IR 05000498/1993012
| ML20035E928 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/02/1993 |
| From: | Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20035E924 | List: |
| References | |
| 50-498-93-12, 50-499-93-12, NUDOCS 9304200082 | |
| Download: ML20035E928 (9) | |
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Q APPENDIX
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
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NRC Inspection Report:
50-498/93-12 50-499/93-12
Operating License:
NPF-76 NPF-80 Licensee: Houston Lighting & Power Company P.O. Box 1700 Houston, Texas 77251
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Facility Name: South Texas Project Electric Generating Station, Units 1 and 2 Inspection At: Matagorda County, Texas Inspection Conducted:
February 13 through March 17, 1993 Inspectors:
J. I. Tapia, Senior Resident Inspector Approved:
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SI/W 4-!2[93 T. F. Stetka, Chief, Project Section D Dat6 Inspection Summar_y j
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Areas Inspected: Nonroutine, announced, special inspection of technical issues associated with undersized 120 volt vital ac fuses.
Results:
Reactor operators responded well to a loss of Residual Heat Removal
during Mode 5 operation (Section 1.1).
The licensee's investigation to define the scope of undersized fuses was
extensive and did not disclose other operability or safety concerns (Section 1.3).
Cracks have been identified in spare fuses. This item is an inspection
followup item (IFI) (Section 1.3).
There has been one other similar fuse failure for which a root cause was
never defined (Section 1.4).
The licensee's responses to notifications from the industry and from the
NRC concerning related issues has been adequate (Section 1.4).
The licensee did not adequately incorporate all design loads in the
design of the circuit between the Solid State Protection System (SSPS)
9304200082 930414 PDR ADOCK 05000498 G
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l Actuation Cabinets and their associated power supplies.
This item was identified as an apparent violation of 10 CFR Part 50, Appendix B, Criterion III (Section 1.5).
Since plant startup the licensee operated both units in violation of
Technical Specification (TS) 3.3.2 requirements for having the actuation
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relays for safety injection, containment isolation, main steam line isolation, turbine trip, main feedwater isolation, and auxiliary feedwater operable. This item w?s identified as an apparent violation i
(Section 1.5).
l Summary of Inspection Findings:
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IFI 498;499/9312-01 was opened (Section 1.3).
- Apparent Violation 498;499/9312-02 was opened (Section 1.5).
- Apparent Violation 498;499/9312-03 was opened (Section 1.5).
- Attachment:
Attachment - Persons Contacted and Exit Meeting
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i DETAILS
I UNDERSIZED 120V VITAL AC FUSES i
1.1 Unit 1 Loss of Residual Heat Removal and Reactor Trip Signal On February 13, 1993, Unit I was in Mode 5 operation with residual heat removal (RHR) Pump 1A and two reactor coolant pumps in service. A trip i
actuating device operability test (TADOT), Procedure OPSP03-SP-0006S, " Train S Reactor Trip Breaker TADOT," was being conducted to verify the operability of reactor trip Breaker S, following replacement of the undervoltage and shunt trip coils. The test was commenced at 1:55 p.m. and caused an SSPS logic Train S urgent alarm to be generated at 2:26 p.m.
The SSPS actuates the engineered safety features (ESF).
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l During the performance of this TAD 0T, plant operators noticed that, in addition to the expected opening of reactor trip bypass Breaker S both normal reactor trip breakers also opened as the result of a reactor trip signal.
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They also noticed that an SSPS logic Train R urgent alarm actuated.
If the event had occurred with the reactor in operational Modes I or 2, a reactor
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trip would have been experienced. The NRC Operations Center was notified of
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the unexpected ESF actuation signal on February 13, 1993, at 5:59 p.m.
j RHR Pump 1A tripped on low discharge flow approximately 5 seconds after the t
Train A SSPS actuation cabinet lost power even though low flow conditions were l
not actually present.
At 2:31 p.m., plant operators attempted to restart this RHR pump, however, the pump again tripped on low discharge flow.
Approximately 1 minute later, RHR Pump IC was placed in service. The reactor coolant system heated up 6 F, from 181of to 187 F, during the 6-minute time frame that RHR flow was stopped.
Station Problem Report (SPR) 930503 was
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issued to further investigate the event.
The performance of the TAD 0T caused the expected urgent alarm on SSPS logic Train S.
The unexpected urgent alarm on SSPS logic Train R was caused by a loss of power to the Train A SSPS actuation cabinet from the 120 volt vital ac Channel I Ungrounded Instrumentation and Control Power Distribution Panel (DP 1201). The combination of these signals, i.e., logic Train R and S trips, generated the reactor trip signal. The RHR 1A pump trip was also caused by the loss of the 120 volt vital ac power to the Train A SSPS actuation cabinet, which gave an erroneous RHR pump low discharge flow signal.
A 10 ampere fuse which is upstream of Circuit Breaker 6 in DP 1201 and feeds the Train A SSPS actuation cabinet was discovered to be blown.
The cause of the fuse failure was not immediately identified.
Short-term corrective actions included fuse replacement and complet;on of the Train S reactor trip breaker TADOT on February 14, 1993. On February 16, 1993, SPR 930530 was written to investigate the possibility that the 10 ampere fuses, which were present in all three power distribution panels for the SSPS Trains A, B, and C
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actuation cabinets, could be undersized such that they would not be able to carry the calculated inrush current for selected accident scenarios.
Initial i
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l I-4-calculations made by the system engineer indicated that the expected inrush current was beyond the capability of the installed 10 ampere fuses.
An operability determination was conducted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the licensee
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concluded that the fuses were, in fact, undersized.
This determination
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referenced Westinghouse furnished Outline and Installation l
Drawings 387(1)00336-AWN and 387(2)00339-AWN, which indicated that the SSPS actuation cabinets have internal 20 ampere fuses installed. The sizing of these fuses were based upon a 6.8 ampere full load current, i.e., after the i
main steam line break has occurred and all required relays are energized, and
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a 46.5 ampere momentary inrush current.
The fuse curve supplied by the fuse vendor, Gould Shawmut, for their Catalog No. A60X10, Type 1, Amp-Trap 10 ampere fuse, shows that the fuse will blow in 10 milliseconds at 30 amperes.
The undersized fuses in Unit I were replaced with 20 ampere fuses in Train B on February 19, in Train A on February 22, and in Train C on February 23, 1993.
1.2 Unit 2 TS 3.0.3 Entry and Subseauent Plant Cooldown
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On February 17, 1993, Unit 2 was in Mode 4 operation.
At 10:02 a.m., Unit 2 entered TS Limiting Condition for Operation (LCO) 3.0.3 because of the operability determination in Unit I that undersized fuses existed in the three power distribution panels. The unit also entered TS LCO 3.4.9.3 because the
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overpressure protection system was determined to be inoperable since the power-operated relief valve control circuitry is associated with the SSPS
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actuation circuitry. At 10:30 a.m., unit cooldown was commenced from 343oF to
allow for reactor coolant system depressurization.
The undersized fuses were
replaced with 20 ampere fuses in Unit 2 by 3:17 p.m. the same day. Unit 2 i
l exited TS LCO 3.0.3 at 3:45 p.m. after declaring the three SSPS actuation l
trains operable. The reactor coolant system cooldown was secured about i
15 minutes later, with the temperature at 242 F.
Unit 2 remained in Mode 4 operation and did not change modes during the event. The licensee issued SPR 930545 to document the required shutdown and to provide further investigation of the event.
j 1.3 Subseauent Investigation A review was conducted by the licensee's Design Engineering Department of the Bechtel design drawings and the vendor interface drawings for the safety-i related 125 volt DC and 120 volt ac distribution panels.
A total of 102 circuits were included in the review. This review was performed to determine if conditions existed in other circuits where the fuse rating in the
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distribution panel was less than the protection provided in the vendor supplied cabinet. There were six cases identified where the upstream fuse size was lower than that provided by the vendor.
In all cases, an engineering evaluation was performed which determined that the installed fuses were acceptable. The inspector reviewed the results of the licensee's evaluations and did not identify any operability or safety concerns.
For each of the six cases, inrush current was not a contributor to the fuse load and the distribution panel circuit fuse sizing was adequate.
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i The cause for the fuse failure that occurred on February 13, 1993, has not t
been determined by the licensee.
The fuse was sent to Southwest Research l
Institute for analysis cf the failure mode of the failed fuse.
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various laboratory tests will be performed on other 10 ampere fuses removed
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from the SSPS actuation cabinets.
The results of that analysis were not
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available at the end of the inspection.
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During the replacement of the 10 ampere fuses, several fuses in a box of replacements were found with axial cracks in the end cap ferrules.
Further inspection disclosed that a majority of the spare fuses on hand in the control
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room and in the warehouse had cracked end cap ferrules.
The cracks were found
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in 10, 15, and 30 ampere Gould Shawmut Amp-Trap fuses. The cracks appear to have.been generated during the crimping of the end cap to the fuse body. The cracks were not developed to the point where the silica inside the fuse was leaking nor to the point where the mechanical strength was in question.
The licensee issued SPR 930538 to further investigate the significance of the
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cracks on fuse performance.
Further action on this matter will be followed by
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the inspectors and is considered to be an IFI (498;499/9312-01).
1.4 Previous Experience
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On January 7, 1991, a similar fuse failure occurred during the performance of i
Procedure 2 PSP 03-SP-0006S, " Train S Reactor Trip Breaker TADOT," in Unit 2
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while in power operation Mode 1.
Procedure 2 PSP 03-SP-0006S was similar to the
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current procedure, OPSP03-SP-0006S, except that the later procedure was now l
l applicable to both units.
This failure did not cause a plant trip and resulted in the urgent alarm annunciator remaining in an alarm condition and the turbine trip test lights failing to go to the configuration required by
the procedure.
The SSPS Logic Train S was declared inoperable and the shutdown action statements of TS LCOs 3.3.1 and 3.3.2 were entered.
As a result of the required shutdown, a Notification of Unusual Event was declared and unit power was ramped down until a failed fuse was found in the power
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l supply to the SSPS safeguards test cabinet. The fuse was replaced and the surveillance was successfully completed. The power reduction and Notification of Unusual Event were terminated after SSPS Logic Train S was declared operable.
l Subsequent investigation of this event indicated that the blown 10 ampere fuse was upstream of Circuit Breaker 8 in Panel DP 1203 and fed the SSPS safeguard test Cabinet B.
It was determined that the fuse had blown as a result of overcurrent. This conclusion was supported by the fact that a slow-blow 5 ampere fuse in the circuit had not blown, indicating that the fault was due to a relatively high current applied suddenly.
The cause of the overcurrent condition which resulted in the blown 10 ampere fuse could not be determined, however, it was felt that it occurred when the test switch was reset.
Several f actors were investigated to determine the most probable cause of the failure.
These included calculating the inrush current to the test cabinet when the system was placed in test, the steady state currents, and the inrush current when the system was taken cut of test.
Evaluation of the results did not indicate any reason why the fuse would have blown. Subsequent surveillance
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This event was not reported in the nuclear plant reliability data system. No
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other fuse failures have been reported by the licensee.
The inspector requested information concerning the licensee's handling of previous notifications from both the industry and the NRC which may have provided information about fuse failures. A review of the nuclear plant l
reliability data system did not provide any information which would help to l
define a potential cause of the fuse failure.
The licensee's responses to the following NRC notifications were reviewed for adequacy:
Generic Letter 88-15:
Electric Power Systems - Inadequate Control Over
Design Processes. A review of this Generic Letter was completed by Bechtel Engineering in 1988. All concerns were considered to be covered
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by the plant processes in place at the time, i
Problems in Protective Relay and Circuit l
l Breaker Coordination. The Plant Engineering Department determined that a change to the plant modifications procedure was required to make sure
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l coordination between the protective relays and circuit breakers were
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checked. The Design Engineering Department determined that available
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calculations adequately verified the breaker coordination scheme for all Class IE load centers and motor control centers as well as all Non-Class IE load centers which fed safe shutdown loads.
No actions were necessary to address the design of the circuit.
Inadequate Physical Separation and Electrical
Isolation of Nonsafety-Related Circuits from Reactor Protection System Circuits. The specific issues identified in this Notice concerned the reactor coolant pump undervoltage and underfrequency circuitry.
The concerns noted in the Notice were not applicable to South Texas Project.
Information Notice 91-29: Deficiencies Identified During Electrical
Distribution Functional Inspections.
The deficiencies identified in this Notice involved undervoltage relay setpoints for degraded grid conditions, inadequate procedures to test circuit breakers, and inadequate determination and evaluation of setpoints. No actions were necessary with respect to fuse sizing design.
Inadequate Fuse Control Programs. The
licensee found it necessary to develop a Fuse and Relay List and to revise the procedure for the control of configuration changes to include controls on fuse changeout and type verification.
Because of a lack of fuse failures, the licensee's response focused on the controls for fuse changeout. The Notice did not specifically identify a need to verify that the design of installed fuses adequately considered inrush currents
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and did not address fuse sizing.
The licensee's response to the
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notification of discrepancies between design documents and installed fuses was adequate and the fuse control program is still being developed.
The Notice also did not address fuses that were contained within vendor supplied equipment.
Potential Breaker Miscoordination Caused by
Instantaneous Trip Circuitry, The instantaneous trip function for i
j breakers is not used by the licensee for protection. No concerns l
regarding fuses were applicable.
t Information Notice 92-51: Misapplication and Inadequate Testing of
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Molded-Case Circuit Breakers. No concerns regarding fuses were applicable.
The licensee's response to these previous notifications from the NRC appears to have been adequate. The notifications did not address the potential for a design inadequacy involving installed fuse sizes due to a failure to consider the effect of inrush currents.
1.5 Analysis of Fuse Design Capability l
An evaluation of the effect of the less than adequate fuse design was performed by the licensee. Of all the accident scenarios considered, the
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l worst case scenario was identified as a main steam line break which would coincidently initiate main steam line isolation and safety injection. This condition initiates energization of 47 relays (43 latching and 4 nonlatching)
in the Trains A or B actuation panels and 37 relays (33 latching and
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4 nonlatching) in the Train C actuation panel. The manufacturer of the relays, Potter-Brumfield, provided the inrush current for each relay based upon nominal relay coil resistances. The inrush current for the latching relays was given as 0.520 amperes and for the nonlatching relays as 0.906 amperes. This calculates to 26.0 amperes each for Trains A and B.
Utilizing the minimum coil resistances (nomimal value minus 15 percent) gives j
inrush currents of 0.60 amperes and 0.94 amperes, respectively. This i
calculates to 29.5 amperes inrush current each for Trains A and B.
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Train C, inrush current was calculated to be 23.5 amperes using these minimum coil resistance values. The normal running current for each of these panels, l
i.e., without a main steam line break event, is about 1.5 amperes. When the i
inrush currents, that would occur during initiation of the main steam line break event, are added to this normal running current, the total current required by the Trains A and B panels would be 31 amperes each and 25 amperes
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l for the Train C panel. According to the manufacturer's specifications, the l
relays will pick up in 10 to 12 milliseconds. The fuse manufacturer's melting time versus current curve indicates that the fuse will open in 10 milliseconds at 30 amperes. A review of the original design calculation, Calculation No. EC-5008, Revisions 1-8, " Class lE Battery, Battery Charger and Inverter Sizing," disclosed that these relay inrush currents had not been considered in the design process by the AE (Bechtel) even though Westinghouse documents
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specifying inrush currents were available. The difference between the inrush current calculations made by Westinghouse (46.5 amperes) and those calculated by the licensee (31 amperes) were apparently due to the fact that the licensee's calculations were based upon actual plant conditions without the conservatism that was added in the Westinghouse calculations.
The licensee
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I subsequently issued Document Change Notice No. EC-0071 on March 9, 1993, to incorporate the evaluation of inrush currents for the SSPS actuation panels.
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The licensee does not have an integrated test that includes the main steam line break event.
While all applicable relays are tested, they are not tested as an integral unit in the manner that would occur during a main steam line break event.
The failure to include all design loads for the design of the
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10 ampere fuses is considered to be an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, " Design Control" (498;499/9312-02).
The SSPS Actuation Trains A and B had the potential to become inoperable had a main steam line break event occurred. A main steam line break signal would have energized the 47 actuation relays in Trains A and B which had the
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potential to cause the failure of the installed 10 ampere fuse. While redundant Train C would have been unaffected, there are certain functions included in Trains A and B that would not have been available with Train C.
These functions were the main feedwater pump trip, main steam line isolation, i
main turbine trip, and main feedwater isolation.
It appears that this condition has existed on both units since initial plant operation that occurred in March 1988 for Unit I and March 1989 for Unit 2.
TS LCO 3.3.2 requires that the ESF actuation system instrumentation channels
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and interlocks shown in Table 3.3-3 be operable. Table 3.3-3 specifies that the actuation relays for safety injection and containment isolation be operable in Modes 1-4.
Table 3.3-3 also requires that the actuation relays for main steam line isolation, turbine trip, main feedwater isolation, and auxiliary feedwater be operable in Modes 1-3.
The failure to have operable Train A and B actuation relays for all design conditions since initial plant l
l operation is considered to be an apparent violation (498;499/9312-03).
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ATTACHMENT I
1 PERSONS CONTACTED
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1.1 Licensee Personnel l
C. Bowman, Administrator, Corrective Action Group J. Calloway, Staff Consultant, Participant Services M. Coughlin, Senior Licensing Engineer, Licensing M. Chakravorty, Executive Director, Nuclear Safety Review Board
- R. Dally-Piggott, Engineering Specialist, Licensing D. Hall, Group Vice-President, Nuclear
- S. Head, Deputy Licensing Manager, Licensing H. Hesidence, Acting Director, Independent Safety Engineering Group T. Jordan, Nuclear Engineering General Manager W. Jump, Nuclear Licensing General Manager W. Kinsey, Vice President, Nuclear Generation M. McBurnett, Manager, Integrated Planning and Scheduling T. Meinicke, Deputy Plant Manager
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M. Pacy, Manager, Design Engineering Division G. Parkey, Plant Manager S. Parthasarathy, Supervising Engineer, Design Engineering Division R. Rehkugler, Quality Assurance Director T. Riccio, System Engineer, Plant Engineering Department S. Rosen, Vice-President, Nuclear Engineering R. Schiavoni, Division Manager, Design Engineering Division E. Stansel, Division Manager, Design Engineering Division P. Travis, Supervisor, Instrumentation & Controls Engineering T. Underwood, Maintenance Manager D. Wohleber, Director, Records Management / Administration 1.2 Other Personnel B. McLaughlin, Owner's Representative, Central Power & Light All personnel listed above attended an exit meeting conducted on March 9,
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1993.
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- Designates at;endance also at the March 18, 1993, exit meeting.
In addition to the personnel listed above, the inspectors contacted other personnel i
during this inspection period.
1.2 EXIT MEETING
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l An exit meeting was conducted on March 9, 1993.
During this meeting, the I
inspector reviewed the scope and initial findings of the inspection.
On i
March 18, 1993, the inspector conducted another exit meeting to present additional findings from the inspection as discussed in this report.
The licensee did not identify as properietary any information provided to, or reviewed by, the inspector.