ML20059L964
| ML20059L964 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 11/16/1993 |
| From: | Siegel B Office of Nuclear Reactor Regulation |
| To: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| References | |
| TAC-M86481, TAC-M86482, NUDOCS 9311180036 | |
| Download: ML20059L964 (7) | |
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November 16, 1993 Docket Nos. 50-348 and 50-364 Mr. D. N. Morey, Vice President Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295
Dear Mr. Morey:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATING TO ISSUES RESULTING FROM THE ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL INSPECTION PERFORMED AT THE JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 (TIA 93-05) (TAC N05. M86481 AND M86482)
The staff has reviewed your response to the six issues resulting from the Electrical Distribution System functional Inspection (EDSFI) that was provided in your November 5, 1992, response to Unresolved and Inspector followup Items identified in Inspection Report Numbers 50-348/92-17 and 50-364/92-17. We have reviewed the information contained in these documents and determined that additional information is required to fully understand your response to these issues. Thus, the enclosure provides a request for additional information and identifies each issue resulting from the EDSFI that relates to each request' for additional information.
It is requested that your response be provided within 45 days of receipt of this letter. Should you have any questions regarding this request or the schedule for responding, please contact me.
ORIGINAL SIGNED BY:
Byron L. Siegel, Senior. Project Manager Project Directorate 11-1 Division of Reactor Projects - I/II
Enclosure:
Request for Additional Information DISTRIBUTION w/ enclosure
+ Docket File P. Anderson NRC & Local PDRs B. Siegel PD#11-1 Reading OGC S. Varga ACRS (10)
G. Lainas E. Herschoff, Region 11 S. Bajwa cc:
Plant Service list 0FC LA:PD2(hitPE PM:PD2-M Jr AD:PD2):DRPE NAME PAnd'ekkY BSbg[M SBajw[
11/h/93 11/ / b[/93 11/Ib /93
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/93 DATE 0FFICIAL RECORD COPY DOCUMENT NAME:
G:\\FARLEY\\rAR86481.RAI 9311180036 931116 PDR ADOCK 05000348 N CM wy eoa 1700fn g
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s' Mr. D. N. Morey Joseph M. Farley Nuclear Plant Southern Nuclear Operating Company, Inc.
cc:
Mr. R. D. Hill, Jr.
State Health Officer General Manager - Farley Nuclear Plant Alabama Department of Public Health Southern Nuclear Operating Co., Inc.
434 Monroe Street Post Office Box 470 Montgomery, Alabama 36130-1701 Ashford, Alabama 36312 Chairman Mr. B. L. Moore, Licensing Manager Houston County Commission Southern Nuclear Operating Co., Inc.
Post Office Box 6406 Post Office Box 1295 Dothan, Alabama 36302 Birmingham, Alabama 35201-1295 Regional Administrator, Region II James H. Miller, III, Esquire U. S. Nuclear Regulatory Commission Balch and Bingham Law Firm 101 Marietta St., N.W., Ste. 2900 Post Office Box 306 Atlanta, Georgia 30323 1710 Sixth Avenue North Birmingham, Alabama 35201 Resident Inspector U.S. Nuclear Regulatory Commission Mr. J. D. Woodard Post Office Box 24 - Route 2 Executive Vice President Columbia, Alabama 36319 Southern Nuclear Operating Company P.O. Box 1295 Birmingham, Alabama 35201
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N Enclosure RE00EST FOR ADDITIONAL INFORMATION R G ilNG TO ISSUES RESULTING FROM THE ELECTRICAL DISTRIBUTION SYSTEM FUNCTIONAL INSPECTION PERFORMED AT THE JOSEPH M. FARLEY NUCLEAR PLANT. UNITS 1 AND 2 (TIA 93-050 (TAC NOS. 86481/2) 1.
Unresolved Item 348, 364/92-17-05, Degraded Grid Voltage Relay Settings Specified by Technical Specifications (TS) are Inadequate.
Issue:
The licensee's calculations indicated that the degraded voltage relay setpoints currently utilized in the Technical Specifications (TS) would not ensure sufficient voltage was maintained to all safety loads. As a result before the :r Aum settings would be reached, some safety-related loads below the 6s0 '
'el would become inoperable. Administrative controls had been imp:..nted to ensure adequate voltage, with the offsite dispatch center having the primary responsibility.
The administrative lower limit resulted in a minimum expected steady state voltage for an accident scenario of 93.7 percent at the 4160 V safety-related buses.
Region 11 confirmed that this voltage translated to adequate voltage throughout the safety-related systems. The minimum acceptable degraded voltage relay setpoint specified by TS Table 3.3.4 was 87.45 percent at 4160 Vac safety-related busses.
Recuest for licensee:
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a.
The NRC staff issued generic correspondence dated August 8, 1979, addressing " Adequacy of Station Electric Distribution System i
Voltages." This correspondence required each licensee to review the electric power systems at each of their nuclear power plants and determine analytically if the offsite power system is of sufficient capacity and capability to automatically start, as well as operate, all required safety loads. This correspondence also required that the analytical results be verified.
In response to this generic correspondence, Alabama Power Company (APC) submitted voltage analyses and verification results on December 11, 1979.
This submittal was reviewed and evaluated by the Idaho National Engineering Laboratory (EG&G).
By letter dated February 18, 1981, the NRC technical staff provided a safety evaluation (SE) that concluded the voltage analyses and test results provided by APC were acceptable. Table 2, " Comparison of Analyzed Voltage and Undervoltage Relay Setpoints," in the Technical Evaluation Report, dated January 1981, and provided to support the staff SE indicates that the relay setpoints for degraded grid are 89.65 percent ( 1) at 4.16kV safety-related buses with 20-30 second time delays.
Currently, minimum acceptable degraded voltage relay setpoints specified by TS Table 3.3.4 are 87.45 percent at 4.16kV safety-related busses. This being the case, provide technical bases for changing the degraded voltage relay
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In addition, provide a list of safety and/or non-safety-related loads that have been added or deleted since the i
last degraded grid voltage study was performed.
Further, explicitly i
identify the 120 Vac safety-related loads that are not supplied by an inverter or a voltage regulation transformer.
b.
In recent years, Southern Company Services (SCS) has developed a load study to evaluate the adequacy of voltages to safety-related components using updated analysis models for the electrical system to a level of detail that includes 208 Vac and 120 Vac systems.
Describe how SCS verified the accuracy of the updated analysis models and associated analyses results.
In addition, identify what safety-related loads and non-safety-related loads have been included in these analyses and describe how the updated load study ensures that the current degraded grid soltage relay setpoints are l
technically adequate and acceptable.
2.
Unresolved Item 348, 364/92-17-07, Auxiliary Building Battery Voltage is l
Marginal for Present Load Requirements.
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Issue 1:
The TS and FSAR imply two hours of adequate battery voltage in the I
absence of other DC sources.
The licensee's calculation demonstrated l
only a one minute capability. The licensee's personnel stated that this was sufficient in that, within one minute of a loss of offsite power, l
EDGs would be in operation providing DC power through the chargers.
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Reouest for licensee:
a.
It is our understanding that the 2-hour battery capacity requirement is derived from a recommendation contained in NUREG-0611 (referenced in Standard Review Plan (SRP) Section 10.4.9) in which turbine-driven auxiliary fndwater (AFW) pumps are to be available for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> assuming a total loss of AC power and control power to these pumps is from DC sources. This being the case, each auxiliary building safety-related battery is required to have adequate storage capacity to carry vital DC loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> assuming a total loss of AC power.
Thus, explain in detail how each auxiliary building safety-related battery complies with this requirement.
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b.
Section 4.8.2.3.2.c.5 of the TS requires verification at least once per 18 months that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual emergency loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the battery is subjected to a battery service test or the individual cell voltage does not decrease below 1.75 volts when the battery is subjected to the equivalent load profile based on anticipated breaker operations required during loss-of-offsite power (LOSP) and loss-of-coolant accident (LOCA) conditions as described in the Final Safety Analysis Report.
Provide the current vs. time curve for the equivalent load profile referred to above and verify that each auxiliary building battery is tested to this profile.
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Issue 2:
l Region II questioned the licensee's justification for the adequacy of l
DC voltage to close 4kV breakers following a loss of AC power. The l
vendor's data for 4kV breaker close and trip coils had been ignored in favor of measured test data that produced a lower calculation volt drop.
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These measurements were made by the licensee on a small sample and l
employed neither statistical sampling techniques nor an approved test l
procedure.
Further, new acceptance values were not included in the procurement specification for replacement parts to ensure capabilities of the replacements.
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a.
Using the auxiliary building battery minimum voltage values as specified in the TS and using 4 amperes in calculating voltage drops to the 4kV breaker closing and tripping coils, provide DC voltage values at the terminals of the closing and tripping coils.
b.
Provide information from the manufacturer's documents that describes the method / approach to be used to determine the minimum current values necessary to operate the 4kV breaker closing and tripping coils.
Explain in detail why the measured current values obtained using a small sample population and employing neither statistical sampling techniques nor an approved test procedure are considered technically adequate and acceptable for use in calculations.
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Provide technical justification for not including the new acceptable current values in procurement specifications for 4kV' breaker replacement parts to provide assurance of the capabilities of these parts.
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3.
Inspection Followup Item 348, 364/92-17-13, Inadequate Motor Overload
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Protection l
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The reactor coolant pump motors have narrow tolerances between the I
thermal limits and the full load and starting currents. Their protective relay settings were selected to avoid spurious tripping during acceleration and running, thus making it impossible to protect the motors on overloads. Both relays used in motor protection (IAC66K and C0ll) were set above the thermal limits of the motor for i
acceleration and running conditions. Therefore, the motors were not provided with protection for overload conditions. The thermal limit curve indicated that, if the motors were allowed to run at 90 percent of the rated voltage, they would be damaged in nine minutes, i
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Reauest for licensee:
a.
Provide time-current characteristic curves for equipment items i
associated with the reactor coolant pump motors.
The response should include:
(1) inrush current (1s) curves with 80 percent and 100 percent bus voltage available, (2) both relay curves used for j
the RCP motor protection (IAC66K and C011), (3) thermal limit curves j
for RCP motor stator and rotor windings, (4) RCP motor bus upstream main breaker thermal limit curves, (5) RCP motor cable thermal limit curves, (6) containment penetration thermal limit curves, and (7) fuse thermal limit curves, if fuses are connected in series between the RCP motor feeder breaker and the RCP motor.
In addition, i
provide relay trip setpoints for protective relays IAC66K and C011.
b.
Assuming that the thermal limits of RCP motor windings are exceeded l
frequently, discuss the possibility that the RCP seal could be damaged due to overheating and/or excessive vibration.
If this is
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not considered possible, provide a detailed explanation to include technical bases in the response.
Issue 2:
Long time delay trip settings of circuit breakers supplying 600 Vac i
motors appeared to be too high, permitting operation of the motors at i
excessive currents. A typical example related to the control rod drive i
mechanism motor generator set lA motor (150 HP), where the circuit breaker setting could cause tripping at 160 percent of motor full load current (bandwidth of 132 to 162 percent). ANSI Standard C37.16-1988 recommends that the trip device be set not greater than 130 percent of i
full load rating for a service factor of 1.0 (the service factor for the motors in question).
Reauest for licensee:
a.
Provide the technical basis for the Bechtel criteria used to select circuit breaker settings at 160 percent of motor full load current.
b.
Explicitly identify the 600 Vac safety-related motors whose long time delay trip settings of their feeder breakers are set at 160 percent of motor full load current.
c.
Based on historical data, explicitly identify the 600 Vac safety-related motors that were damaged due to overload because their feeder breaker has been set at 160 percent of motor full load I
current.
d.
Provide time-current characteristic coordination curves for 600 Vac l
safety-related motors. The response should include a motor feeder breaker curve and the upstream breaker curve.
4.
Inspection Followup Item 348, 364/92-17-14, No Periodic Testing to i
Verify Continued Capabilities of Most Safety-Related Molded Case Circuit Breakers (MCCBs).
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! s 1 s Issue:
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The licensee originally established a program for extensive testing of I
MCCBs but reduced the scope of the program based on a letter from the Nuclear Management and Resources Council (NUMARC).
The NUMARC letter, dated October 17, 1990, noted that there had been problems with factory i
calibrations but that these problems could be identified and corrected through acceptance or pre / post maintenance testing.
It concluded by stating that "it is not clear that the perceived benefits achieved by l
periodic testing of MCCBs are commensurate with the resources required j
to conduct such testing." The licensee currently tests MCCBs for l
safety-related use upon receipt or prior to use.
However, once installed in the plant, the only routine testing of MCCBs is the testing required by Unit 2 TS for containment penetration MCCBs (Note: The Unit.
l 1 TS do not require testing of containment penetration MCCBs.). The licensee plans no other preventive maintenance or testing of MCCBs over the life of the plant. Aging could cause the lubricant in the breaker to dry out or other degradation might occur which would affect breaker t
trip calibration.
Region II finds that many other licensees perform periodic testing of their more important MCCBs and believes this is appropriate over the 40 year life of a plant.
Reauest for Licensee:
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Identify safety-related and non-safety-related MCCBs that have been tested.
In addition, provide detailed descriptive information explicitly addressing the number of MCCBs tested, the number of MCCB failures, and the types of MCCB failures that have been identified for i
these breakers as a result of this testing.
Further, provide a description of any analysis of the testing results that has been used to determine actual failure rates (include actual failure rates if available) for the MCCBs tested.
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