ML20112G139

Forwards Summary of Disposition of Justification for Continued Operation Re Environ Qualification of safety-related Electrical Equipment in Effect When Discussed in But No Longer in Effect

ML20112G139
Person / Time
Site:	Fort Calhoun
Issue date:	01/10/1985
From:	Andrews R OMAHA PUBLIC POWER DISTRICT
To:	Harold Denton Office of Nuclear Reactor Regulation
References
LIC-85-009, LIC-85-9, NUDOCS 8501160207
Download: ML20112G139 (14)

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Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 402/536-4000 Janua ry 10, 1985 LIC-85-009 Mr. Harold R. Denton, Director U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Licensing Operating Reactors Branch No. 3 Washington, D.C. 20555

Reference:

Docket No. 50-285

Dear Mr. Denton:

Envirorriental Qualification of Safety-Related Electrical Equipment In a letter dated May 31, 1984, the Omaha Public Power District provided the NRC with documentation based on a March 23, 1984, meeting between District personnel and your staf f. This meeting was to provide the staff with curreat information concerning outstanding items identified by Franklin Research Cen-ter in their Technical Evaluation Report, to discuss the licensee's proposed resolution of these items and to establish the work remaining to be done to enable the staff to issue its final Safety Evaluation Report (SER).

This May 31, 1984, submittal included, among other items, a discussion of com-pliance with 10 CFR 50.49, the current status of equipment qualification, and discussion of those justifications for continued operation (JC0's) in ef fect at that time.

A number of events have transpired since the May 31, 1984, l e tter. Notably, the District has completed qualification of all but one item, and has sub-mitted a request for extension for that item.

Attachment 1 to this letter provides discussion which summarizes the disposi-tion of those JC0's in effect according to the May 31, 1984, letter which are no longer in effect. Attachment 2 provides additional information concerning 8501160207 850110 p DR ADOCK 05000285 PDR 1

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Mr. Harold R. Denton January 10, 1985 Page Two one of the notes in Attachment 2 of the May 31, 1984, letter. These attach-ments are clarifying in nature and do not alter our previous conclusions.

Sinc.grely,

/I8 R. L. Andrews Division Manager Nuclear Production RLA/DJM/rh-E

, Attachments cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hanpshire Avenue, N.E.

Washington, D.C. 20036 Mr. E. G. Tourigny, NRC Project Manager Mr. L. A. Yandell, Senior Resident Inspector

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Attachment 1 k

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Justifications for Continued Operation i The District's May 31, 1984, letter b'riefly detailed those items whose quali-fication was not yet established and, as such, required a Justification for Continued Operation (JCO). At that time, the following items (identified by Franklin Research Center Technical Evaluation Report (TER) number) were noted as still requiring qualification:

TER #1 TER #2 TER #3 Foxboro Transmitters i TER #4 TER #85 TER #86 TER #87 Cable -

TER #88 TER #89 TER #92 Cable Splices TER #99 Conax Cable Penetrations At the time of the May 31, 1984 submittal, the District had requested an ex-tension until September 30, 1984, for these items (This was done in a letter dated April 3,1984 which appeared also as Attachment 7 to the May 31, 1984, letter) . The requested extension to complete qualification was granted in a i

letter dated May 18, 1984. Extension was required for TER Items 1 through 4 to complete the review of recently-received documentation. The extension for the remaining items was requested because qualification testing was in pro-gress and would not be completed in time to meet the "end of 1984 refueling 7

outage" date for qualification.

The qualification of the Foxboro transmitters has been carpleted and the qual-ification documentation is on file. Justifications for continued operation

for TER Items 1 through 4 are no longer required.

The District conducted testing on TER Items 85, 86, 87, 88, 89, 92, and 99.

These items were being tested as a system to answer questions raised by Franklin Research Center -(FRC) in the TER. FRC raised questions concerning the interfaces between these individual camponents.

The District constructed test specimens to simulate the installed configura-

< - tion at Fort Calhoun Station and began testing. As detailed in previous cor-respondence, numerous testing problems were encountered prior to May 31, 1984. Before the end of the refueling outage, testing had shown that Itan

1. 92, Cable Splices, which were part of original plant equipment, had failed.

Consequently, original equipment splices on safety-related EEQ equipment in containment were replaced with qualified Raychem Splices. Item #92 was quali-7, fied by replacement and a JC0 was no' longer necessary. Meanwhile, the Dis-

. . trict instructed the testing laboratory to continue with the test sequence.

Testing continued from the point where the failure had occurred. During the

, resumed testing and prior to plant startup, abnormal conditions associated with the Conax cable penetration subassemblies appeared. Specifica1ly, the

' insulation on the lead wires to the penetration subassemblies had degraded such that electrical integrity was no longer assured. The District performed 4

' analyses to determine the failure mechanism. It was determined that the cause of failure was induced by environmental stress parameters which would only result from a large Break LOCA.

As detailed in the District's letter dated July 3,1984, we informed you of this and requested an additional extension until November,1985, for TER item

1. 99. A revised JC0 was submitted. This matter is being handled separately.

After the penetration lead wire failure, the District elected to resume the testing to address aspects of the qualification of Rockbestos cable. IE Information Notice 84-44 raised questions regarding the validity of qualifica-tion data for Rockbestos cable because of discrepancies in Rockbestos' QA pro-gram. During the resumed testing (of the same cable which had been part of the splice and subassembly failures), certain anamalies appeared. The Dis-trict ceased testing and began an in-depth analysis.

It was noted that after the penetration lead wire failure the Rockbestos cable was subjected to significantly abnormal stress in the way of handling, cutting, etc., not indicative of normal operating practices. This included misapplication of Raychem heat-shrink sleeves to cable which had already been through the equivalent of a LOCA. Additionally, based upon the sequence of events, the District determined that the as-tested configuration which result-ed in the anomalies was not indicative of installed conditions at Fort Cal-houn Station.

The original intent of the testing was to demonstrate the qualification of the penetration-splice-cable interface, not cable qualification. Testing done as described is not indicative of installed conditions and should be considered invalid. The District has again reviewed existing cable test data and performed analyses consistent with the D0R Guidelines. The District be-lieves qualification is demonstrated based upon this data and appropriate

, documentation is on file. Thus, items 85, 86, 87, 88 and 89 are qualified and a JC0 is no longer needed.

In summary, the District currently has one JC0 in effect, that being for TER

' Item #99. Other JCO's are no longer necessary.

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9 Attachment 2 1

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k Fort Calhoun Station f' Environmental Qualification of Safety-Related Electrical Equipment Revision to TER Meeting Response Item 7 i:

Discussion As pa'rt of the District's TER Meeting Response to TER Item 7, the District included a. discussion (identified as Attachment 2-Note 2 to the response) to provide more infomation regarding the limit switches on certain pilot sole-

< noid air operated valves in containment. In this note, nine valves were identified as having limit switches which could be submerged in the event of a large break LOCA.

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This note discussed the District's belief that limit switches are qualified for this service. However, to provide added assurance of safe operation, a fuse would be installed in the negative lead of the limit switch which could

-be grounded to insure that if another positive lead grounded somewhere else, this fuse would blow, isolating the limit switch and insuring valve operabil-

- i ty. This emergency modification was installed during the 1984 refueling

. outage.

During a post-modification document update, it was discovered that the de-signed location of the fuses did not accomplish the intended purpose. The

-lead which was. already isolated from ground by the indicating lights was fused.- Because of this, the fuse protection discussed in the preceding

paragraph is not available.

The District still believes that the system (limit switch, conductor' seal assembly, cable splice. and cable) will function properly. This is discussed

~in detail in the revised Note 2 contained in Appendix A. Changes from the E - currently docketed Note 2 are designated by vertical black lines in the right hand margin.

The District has also completed a failure mode and effects review to deter-p mine if there is a safety concern in the event of failure. This review is .

. provided in Appendix B.

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' Based on the belief in qualification for submergence and support of the fail .

. ure modes and effects review the District believes there is no concern for F safe operation or need for an extension.

. Plant Modification, Operation, or Administrative Control Changes Since there always exists the potential for even qualified equipment failure, the District has reviewed the need for any modification to enhance the' plant >

safety as related to these limit switches.

Based'on the failure modes and effects review, it appears thn operational en-hancements can be gained for only HCV-238-and HCV-239 (long Mrm core cool-

.ing), and HCV-438A and HCV-438C (component cooling water to kc pcmp seals)

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~ since these valves have both .open and. closed accident functions.

p LThe District?is reviewing the replacement of the fuse in the potentially

. grounded lead with a. higher value fuse to clear the other grounded circuit

=.first. , 'In -addition,"a note will be placed in the emergency ' procedures to key the- operator to these valves, in~ the event position indication is lost, and L to have-the circuit checked -(and re-fused should simultaneous' grounds blow the ' oversized fuse).

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In the long tenn (1985 refueling outage) the modification as originally des-cribed in Attachment 2, Note 2, to the District's letter dated May 18, 1984, will be complete.

' Conclusion Based on the belief of qualification, identification that no safety concern exists, and upgrade of administrative controls, the District believes this item is closed. It is requested that the TER response be updated with infor-mation as provided in Appendix A.

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9 Appendix A Revised Note 2 4

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! FRC-TER ITEM BY ITEM RESOLUTION NOTES 4

TER Items 1, 2, 3, 4 - Foxboro Pressure and Differential Pressure Trans-E - l' .

mitters.

! The District is upgrading all transmitters which are required to oper-ate in a harsh environment to latest models of NE-10 Series Foxboro which have been tested to meet the requirement of IEEE/323-1974. The

. District will have completed any necessary hardware upgrade by the end of the current refueling outage scheduled to be completed in early May i of 1984. Due to the late arrival of the test reports and the work re-l quired to insure similarity, review power supply voltages, review accur- i acy data, update the qualified life program and update the central file

.an extension to September.30,1984 has been requested. This includes a justification for continued operation.

t The transmitter upgrade will resolve the radiation exposure, T/P (temp-

.erature and pressure) exposure duration, and provide a transmitter qual-ifled 11fe which will be factored into the District's Qualifled Life 4

< Program. All transmitters are now installed above the flood level.

' 2. . TER Items 7 - NAMCO Limit Switches

! The limit switches listed in Table 1-Note 2 are limit switches which are installed on valves whose location is below the maximum projected

! flood level in containment. It is the District's engineering judgement

. 7that the limit switch system (limit switch, conductor seal assembly, ,

cable splice, and cable), is qualified for operation while submerged

following a LOCA design basis event.

' The NAMCO limit- switch report (#QTR 105)-indicates that through test-E equipment malfunction, the limit switches were in fact submerged during testing. .This is a very straight-forward indication of qualificatio_n.

F The only additional' hardware requirement for these limit switches is that the cable entrance b'e sealed.. To accomplish this, a Conax elec-

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[ trical conductor seal assembly was used which Lis trated for-75.-psig 'and l chemical . spray providing an adequate seal.

L The Raychem sleeves, and Pyrotrol III cable were also LOCA tested. The

- District believes that the only concern here is insulation degradation

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p due to submergence., Please. note that only a short length of cable,'if any, would potentially be submerged under worst case conditions (See -

- ' Table 1-Note 2 for elevation). It is the District's-belief that catas-ttrophic failure will not occur,:but rather, the insulation resistance will bc reduced.- 'It is also the District's belief that since the volt-age is relatively 1ow - nominal 130:VDC - and used with'a 10 amp fused supply,'a small increase in leakage currents will not affect the cir-cuit operation.

The. District believes this adequately establishes the acceptability of.

the use of this equipment for this. application.

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TABLE 1 NOTE 2 I Tag Limit Switch Elevation

• Accident Fail No. System Function Position Position Position HCV241 Let Down Position 1,000' C C CVCS Indication Containment

Isolation HCV-238 Charging Position 999' 0 Accident 0 Long Term Indication **C Long Term Core Cooling Core Cooling HCV239 Charging -

Position 1,000' 0 Accident 0 Long Tenn Indication **C Long Tenn Core Cooling Core Cooling

-HCV438A CCW to RC' Position 996' 9" 0 0 Pump Seals Indication **Close on CIAS Containment + Low CCW Isolation Pressure HCV438C CCW to RC Position 997' 0 0 Pump Seals Indication **Close on CIAS Cor.tainment + Low CCW

Isolation Pressure HCV467A- CCW to Position 999' 5" C C Nuclear Indication Detector Well Cooling Containment Isolation HCV467C CCW to Position 999' 2" C C Nuclear Indication Detector .

.Weli Cooling Containment Isolation l .HCV1387A B1owdown Position- 1000' C C I

from SGB Indication

-Containment Isolation HCV1388A Blowdown Position 1000' C C l

from SGA: Indication L Containment Isolation

• MAXIMUM PROJECTED FLOOD LEVEL 1000.9' INCLUDING ENTIRE RCS INVENTORY

** EQUIPPED WITH AIR ACCUMULATORS FOR REPOSITIONING 2

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o Appendix 8 Failure Modes and Effects Review 1

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Failure Modes and Effects Review Criteria Discussion A review has been conducted to detennine if a submergence induced failure of the NAMCO limit switches addressed in Note 2 of the TER Meeting Response poses any safety concern. The requirements can be grouped into 3 categories:

1. Those valves which have two Large Break LOCA functions which directly contribute to reactor safety. These are HCV-238 and HCV-239 which initially open to . inject concentrated boric acid into the RCS and later close to permit the HPSI pump to use auxiliary spray for long tenn core cooling.

2. Those valves which have two accident positions but are not used in a large break LOCA. These are HCV-438A and HCV-438C which provide compo-nent cooling water (CCW) to the Reactor Coolant Pump Seals and Motor.

These normally remain open during a DBA and close automatically only on low CCW pressure and containment isolation.

3. Those valves which go to their accident position and fail to the same

. posi tion. These valves are HCV-241, HCV-467A, HCV-467C, HCV-1387A, and HCV-1388A.

Failure Modes The only two failure modes expected are either a short of the switch or a ground of the floating lead. A short would cause dual position indication but would not cause loss of function. A ground could cause the circuit fuse to blow resulting in the valve going to its fail position. The loss cf func-tion is considered to be the only area which could result in a safety con--

Cern.

Failure Effects

- Category 1. HCV-238 and HCV-239: The concern here is the inability to close the valves for long term core cooling.

Long term core cooling is not established until several hours af ter-a LOCA. iii this case, adequate time exists to clear the bus of grounds and close the valves to correct the problem.

Actual inability to establish long tenn -

core cooling would require limit switch failure, failure to clear the bus of grounds, and failure of valves redundant to 238 and 239.

Category 2. HCV-438A and HCV-438C: The concern here is the inability to close the valves on low CCW pressure.

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This is a safety concern only if CCW pressure is low, the RCP cooling system (ASME Section III, Class II) fails, the limit switches fail, and the redundant valves outside containment fail .

Category 3. HCV-241, HCV-467A.

HCV-46/C, HCV-138/A and HGV-1388A: There is no concern here, no credible failure of the ifmit switches can prevent the valves from accomplishing their acci-dent function of containment isolation.

Conclusion Due to the multiplicity of failures required, or lack of failure effect, there should be no safety concern.

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