when the test signal was removed, the true, low TAVG signal in conjunction with the two high steam flow signals completed the required inputs to the protective system to initiate ESF actuation. PWO 7922 was inadequate in that: 1. Proper consideration of the safety consequences of concurrent main-tenance and actual operating conditions was not included, and 2. The PWO did not include written steps which authorized the insertion of test signals to override the existing protective system output. Failure to include such steps prevented meaningful review of the consequences of that action by supervisory personnel as required by AP 0190.19 'n January 6, 1981 maintenance personnel working under PWO 8927 on Unit 4 steam dump valves obtained a clearance to electrically 'deenergize the valve 'A i control circuits by removing the power supply fuses F-3, and F-4 to Safe-guards Panels 4gR43 and 45. The plant was in a cold, refueling shutdown condition. Work was still suspended on the steam flow surveillance and two channels were still in a-tripped (high flow) condition. Removing F-3 and F-4 failed to deenergi ze that portion of the valve control circuit desired; however, removing F-3 and F-4 had the effect of deenergizing the ESF block "seal-in" relay. This relay energizes and seals in the ESF block signal so that the signal is maintained after the block switch is allowed to return to its neutral position. When F-3 and F-4 were subsequently reinstalled, the operations personnel failed to observe requirements of off-normal operating procedure 4008.1, Re-energizing Safeguard Racks After Loss of Single Power Supply. This would have provided for establishing 'the ESF block signal as 'he racks were re-energized and prevented the inadvertent. ESF actuation which occurred. The ESF resulted because,. the relays in 4gR43 and 45 which energize to cause operation of ESF equipment (when the required protection signals are prevent) were energized without the presence of the block signal and in conjunction with the two. high steam flow signals and existing-, low.- TAVG. PWO 8927 was inadequate in that (1) it did not properly consider the safety consequences of concurrent maintenance and actual operating conditions, and (2).it did not call out the provisions of off-normal operating procedure 4008.1 for re-energizing the safeguards racks. Both of the inadvertant. ESF* actuations.- were involved with the partial completion of Maintenance procedure 14007. 16 Steam; Generator Level Feed-water Flow and Steam'. Flow. Instrumentation Calibration such that the two steam flow channels were left in a tripped. (high flow) condition. This action is allowed by the procedure in that the precautions of paragraphs 4.3 and 4.4 which require testing of only one channel at a time with all other channels untripped are not required by the note immediately following paragraph 4.2 when the plant is in a refueling shutdown. The licensee should consider modifying MP 14007. 16 and other simi liar procedures to require circuit conditions to be returned to normal if work is suspended prior to completion to reduce the likelyhood of future unnecessary challenges to safety systems. The licensee should also consider modifying AP0190. 19 to recognize that procedures other than maintenance procedures may be required to be followed in accomplishing maintenance under a Plant Work Order to preclude unnecessary challenges to safety systems. PWO 7922 and PWO 8927 are two examples of inadequate procedures which constitute a single apparent violation of Techincal Specification 6.8 procedures (251/81-02-01). ~ g ~