IR 05000101/2002002
| ML17346A907 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point, 05000101  |
| Issue date: | 02/26/1985 |
| From: | Brewer D, Elrod S, Elvod S, Peebles T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17346A905 | List: |
| References | |
| 50-250-85-02, 50-250-85-2, 50-251-85-02, 50-251-85-2, NUDOCS 8503200179 | |
| Download: ML17346A907 (16) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTAST., N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-250/85-02 and 50-251/85-02 Licensee:
Florida Power and Light Company 9250 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:
January 1 - February 2,
1985 Inspectors:
. A. Peebles, Seni Resident Insp ctor D
R. Brewer, Resid t Inspector at Signed Z 2Z ate igned Approved by:
Stephe A. Elrod, Section Chief Division of Reactor Projects Date igned SUMMARY Scope:
This routine, unannounced inspection entailed 284 direct inspection hours at the site, including 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> of backshift, in the areas of licensee action on previous inspection findings, annual and monthly surveillance, annual and monthly maintenance, operational safety, engineered safety features walkdown, plant events, independent inspection, and spent fuel pool re-rack.
Results:
Of the eight areas inspected, no violations or deviations were identified in six areas; three violations were identified in two areas (two separate examples for failure to meet Technical Specification surveillance requirements, paragraph 5;
and failure to obtain quality control approval prior to beginning maintenance, paragraph 6).
8503200 05000250 9 850228 PDR ADOCK PDR
REPORT DETAILS Licensee Employees Contacted K.
N. Harris, Vice President-Turkey Point C. J.
Baker, Plant Manager-Nuclear D.
D. Grandage, Operations Superintendent-Nuclear T.
A. Finn, Operations Supervisor
"J.
A. Labarraque, Acting Plant Manager Nuclear and Technical Department Supervisor
"B. A. Abrishami, IST Supervisor 0 ~ Tomaszewski, Plant Engineer Supervisor D.
A. Chancy, Corporate Licensing
~J. Arias, Regulation and Compliance Engineer
"J, W.
Kappes, Maintenance Superintendent-Nuclear W.
R. Williams, Asst.
Superintendent, Electrical Maintenance R.
A. Longtemps, Asst. Superintendent, Mechanical Maintenance E.
F.
Hayes, Asst. Superintendent, I and C Maintenance V. A. Kaminskas, Reactor Engineering Supervisor R.
G.
Mende, Reactor Engineer R.
E. Garrett, Plant Security Supervisor
~P.
W. Hughes, Health Physics Supervisor 8 Acting Operations Superintendent R.
M. Brown, Asst. Health Physics Supervisor W.
C. Miller, Training Supervisor P. J.
Baum, Asst. Training Supervisor J.
M. Donis, Site Engineering Supervisor J.
M. Mobray, Site'echanical Engineer L.
C. Huenniger, Start'-up Superintendent T. Young, Project Site Manager M. J. Crisler, equality Control Supervisor
~R.
H. Reinhardt, equality Control Inspector
~K.
L. Jones, equality Assurance Supervisor J.
E.
Moaba, PEP Program Manager
~J.
E. Price, Safety Engineering Group
~G.
M. Vaux, Safety Engineering Group
"T.
C. Grozan, Licensing Engineer
~P.
Pace, Licensing Engineer Other NRC personnel
~D.
G. McDonald, NRR Project Manager
~Attended exit interview Exit Interview The inspection scope and findings were summarized during management interviews held throughout the reporting period with the Plant Manager-Nuclear and selected members of'is staff.
The exit meeting was held on January 31, 1985, with the persons noted above.
The areas requiring management attention were reviewed.
The items which were identified as potential violations were:
failure to meet the requirements of Technical Specification (TS) 4.8. 1.C during emergency diesel generator (EDG) load rejection testing (250,251/85-02-01):
failure to meet the requirements of TS 4. 10.4 during auxiliary feedwater (AFM) pump test (250,251/85-02-02):
failure to obtain quality control (gC) approval prior to beginning maintenance on safety-related equipment (250,251/85-02-03).
The licensee acknowledged the findings; however, it was noted that the licensee had not agreed to the EDG load rejection TS interpretation.
Another exit was held with the Plant Manager on February ll, 1985, to confirm that management was aware of NRC concerns in the following areas:
a safety relief valve program will be set-up and all safety relief valves on Unit 3 will be appropriately tested during the upcoming refueling (IFI 250,251/85-02-05);
an engineering task will be expedited to evaluate the monitoring of loss of control voltage at the emergency diesel generators (IFI 250,251/85-02-04);
and torque switch and limit switch settings for the motor operated valves (MOY) will be researched and documented (IFI 250,251/85-02-06).
The licensee agreed that these items will be updated by the end of February 1985.
In neither of the above exit meetings did the licensee identify as proprietary any of the materials provided or reviewed by the inspectors during this inspection.
Licensee Action on Previously Identified Items a.
Monthly update of Performance Enhancement Program (PEP)
The PEP was reviewed to determine if commitments were being met.
Status was discussed with the PEP Manager and with other members of management.
The new PEP Manager is J. E.
Moaba.
The facility upgrade project has completed driving pilings for the new administrative building.
The third floor framework has received approval.
The steel skeleton of the new Health Physics building has been erected and some of the roof has been ins'tailed.
The permit to begin clearing for the training/simulator building has been received.
The meeting of scheduled commitments for the PEP continues to be within acceptable limits and all modifications have been cleared by the Region.
b.
(Open) Violation 250/84-28-02 8 251/84-29-03 In January 1985, at the request of the licensee, engineers from the Mestinghouse Electric Corporation, Nuclear Fuels Division, visited the site to evaluate estimated critical condition (ECC) calculations.
The review revealed that improvements could be made to improve the level of
accuracy of the ECC calculations.
It was determined that no additional physics testing was required to verify the data used in the ECC.
The cycle specific Nuclear Design Report (NDR) provided by Westinghouse has been the primary source of data used in the ECC.
However, the NDR was not intended to provide data of the accuracy needed to prevent previously documented ECC discrepancies from occurring.
To alleviate the problem the Westinghouse engineers recommended the following actions:
(1)
Utilization of integral rod worth curves for beginning, middle and end of core life, referenced to hot zero power (2)
Utilization of integral boron worth curves for beginning, middle and end of core life (3)
Utilization of burnup specific Xenon and Samarium curves (4)
Application of a
reasonability check to the last measured pre-shutdown boron concentration to ensure the ECC is not based on an anomalous point.
The licensee has implemented the use of core life dependent integral rod worth curves for both units.
Burnup specific Samarium curves have been incorporated into the plant curve book.
Use of the recommended integral boron worth curves will begin as soon as a revised ECC procedure can be developed to accommodate the necessary calculations.
The revised procedure will incorporate the use of the recommended reasonability check.
Burnup specific Xenon curves will be implemented.
Some necessary computer programming changes have been completed on the digital data processing (DDPS) system.
The licensee expects to have completed all recommendations by the end of the Unit 3 refueling outage which,is scheduled to begin on March 31, 1985.
4.
Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations.
No unresolved items were identified during this inspection period.
5.
Monthly and Annual Survei1 lance Observation (61726/61700)
The 'inspectors observed TS required surveillance testing and verified that testing was'erformed 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 that any deficiencies
identified during the testing were properly reviewed and resolved by management personnel; and that system restoration was adequate.
For completed tests, the inspector verified that testing frequencies were met and tests were performed by qualified individuals.
The inspectors witnessed/reviewed portions of the following test activities:
Emergency Diesel Generator - Periodic Test Load on 4KV Bus Emergency Diesel Generator Partial Load and Full Load Rejection Test Emergency Diesel Generator - Eight Hour Full Load Test and Load Rejection Test Power Range Nuclear Instrument Periodic Channel Functional Test Auxiliary Feedwater System - Periodic Test Operating Procedure (OP) 4304.3,
"Emergency Diesel Generator - Eight Hour Full Load Test and Load Rejection",
was performed on December 22, 1984, for the
"A" Emergency Diesel Generator (EDG), and on January 14, 1985, for the
"B" EDG.
During each test the voltage regulator for the applicable EDG failed to limit the full load rejection voltage spike to less than 4784 volts as required by TS 4.8. 1.C.
Following the December test, the licensee developed Temporary Operating Procedure (TOP)
138, "'A'mergency Diesel Generator Partial Load and Full Load Rejection", for the purpose of checking and adjusting the EDG voltage regulator and collecting additional load rejection data.
Subsequent testing of the "A" EDG revealed that the voltage regulator was not capable of suppressing a voltage spike of approximately 5000 volts during the seven tenths (.7) of a second immediately following the full load rejection.
A stable voltage of 4320 volts was reached approximately two seconds after the load rejection.
An engineering evaluation concluded that:
the voltage transient experienced during the full load rejection was acceptable; that all equipment operated as designed; that no equipment degradation had occurred; and that better monitoring equipment used during this test showed the transient which had not been noticed previously when only control room meters had been used.
Also, it was determined that USNRC Regulatory Guide 1.9, "Selection, Design, and gualification of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants",
and Regulatory Guide 1. 108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants",
were issued after the installation of the Turkey Point EDGs.
Consequently, the EDGs were not designed to meet the criteria set forth in these two documents.
Similarly, TS 4.8. 1. C was also issued after the EDGs were designed and installed.
The observed voltage transient was determined to be inherent in the system design.
The Plant Nuclear Safety Committee (PNSC)
reviewed the above engineering evaluation and the TS and concurred with an Engineering Department recommendation that, based on the design of the voltage regulator, compliance with the TS limit during full load rejection should be verified by comparing steady state values before and after the rejection.
The inspectors observed the testing of the
"B" EDG in accordance with OP 4304.3 on January 14, 1985.
The voltage trans'ient was observed to exceed the 4784 volt limit established by TS 4.8. 1.C.
The inspectors began an independent assessment of the intent of the TS.
Review of available information, including Regulatory Guides 1.9 and 1. 108, Westinghouse Standard TS (STS)
4.8. 1. l. 2. d. 3, STS Bases 3/4.8. 1 through 3/4.8. 3, and conversations with USNRC Staff in the Region and in the Office of Nuclear Reactor Regulation, resulted in the conclusion that the 4784 volt limit of TS 4.8. 1.C applied to momentary transients or voltage spikes as well as to steady state values.
The residents concluded that TS 4.8. 1.C requires that each EDG be demonstrated operable at least once each 18 months by verifying the EDG's capability to reject complete load without exceeding 4784 volts.
Contrary to the above, on December 22, 1984, during a test of the "A" EDG, and on January 14, 1985, during a test of the "B" EDG, the requirements of TS 4.8. 1.C were not met, in that the EDG's exceeded 4784 volts during full load rejection testing.
The PNSC reviewed an engineering evaluation of the test data, concluded that the EDG voltage regulators had performed as designed and concluded that the TS requirements had not been exceeded.
'Failure to meet the requirements of TS 4.8. 1.C is a violation (250,251/85-02-01).
The licensee submitted a voluntary report on the above, LER 250/84-40, documenting the discrepancies encountered during the performance of OP 4304.3.
The licensee has committed to revising TS 4.8.1 by April 1, 1985, in response to Generic Letter 84-15.
The licensee intends to evaluate the EDG testing and acceptance criteria and any necessary changes will be included as part of the proposed amendment.
In January 1985, the inspectors reviewed OP 7304. 1, "Auxiliary Feedwater System - Periodic Test",
a procedure used to perform surveillance requirements in accordance with TS 4. 10.
The procedure did not address testing the installed, safety-related seismically qualified, bottled nitrogen system which is used to backup the non-safety-related instrument air system in operating the AFW flow control valves.
The AFW flow control valves fail closed on loss of control air pressure.
Since the instrument air system is neither designed, nor operated as a qualified system, the bottled nitrogen system must function for the AFW system to be operable during all postulated accidents.
A separate nitrogen station is provided for both Unit 3 and Unit 4.
Each station consists of five nitrogen bottles, three of which are dedicated to supply train 1 and 2 to supply train 2.
One nitrogen bottle is aligned to each train through a pressure regulator.
Low pressure alarms on each train indicate the need to valve in an additional nitrogen bottle and replace depleted bottles.
TS 4.10 requires that periodic testing of the AFW system be performed to verify the operability of the system and its ability to respond properly when required.
TS 4. 10. 4 requires that tests shall be considered-satisfactory if control panel indication and visual observation of the equipment demonstrate that all components have operated properly.
Contrary to the above, prior to January 31, 1985, periodic testing of the AFW system was considered satisfactory without demonstrating that all system components operated properly, in that the installed, safety-related AFW nitrogen system was not shown to be capable of controlling 'the AFW flow control valves as designed.
Failure to meet the requirements of TS 4. 10.4 is a violation (250,251/85-02-02).
6.
Monthly and Refueling Maintenance Observations (62703 8 62700)
Station maintenance activities of safety-related systems and components were observed/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:
LCO were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; fire prevention controls were implemented; and housekeeping was actively pursued.
The following maintenance activities were observed and/or reviewed:
Repair of Unit 4B EDG output breaker which had failed to close and was found to be bent due to a cell switch being misaligned Repair of the B
EDG governor solenoid Packing replacement on the A AFW pump shaft Replacement of Unit 3 main steam hanger number
Inspection of B EDG - mechanical (MP 4307.3)
and electrical Repair of B EDG cooling water fill/vent line Replacement of an A AFW pump steam inlet pipe elbow Bench testing of relief valve, RV-4-857 Repair of C AFW pump trip and throttle valve On January 24, 1985, during an operability check of the B EDG, the diesel engine failed to automatically shutdown following the normal 11 minute cooldown period.
Operator action using the local and remote shutdown switches was ineffective.
The turbine operator operating the EDG stopped the diesel engine by manually tripping the fuel racks.
Troubleshooting activities associated with the repair of the EDG were observed.
Maintenance was begun using plant work order (PWO) 5090 and the problem was determined to be the result of a sticking governor solenoid valve.
While stuck, this valve prevented dumping governor control oil by either the local or the remote shutdown switches.
The sticking governor solenoid was turned over to Instrumentation and Control (IC) personnel by Electrical personnel, who cycled the solenoid many times without it sticking again.
The EDG was turned over to Operations personnel and the test was successfully re-run.
Maintenance personnel contacted the vendor and a
representative was sent.
Subsequent dismantling of the solenoid revealed a
carbon-like buildup on the solenoid which had been restricting the solenoid movement.
While initial electrical repair efforts were in progress, PWO 5090 was reviewed.
It was noted that a quality control review had not been conducted prior to commencing the maintenance work.
AP 0190. 19; "Control of Mainte-nance on Nuclear Safety Related and Fire Protection Systems",
section 8. 1.5, requires that this review be conducted.
Section 8. 1.8 requires that a
equality Control Inspector certify the completion of the review by initialing the PWO.
Since the review was not conducted this signature was missing.
TS 6.8. 1 requires that written procedures and administrative policies be established, implemented and maintained that meet or exceed the requirements and recommendations of section 5. 1 and 5.3 of ANSI 18.7-1972 and Appendix
"A" of USNRC Regulatory Guide 1.33.
Contrary to the above, on January 24, 1985, AP 0190. 19 was not implemented, in that maintenance was begun on the B
EDG using PWO 5090, which had not received a equality Control review and was.not approved by a equality Control Inspector.
Failure to meet the requirements of TS 6. 8. 1 is a violation (250,251/85-02-03).
On January 25, 1985, during surveillance testing of the C
AFW pump, the turbine received an electrical overspeed signal and the Trip and Throttle (T and T) valve closed.
This valve is electrically released and spring driven closed and is opened by the motor operator.
When the control room operator attempted to open the T and T valve the motor operator would not perform.
Electrical Department personnel found that the torque switch setting was at minimum and reset it so that the valve would open.
Discussions with Electrical Department personnel led to a licensee commitment that a program to research, document and then set the torque switch and limit switch settings for all motor operated valves would be actively pursued.
This is an inspector follow-up item (IFI 250,251/85-02-06).
On January 29, 1985 at 3:50 a.m.,
a control room operator found the EDG lights in the control room were off but no annunciator alarm was received.
There are two lights for each EDG on each unit's control board, NORMAL and READY TO START.
The turbine operator found that a light bulb on the EDG local control panel had exploded.
Subsequent investigation found that the light holder had shorted to the control panel frame and had blown two fuses.
These fuses supply a portion of the control power and the EDG would not have
started when called upon.
This portion of control power is not monitored by the loss of control power circuit and thus no alarm was received.
The licensee has agreed to expedite an engineering task to evaluate the monitoring of loss of control voltage at the EDG.
This will be followed as (IFI 250,251/85-02" 04).
During an inspection by a regional inspector on January 21-25, 1985, it was discovered that<<the only program of safety relief valve testing was what the Inservice Test Program required and that included only the Pressurizer and-Steam Generator safety valves.
Licensee management agreed to implement a
program to test safety-related relief valves on a five year period and to test all such valves on Unit 3 during the upcoming refueling.
This will be followed as (IFI 250,251/85-02-05).
Also, they agreed to test essential safety valves during the current outage of Unit 4.
The valves which were tested prior to the unit returning to service on January 30, 1985, were as follows:
accumulator primary letdown, boron injection, seal water return, and residual heat removal.
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, reviewed tagout records, verified compliance with TS LCO and verified the return to service of affected components.
The inspectors, by observation and direct interviews verified that the physical security plan was being implemented in accordance with the 'station security plan.
The inspectors verified that maintenance work orders had been submitted as
.required and that followup and prioritization of work was accomplished.
The inspectors observed plant housekeeping/cleanliness conditions and verified implementation of radiation protection control.
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 on Unit 3 and Unit 4 to verify operability and proper valve/switch alignment:
Emergency Diesel Generators Auxiliary Feedwater Intake Cooling Mater
No Steam Generator Blowdown Containment Spray Containment Penetrations Nuclear Instrumentation Drawers Refueling Mater Storage High Head Safety Injection violations or deviations were identified.
8.
Engineered Safety Features Walkdown (71710)
The inspector verified operability of the AFW system, which is common to Units 3 and 4, by performing a complete walkdown of the accessible portion of the system.
The following specifics were reviewed and/or observed as appropriate:
a.
b.
C.
d.
e.
that the licensee's system lineup procedures matched plant drawings and the as-built configuration; that the equipment conditions were satisfactory and items that might degrade performance were identified and evaluated (e.g.
hangers and supports were operable, housekeeping was adequate, etc.);
that instrumentation was properly valved in and functioning and that calibration dates were not exceeded; that valves were in proper position, breaker alignment was correct, power was available, and valves were locked/lockwired as required; local and remote position indication was compared and remote instrumentation was functional; breakers and instrumentation cabinets were inspected to verify that they were free of damage and interference.
No violations or deviations were identified.
9.
Plant Events (93702)
a.
On January 13, 1985, at 4:28 a.m.,
a Unit 3 blowdown containment isolation valve did not close on command and an Unusual Event was declared because the TS have no time allowed for action.
The valve was manually closed at 4:37 a.m.
It was found that the air control solenoid valve had positioned to vent but failed to vent the air pressure due to a blocked strainer on the vent port.
These valves are manufactured by Hiller.
The only valves at the plant of this type are the steam generator blowdown containment isolation valves and the breathing air containment isolation valves (a total of eight valves).
The licensee has put these solenoid valve vent port strainers on a
preventative maintenance program.
b.
On January 19, 1985, at 5:30 a.m.
while shutting down Unit 3, the reactor tripped at 10K power from a turbine trip caused when the unit power increased slightly and reinstated permissive P-7.
The steam generator chemistry was found out of specification which put the unit in an administrative action statement requiring a
load reduction.
Indications were of a condenser tube leak.
It was necessary to plug three leaking tubes and nine surrounding tubes.
C.
On January 29, 1985,.at 11:45 a.m.
a Unit 3 reactor trip occurred when a turbine operator tripped open the Motor Generator (MG) set breakers for the control rod power supplies.
The rods tripped, causing the nuclear instrumentation to sense a
dropped rod and resulted in a turbine runback.
All safety systems responded within parameters.
The turbine operator had been instructed to tag open the Unit 4 MG set output breakers however he entered the wrong room and opened the wrong breakers.
No violations or deviations were identified.
10.
Spent Fuel Pool Rerack (50095 8 37700)
The inspector witnessed several of the evolutions of the spent fuel pool rerack on Unit 3.
This consisted of the decontamination of the existing racks before taking them from the pool; the lifting of the racks out of the pool and final rinsing; the modification of the new racks before placing them in to the pool; the placing of the new racks into the pool; the health physics activities around the pool area; the operations of,moving the spent fuel and raising and lowering the pool water level; and the monitoring of the pool boron concentration.
Also, the crating and loading of the old racks for shipment was witnessed, including the health physics survey activities.
At the end of the report period, the old racks had been removed and half of the new racks had been placed into the pool.
No violations or deviations were identified.
ll.
Independent Inspection (92706)
During the reporting period, the inspectors routinely attended meetings with licensee management and shift turnovers between licensed operators.
These meetings and discussions provided a daily status of plant operating and testing activities as well as a discussion of significant problems or incidents.
On January 22, 1985, during a walkdown of the AFW system, a constant force spring hanger was found broken.
The hanger was adjacent to the C
AFW pump and was supporting a
The site engineering office was informed, and they placed a temporary support on the main steam line, until the support could be analyzed and repaired.
The replacement was completed by January 22.
No further questions on this support remain.
Several new fuel shipments have been arriving on a weekly basis.
The inspector has reviewed their receipt inspection.
No violations or deviations were identified.
ENCLOSURE
L-85-l 30 Dr. 3. Nelson Grace Regional Administrator, Region ll U. S. Nuclear Regulatory Commission Suite 2900 101 Marietta Street N.W.
Atlanta, Georgia 30323
Dear Dr. Grace:
Re:
Turkey Point Units 3 and 0 Docket Nos. 50-250, 50-251 Ins ection Re ort 250-85-02 and 251-85-02 Florida Power R Light Company has reviewed the subject inspection report and a response is attached.
C There is no proprietary information in the report.
Very truly yours, 3. W. %'illiams, 3r.
Group Vice President Nuclear Energy Department 3%'W/SAV/js Attachment cc:
Harold F. Reis, Esquire PNS-LI-85-127v
ATTACHMENT Re:
TJ.-'i Point Units 3 and
Docket No. 50-250, 50-251 IE Ins ection Re ort 250-85-02 and 251-85-02 FINDING 1-Technical Specification (TS)
4.8.1.C requires that each diesel generator be demonstrated operable at least once each
months by verifying the diesel generator's capability to reject complete load without exceeding 0784 volts and without exceeding overspeed limits.
Contrary to the above, on December 22, 1980, during a test of the "A" diesel generator, and on 3anuary 0, 1985, during a test of the "8" diesel generator, the requirements of TS 0.8.1.C were not met, in that the diesel generators exceeded 0780 volts during full load. rejection testing.
The Plant Nuclear Safety Committee (PNSC) reviewed an engineering evaluation of the test data, concluded that the diesel generator voltage regulato'rs had performed as designed and concluded that the TS requirements had not been exceeded.
RESPONSE:
1)
FPL does not concur with the finding that the diesel generator testing did not meet Technical Specification requirements.
2)
FPL does not agree with the finding, because it has been FPL's understanding that the Turkey Point Technical Specifications, concerning the voltage criteria during diesel generator load rejection testing, were based on steady state voltage values rather than maximum short-duration transient values.
These load rejection criteria were adopted by FPL for the Turkey Point Technical Specifications based on NRC staff guidance at their request.
This NRC staff surveillance guidance was not specific in this area i'n that it did not clearly identify whether the voltage criteria was a
steady state or maximum transient voltage criteria.
The steady state load rejection voltages measured for both diesel generators during the surveillance testing of December 22, 1980 meet the Technical Specification requirements.
In addition, an engin=ering evaluation was performed which clearly demonstrates that appropriate design and operational criteria were met.
This engineering evaluation was performed to analyze the voltage and overspeed testing results, obtained on December 22, 1980, to ensure that these did not exceed equipment design criteria or result in any degradation of equipment performance.
The results of that evaluation which are fully documented in a Voluntary LER Report (250-80-000) dated 3anuary 22, 1985, concluded that the transient voltage and overspeed performance of both diesel generators was normal, meeting all original design criteria and was consistent with the factory load rejection test results.
In addition, the engineering evaluation concluded that the voltage and overspeed performance of the diesel generators did not result in any equipment performance degradation as a consequence of the load rejection transients.
I E Ins ction Rc rt.250-85-02 a~d 251-85-02 Page
A separate engineering evaluation has been completed which reviewed the load rejection voltage, frequency, and overspeed criteria, contained in Specification 0.8 and recommended revised criteria appropriate to the Turkey Point Emergency Diesel Generators for use in the diesel generator surveillance technical specifications.
With the completion of this engineering evaluation, a revision to Technical Specification 0.8 will be initiated to incorporate the load rejection criteria developed during this engineering review.
The revision of Technical Specification 0.8 will be submitted to the NRC by May 1, 1985, as a part of those revisions to this Technical Specification in response to Generic Letter 80-15.
I FINDING 2:
TS 0.10 requires that periodic te ting of the Auxiliary Feedwater (AFW) system be performed to verify the operability of the system and its ability to respond properly when required.
TS 0.10.0 requires that tests shall be considered satisfactory if control panel indication and visual observation of the equipment demonstrate that all components have operated properly.
Contrary to the above, prior to 3anuary 31, 1985, periodic testing of the AFW system was considered satisfactory without demonstrating that all system components operated properly, in that the installed safety-related AFW nitrogen system was not shown to be capable of supplying pneumatic control pressure to the AFW flow control discharge valves as designed.
RESPONSE:
1)
FPL concurs with the violation and agrees that the AFW Backup Nitrogen System should be included in a periodic surveillance to demonstrate the operability of that AF W subsystem.
')
The reason for the finding was inadequate procedural guidance to control the operability surveillance of the AFW Backup Nitrogen System.
In the past, the AFW backup nitrogen system was not considered as part of the operability acceptance criteria for the AFW system.
3)
As an interim corrective measure, a temporary surveillance procedure has been developed to test the AFW Backup Nitrogen System in conjunction with inservice valve testing on a quarterly surveillance schedule.
As part of this temporary quarterly surveillance procedure, the proper operation of the minimum nitrogen pressu. e annunciator alarm instrumentation will be verified. This tempo. ary procedure ha-been P4SC -pprpved for p!ant use.
, ENCLOSIJRE
EVALUATION OF EAERCENGY DIESEL GENERATOR LOAD REJECTION
. TKS OP DELIBER 22 and
1984 Purpose:
This evaluation has been conducted to determine if. the transients obtained during the subJect test exceeded any equipment design limits.
Scope:
The scope of this evaluation consists of analyzing the equipment which experienced the voltage transient to determine if any equipment design limits were exceeded, analyzing the increase in speed experienced to determine if it is within acceptable limits and analyzing the voltage regulator performance to determine if it is degraded from original design.
Additionally the applicability of Regulatory Guide 1.9 and 1.108 with respect to the design of the emergency diesel generators will be determined.
Discussion:
The maximum voltage transient experienced during the full load (2500 KW) rejection tests was in excess of 5,000 volts for approximately one second.
The following analysis shows that the voltage transient is within the voltage ratings of the affected equipment.
Per the manufacturer's data, the diesel generator breaker, potential and current transformers are designed to pass an applied A.C.
potential test for one minute at
kV and to withstand an impulse of
kV.
The diesel generator cables were designed and tested in accordance with IPCEA S-lg"81
~hich states that, this size cable shall withstand without failure a
kV A.C. 'voltage for five minutes.
Per the manufacturer's instruction manual, the diesel generator stator was designed tq pass a high potential test of 4.8 kV phase to ground (8.3 kV phase to phase)
for one minute.
As can be seen above, the applied potential tests for the affected equipment envelope the voltage transient experienced and therefore no equipment degradation is expected.
The engine speed transient also experienced during the full load rejection test was determined to be inherent to the system design, and to be within acceptable limits.
Regulatory Guide 1.9 states that "the transient following the complete loss of load should not cause the speed of the uriit to attain the overspeed trip setpoint".
Based on this and the fact that the manufacturer's recommended trip setpoint was not reached we conclude that the overspeed experienced is normal and resulted in no equipment degradation.
The voltage transient was also determined to be inherent to the system design.
By comparison of the original factory load rejection test results with the subject test results, we have determined that the voltage regulator performance is not degraded.
Conclusion:
Ve conclude that the voltage and speed transients experienced during the full load rejection tests are acceptable, all equipment operated as designed and that no equipment degradation occurred as a result of these transients.
In addition we have determined that Regulatory Guide 1.9 and 1.108 were not included in the design basis of the emergency diesel generators since they were manufactured 'prior to issuance of the Regulatory Guides.