ML20042H020
| ML20042H020 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/10/1990 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| CON-NRC-90-046, CON-NRC-90-46 VPNPD-90-222, NUDOCS 9005170065 | |
| Download: ML20042H020 (11) | |
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' Electnc POWER COMPA'T 231 W MicNgon. Po box 2046. Miwoukoo. WI 53201 (414)221-2345 VPNPD 222 NRC 046 May 10, 1990 Document' Control Desk U.S. NUCLEAR REGULATORY COMMISSION 1
Mail Station Pl-137 Washington, D'.C.
20555 Gentlemen i
DOCKET NOS. 50-266 AND 50-301 ELECTRICAL INSPECTION EXIT MEETING 1
POINT BEACH NUCLEAR PLANT On April 17, 1990, a meeting was held in the Wisconsin Electric 3
corporate office to discuss the preliminary findings of a.special electrical inspection conducted by the NRC staff during the period from March 19 through April 12, 1990.
The objective of.this inspection was to review the adequacy of.the electrical distribution system at'the Point Beach Nuclear Plant.
NRC and contractor inspection team members were atLour corporate offices.
from March'19 - 23, 1990 and at.the Point Beach Nuclear Plant from
-April'2 - 11, 1990.
During an informallmeeting at'the plant on
-t April 11 and at the exit meeting on April 17,: the inspection' team identified 23 items of concern'resulting from this inspection.
Wisconsin Electric responded byfprovidingLan.up-to-date status report on each of the 23 items.
Attached to this letter is a' listing of the 23 items:of' concern identified to us by the inspection team.
Many'of-theseLitems were initially identified during the-course'of the inspection and prompted various corrective actions taken by Wisconsin Electric i
personnel during and immediately after the inspection. 1The j
attachment identifies those' actions already taken or planned to i
address the specific concerns and-serves to document the information provided by our staff to the.NRC team members at the l
April 17, 1990 meeting.
We believe these-actions demonstrate our 1
responsiveness to those concerns identified during the inspection i
and our desire to take the required remedial steps on a timely and' l
thorough basis.
4 1
9003170065 90osto e
f I
$DR ADOCK 05000266 l
l PDC i
ilsuMhmetIlinomh thwrorpvaSn 1
Document Control Desk M&y 10, 1990 Page 2 The NRC inspection team noted that the inspection was difficult i
due to the lack of supporting engineering information.
As discussed at the April 17 exit meeting, we have initiated a major design basis reconstitution project which will take place over the next several years.
This project is particularly difficult because Point Beach was a turnkey plant.
However, completion of i
this project will facilitate the retrieval of supporting engineering information regarding the plant design basis.
4 Please notify us if any of the information identified herein does not correspond to your understanding of the information presented during the April 17 meeting.
We will, of course, respond to your inspection report on these matters when issued.
-1 Very truly yours,
{ff oc C. W. Fay l
Vice President Nuclear Power Attachment Copy to NRC Resident Inspector Regional Administrator, Region III CWK/alg j
bcc:
J. W.
Boston i
G. M.
Krieser E. J.
Lipke R. A. Newton J. J.
Zach i
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i DISCUSSION OF ITEM IDENTIFIED AT ELECTRICAL INSPECTION EXIT MEETING APRIL 17, 1990 The items of concern identified at the Electrical Inspection Exit Meeting on April 17, 1990, were categorized in four major areas.
These categories were:
1.
Single / Common Mode Failure 2.
Emergency Diesel Generator Operability 3.
Battery ar d DC Systems 4.
Miscellaneous This attachment presents the specific items identified in each category and a brief discussion of our completed or anticipated response.to each item.
1.
Single / Common Mode Failure
-A.
Inadequate Redundant Cable Separation During the inspection, NRC identified two cases of inadequate cable separation and questioned whether other cases existed.
By using the Point Beach Cable and Raceway Data System (CARDS), 25 cases of inadequate cable separation involving redundant train cables in the same raceway were identified.
By the date of the exit meeting, we had completed an operability evaluation for all of
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these cases.
Immediate corrective actions included modifications to the Unit 2 steam driven auxiliary feedwater. pump starting circuit and the component cooling water pump starting circuits.
Procedure changes for operation of the power operated relief valves were also-completed.
In no other cases were plant operability problems identified.
In the near term, we will correct each of the 25 identified situations by physical modification, correction.of errors in the CARDS database, or evaluation.
In the longer term, we will activate.
features of the CARDS program which provide for an automatic check of cable separation violations.
In addition, we will take further measures to assure the accuracy of the CARDS data.
B.
480 Volt Bus Tie Breakers Safety-related Bus 1(2)B03 has a single maintenance breaker capable of tying it to 1(2)B04.
This breaker is provided for maintenance and operational flexibility.
The NRC team expressed the concern that a single DC failure in the control circuitry could cause this breaker to close and tie the independent safety-related busses together.
Our corrective action included removing the DC control power fuses so that DC power cannot operate the breaker a
- end po: ting of operctor: cide cnd a night order bookLentry-so that the-operators are= aware of this action.
Operating procedures have also been modified'to reflect this action.
C.
DC Control Power Knife Switches DC control power to each vital 4160 v. and 480 v.. bus includes a knife switch, which allows selection of normal' or alternate power supply.
Concern.was expressed by-the:
NRC staff regarding the seismic adequacy of these switches.
We have calculated that the peak seismic force acting on the switch center of gravity during a SSE would be about 1,671b-f.
At the time of the meeting, field measurements had been made on four knife-switches'which-determined a minimum opening force-of between 3.0 and 6.0lb-f.
We thus believe that the switches will remain-closed during a seismic event-and this issue should not be e common mode failure concern.
We are planning to verify this by testing the remaining switches.
D.
4160 Volt Tie Breaker The single maintenance breaker lbetween 1(2)A05 and 1(2)A06 has been racked out in accordance with a single failure concern identified in the past.
The inspection team questioned the seismic adequacy of the breaker while in-this racked out position.
We have now physically removed these breakers from the cubicle and placed them in storage outside the safe shutdown areas pending completion of further analysis.
E.
Component Cooling Water Pump Motor Auto Start Circuit The component cooling water (CCW); systems at Point Beach were not originally designed as.a nuclear safety-related system.
In 1988, we undertook a-long' term program to upgrade the CCW system to nuclear safety-related.
While the NRC team recognized-that'our upgrade was not yet complete, it identified an instance of inadequate cable separation.
As previously noted under Item.1.A, we have completed a temporary modification to' disable the auto.
start circuit for the "B" CCW pumpssand have designated those pumps as the preferred pump for continuous operation.
We intend to reroute the cables to provide adequate separation.
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j F.
Unit 2 Auxiliary Feedwater Pumps (AFW) Autostart Circuit Separation 3
During the CARDS review ~of cable separation as discussed in 1.A above, we identified that the auto start circuitry for the motor-operated steam supply valves to the Unit 2 turbine driven AFW pump did not meet our separation l
criteria.
On April 3, 1990, we declared the Unit 2 tur-bine driven AFW pump out of service and entered a-72 hour l:
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bino driven.AFW pump out of service cnd entured c.72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
" Technical' Specification LCO.
The cables were rerouted ~to provide _ required separation and a satisfactory > test of the complete modification was completed on April'5,~1990.
No further-corrective action is necessary.
G.'
Unit 1 MOV-826C Miswired A wiring discrepancy between the as-built component and the drawing was identified during the inspection.
We evaluated _this discrepancy for equipment operability concerns _and identified no problems.
We believe this to be an isolated discrepancy.
We have defined the correct' condition and wired the valve accordingly.
2.
Emergency Diesel Generator (EDG) System Operability A.
EDG Steady State Loading Analysis Accuracy.
The need for a revised loading analysis had been previously identified by Wisconsin Electric personnel and a ocntractor had been hired to complete such an analysis.
The analysis was completed during-the_ inspection.
The 3
revised calculations document a worst case diesel loading of 2939 KW compared to a 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> rating of the.EDG of 2963 KW.
Thus, there were no operability-concerns with diesel loading.
The NRC team then questioned whether the pressurizer heaters and a containment ventilation fan for the unaffected unit should be added to the loading.
We subsequently determined that the pressurizer heaters 1were not needed but'that the containment fan load'should be added.
However, we also determined that certain auxiliary building fans included in'our analysis are'also not needed.
These changes resulted 'irr a net decrease in loading, and the analysis remains valid for operability determination.
l B.
EDG EOP Load List Inaccuracy 1
In the most recent revision to this EOP, specific guidance L
has been provided to the operators to manage EDG loads.
We had recognized that-this guidance. contained some con-f.11cting information with respect to diesel loading and included revising this list as part of the diesel loading study.
The_NRC concern.was that: the present guidance to i
the operators might cause them to violate the design'or licensing basis.
The questioned diesel loading list will' be revised and corrected.
In addition, detailed EOP L
. reviews were conducted by engineers, operators,.and shift l
superintendents on April 10 and 11, and temporary changes to the EOP were approved and distributed on April 12-to provide further guidance to operators for managing diesel i
loads during an accident.
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C..
EDG M: tor.Incccurccy The NRC expressed concern regarding the frequency.of EDG load meter calibration.
The meters which indicate diesel generator load on the main control boards were last-
- calibrated in 1988.
After the NRC' identified this concern, maintenance' work. requests were issued'to.
recalibrate the meters.
Recalibration adjusted the meter to read accurately-(-0% to +0.5%')'at 350 KW.
This has been done for both EDGs.
We are continuing to evaluate the overall metering error, its effects on allowable EDG loading, the potential need for additional or: replacement metering, and the need for'more frequent calibration.
This is. expected to be done by July 1, 1990.
D.
EDG Puel Oil System Piping Seismic Design The NRC inspection team questioned the seismic design basis for the EDGs fuel oil supply system.
'In the process of evaluating the seismic capability of the EDG fuel oil-
- i system piping during the inspection, we determined that.
. piping supports in the fuel oil pumphouse would not meet operability or code allowable stress limits in the event i
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of an SSE occurrence.
As.a result, we declared the EDGs not operable and requested and received from the NRC a temporary waiver of compliance for seven days-to correct' the piping supports.
We completed an; analysis of the transfer piping from the emergency fuel oil tanks to the underground header on April 10, 1990.~
Redesign of the piping supports was completed on April 12 and the supports were installed by April 15, 1990.
Analysis of the fuel oil piping in-the control building was also conducted... Stresses,for this fuel oil piping meet operability allowables but not code allowables. ' We -
plan to complete piping' support modifications by the summer of 1990.
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In addition to the piping analysis,1we also conducted evaluations of the~ seismic adequacy of the fuel' oil pump house structure; the spiral stairs in-the pumphouse; the L
power supplies for the transfer pumps in-the fuel oil pump l
house, facade, and turbine hall; and.the fire protection
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header piping-in the fuel oil pump house.
From these l
evaluations, we determined that the concrete structure ~of the fuel oil pumphouse is adequate to resist-seismic l
loads.
The blockwallslof the pumphouse were judged to'be seismic by comparison to other seismic blockwalls, i
Calculations were done'to show the seismic. adequacy of the i
stairs in the pumphouse.. The transfer. pump power supplies in the pumphouse and the power supply. routed through'the facade were inspected and found-to be supported similar to l
other seismic installations.
The power supply routed in.
l the turbine hall requires further evaluation.
The~ fire 4--
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...'protsetion herderLin the fual oil pumphouso was shownLto be inadequate'for seismic loadings and the' header was conservatively isolated on April 16, 1990.
We have subsequently un-insolated the fire protection-piping.
because the risk of fire was considered to be greater-than the risk of fuel oil pumphouse flooding _in the event of a seismic occurrence.
Modification to-the fire protection piping in the fuel oil pumphouse to enhance the seismic capability will be pursued.
E.
EDG Fuel Oil-Quality The inspection team questioned the absence of a specified cloud point criterion for the fuel oil as suggested by-ASTM Standard D975.
Although we measure the cloud point in our fuel oil sampling, we have accepted fuel oil with cloud point criterion temperatures which exceed the most restrictive seasonal cloud point criterion given in the ASTM D975' standard.
We maintained that for safety-related applications, thefuel-intgeundergound, fuel-oilstorage tank is maintained above 30 F and therefore, our acceptance of fuel oil with a higher cloud point'than recommended in the standard is prudent.
We have committed to complete a thorough engineering evaluation of this issue by the fall of 1990.
F.
EDG Voltage Sensing Relay The FSAR states that the diesel generator output breakers will close when the diesel is up to speed and adequate generator output voltage is present.
The actual as-built circuitry includes a contact that closes'when the diesel is up-to speed and a contact that opens if field voltage does-not exist within six seconds of diesel speed '100
~ RPM; however, a direct interlock which. prevents closing of the output breakers based on measured voltage doesLnot.
exist and the NRC questioned the adeq'uacy of undervoltage protection.
In. order.to demonstrate the adequacy-of' L-diesel output voltage prior to closing the output L
breakers, a special test was performed on April 13 on~EDG G01, which measured = generator voltage as a function of time and compared it to other starting events.
This test demonstrated that a voltage of14432 volts ( 100% of I
nominal) was present when the breaker,would have received a close signal.
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A similar test was conducted on G02 on May-2 which demonstrated that a voltage of 4274 volts ( 100% of nominal) was present when'the breakers would have received a close signal.
We will also evaluate whether periodic testing should be conducted or whether a~ circuit L
modification is appropriate.
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1 3.
- Battery and DC Systems A.
Battery Sizing Calculations and Load Profile Not t
Conservative Formal battery sizing calculations had been created in l
response to a previous SSFI audit finding.
These calculations, however, did not quantify each load supplied by each battery but addressed several of them as being within the margin demonstrated by the calculations.
We have evaluated the magnitude of these!1oads versus trw margin in the battery sizing calculations to demonstrate the adequacy of the battery sizing calculations. - This evaluation was provided1to the inspectionJteam..
Future actions include revision of-the battery sizing calculations to incorporate the above mentioned evaluations.
.In addition, we will' complete a detailed study of all loads connected to the DC systems by mid 1991f Finally, our intention is.to install a new-y non-safety related battery by mid-1991.
This battery will be utilized to supply several large non-safety related-
. loads presently connected to the plant safety related batteries.
B.
High DC System Voltage l
The inspection team expressed a concern that the~ battery float voltage exceeds the manufacturer's recommendation of acceptable voltage levels for equipment supplied from.
associated DC systems.
We are changing our-procedures-to lower the float voltage to 132.75-volts.
We are also processing a modification request to add a fifth safety-related battery,-that will permit us to conduct equalizing charges on the existing batteries..This will-reduce the need to mai~ntain a high' float voltage.- We will also be installing a non-safety,related' battery which will permit removal of-several'large-non-safety related loads
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from the D01 and D02 station batteries.
This could allow for the removal of an additional' cell and permit a further reduction in float voltage.
s C.
Elgar Inverter' Low Voltage Cutout The NRC was concerned.that the DC input voltage level at which the inverter would automatically isolate was not clearly defined.
The inverter supplier provided'us with a letter on April 12 stating:
a)
The inverter will. work with or without low voltage cutout that protection in place, b)
Point Beach has no need for the protection t
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We h:va d3termin d that the_ control board that c:ts tha Iow voltage cutout can be reset and bench calibrated.
This will-be-done-to set the low voltge cutout at approximately 100 VDC.
D.
Swing Battery Charger Interlock
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The NRC expressed a concern that no interlock exists to prevent the interconnection of the plant main DC systems though the input breakers'from the swing battery charger.
However, similar interlock-does exist on the two DC systems added.in the early 1980's.
We have provided administrative controls and training a
which will prevent the-interconnection of the two DC systems.
We completed the installation of specific operator aids on April 20, 1990.- Future action will
. include installation of an interlock device by mid-1991.
E.
DC System Ground Check Criteria-The NRC team was concerned that our routine maintenance procedure has no acceptance criteria for ground checks.
This was identified by Wisconsin Electric personnel approximately one year ago.
At that time, we initiated a trend program to monitor ground readings.
Historically, we have not had hard ground problems.
By the.end of 1990, we will develop ground sensitivities for corrective action follow-up.
We are also placing a purchase order in May 1990 to procure more sensitive test equipment.
F.
DC Bus Breaker Protection This item was first identified by Wisconsin-Electric and involves the absence;of adequate interrupting capability for the thermal only breakers _ installed on the main DC busses.
An initial evaluation of this condition resulted-in some breaker replacements in the fall of 1989 along 1
with'other corrective actions.
This item was the subject.
of an NRC enforcement conference on January 16, 1990.
An i
overall evaluation of DC bus protection with recommended corrective action will be completed by July 1, 1990.
4.
Miscellaneous
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A.
Inadequate Relay Calibration Procedure The NRC team was concerned that_ test equipment, controls, and procedures used by an' expert non-nuclear company group i
did not meet 10CFR50 Appendix B requirements.. Upon identification of this item, we made a decision to stop work on all safety-related relay calibrations.
A-nonconformance report was initiated on April 6, 1990.
Our review of existing conditions verified that certain l
controls were adequately implemented and that additional
cctiono wara rcquircd cnd cubacquently impicm:ntad to allow continuation of the work.
We-have also initiated action to determine the acceptability of previous work and anticipate additional long-term corrective actions.
B.
Steam Damage to Cables
'The NRC team observed damage to cables in the vicinity of a steam vent.
The cables in question were physically examined.on April 4, 1990.
.The cables were determined to be non-safety related and maintenance work requests-were i
issued to wrap two of the cables which was completed on April 5.
Our evaluation demonstrated no operability.
concerns and established the absence of a harsh t
environment as defined by 10CFR50.49.
Additional i
recommendations concerning possible repair or replacement of:the cables will-be available by May 15, 1990.
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.i addition, a modification request has'been issued to prevent further: damage to the cables from.this specific steam vent.
C.
Molded-Case Circuit Breaker Testing The NRC asked about our position with. respect to the subject breakers.
We have not conducted periodic molded
. case circuit breaker testing and have no plans to do so at this time.
We will obtain verification of the adequacy of new or replacement breakers installed at Point Beach and will continue to monitor the industry practices, NRC.
positions, and the manufacturer's recommendations in this matter.
D.
Cable Tray Overfill During the inspection, a possible discrepancy between the FSAR criteria for cable tray capacity (30%) and the-original Bechtel design criteria was identified.
We have identified and written a nonconformance report listing-210-cases of cable tray capacities in excess of 30%.. As of April 17, we had' evaluated several of.these cases and found them to be acceptable.
We are. continuing with the evaluation of the remaining cases.
We intend to issue-written guidance'on cable tray fill requirements by July 1, 1990, and implement automatic cable rating features in our CARDS program by September-l', 1990.
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l U'Ddcument Control D;sk May 10, 1990 Page 2 g
The NRC inspection team noted that the inspection was difficult.
due to the lack of supporting engineering information.
As discussed at the April 17 exit meeting, we have initiated a major design basis reconstitution project which will take place over the next several years.
This project is particularly difficult because Point Beach was a turnkey plant.
However, completion of this project will facilitate the retrieval of supporting engineering information regarding the plant design basis.
Please notify us if any of the information identified herein does not correspond to your understanding of the information presented during the April 17 meeting.
We will, of course, respond to-your inspection report on these matters when issued.
Very truly yours,
{. Y C. W Fay Vic President Nuclear Power Attachment Copy to NRC Resident Inspector Regional Administrator, Region III l
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