ML17320A903
| ML17320A903 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/17/1984 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:0775G, AEP:NRC:775G, NUDOCS 8401200076 | |
| Download: ML17320A903 (45) | |
Text
j REGULATOR INFORMATION DISTRIBUTION TEM (RIDS)
ACCESSION NBRe800)200076 DOC ~ DATE: 80/0)/17 NOTARIZED:
NO DOCKET FACIL;50 315 Donald CD Cook Nuclear Power Planti Unit )i Indiana 8
05000315 50-316 Donald CD Cook Nuclear Power Plantr Unit 2i Indiana 8
05000316 AUTH~NAME AUTHOR AFFILIATION ALEXICHiM,P'.
Indiana 8 Michigan. Electric Co ~
RECIP ~ NAME RECIPIENT AFFILIATION DENTONiH ~ " ~
Office of Nuclear Reactor Regulationi Director
SUBJECT:
Forwards minutes of 830913 meeting w/American Electric Power Svc Corp
'in Bethesda<MD re proposed resolution for numerous environ qualification deficiencies identified by Franklin Research Ctriper 831024 request.
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'NDIANA 8 MI.CHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 Januax:v 17, l984 AEP:NRC:0775G Donald C. Cook Nuclear Plant Unit Nos.
1 and 2 Docket Nos. 50-315 and 50-316 License Nos.
DPR<<58 and DPR-74 MINUTES OF SEPTEMBER 13,
- 1983, NRC/AEPSC MEETING ON ENVIRONMENTAL QUALIFICATION Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
This letter responds to your letter dated October 24, 1983, to Mr.
John E. Dolan of the Indiana 5 Michigan Electric Company (IMECo).
More specifically, your letter granted approval of certain equipment environmental qualification deadline extensions which we requested under the provisions of 10 CFR 50.49. It also noted that the Office of Nuclear Reactor Regulation (NRR) would review the remainder of our environmental qualification program upon receipt of minutes for a meeting held at
- Bethesda, Maryland, on September 13,
- 1983, between members of your staff and American Electric Power Service Corporation (AEPSC) personnel.
The major topic of that meeting concerned our proposed r esolutions for numerous environmental qualification deficiencies identified by an NRR consultant
[i.e., Franklin Research Center (FRC)] in late October, 1982 [reference letter dated December 30, 1982, Mr. S. A. Varga (NRC) to Mr. John E. Dolan (IMECo)].
The requested meeting minutes are contained in Attachment 1 to this letter.
Additional information is also enclosed in Attachments 2, 3, and 4.
In particular, Attachment 2 contains the findings of our r eview regarding radiation qualification of certain penetration extension wire splices and field cable terminations.
Attachment 3 discusses the potential for spurious operation of motor and air operated valves due to short circuits which may result from equipment submergence.
Attachment 4
presents a reference which identifies the Westinghouse test reports applicable to the environmental qualification of the Donald C.
Cook Nuclear t
8401200076 840ii7 PDR ADOCK,O50003i5'"
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Plant hydrogen recombiners.
We trust that these documents will provide an adequate basis upon which your staff can complete review of our environmental qualification program, and we look forward to receipt of a final Safety Evaluation Report (SER) on this important topic.
Furthermore, we request that conditions 4.A and 4.B of Amendment No.
6 to Facility Operating License No. DPR-74 [reference letter dated June 16,
- 1978, R. S.
Boyd (NRC) to J. Tillinghast (IMECo)] be deleted from the Donald C. Cook Nuclear Plant Unit No.
2 operating lioense upon issuance of a final SER.
We believe that these license conditions regarding electric equipment environmental qualification should be treated as open items in the SER.
This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Very truly yours,
. P.
lexic Vice, President MPA/th Attachments cc:
John E. Dolan W. G. Smith, Jr. - Bridgman R. C. Callen G. Charnoff E.
R. Swanson - NRC Resident Inspector, Bridgman
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By letters dated September 23, 1981
[AEP:NRC:0578, G. P. Maloney (IMECo) to H. R. Denton (NRC)], and June 11, 1982 [AEP:NRC:0578B, R. S.
Hunter (IMECo) to H. R. Denton (NRC)], information concerning the environmental qualification of safety related electrical equipment at the Cook Plant was submitted to the NRC for review.
Although numerous submittals have been made regarding this topic, these two letters are of particular importance because their attachments contained, among other items of information, series of System Component Evaluation Worksheets (SCEWs) for those equipment items which we believed to be within the scope of the environmental qualification program to be reviewed as per the guidance contained in IE Bulletin 79-01B (whioh included the DOR Guidelines and NUREG-0588 as Enclosures 4 and 5, respectively).
The acceptability of our environmental qualification program, as outlined in part by the two submittals referenced
- above, was reviewed for the NRC Division of Engineering by the Franklin Research Center (FRC) as part of the Nuclear Reactor Regulation (NRR) Technical Assistance Program in support of NRC operating reactor licensing actions.
The FRC review findings for the Cook Plant are documented in a four-volume report
- entitled, "Technical Evaluation Report:
Review of Licensees'esolution of Outstanding Issues From NRC Equipment Environmental Qualification Safety Evaluation Reports" (TER), dated October 28, 1982.
As noted in our letter dated January 24, 1983
[AEP:NRC:0775, R. S. Hunter (IMECo) to H. R.
Denton (NRC)], we received the final volumes of'the TER on January 17, 1983 and soon thereafter notified the NRC that we believed a working group meeting would be beneficial to our understanding of the FRC TER findings.
On September 13,
- 1983, the requested working group, meeting took place at the NRC offices in Bethesda, Maryland, with Messrs.
J. Calvo, S. Kim, R. LaGrange, P. Shemanski, and D. Wigginton of the
- NRC, and Messrs.
M.
Alexich, L. Caso, J. Feinstein, H. Fouad, T. King, D. Medek, and R.
Shoberg of the American Electric Power Service Corporation (AEPSC) in attendance.
This attachment is intended to document our understanding of the discussions at that meeting (and in follow-up telephone conversations) in order to provide a basis for NRC review of the remainder of the Cook Plant environmental qualification program for electrical equipment important to safety.
The following sections of this attachment discuss our plans and schedule for resolving outstanding concerns identified in the FRC TER.
There are, however, certain points which must be noted:
The FRC TER deficiencies of.concern were identified for equipment items classified as being in NRC Categories I.B.
(Equipment Qualification Pending Modification) and II.A.
(Equipment Qualification Not Established).
No equipment items were identified in NRC Category II.B. (Equipment Not Qualified).
Equipment items in NRC Category IV. (Documentation
Not Mad~
vailable) were addressed in ou etter dated August 10, 1983
[AEP:NRC:0775D, M. P. Alexich (IMECo) to H. R. Denton (NRC)].
Tables I and II identify the speoific defioiencies of concern.
Deficiencies in documentation and similarity were discussed generically at the meeting and are therefore treated generioally herein (speoifio problems in documentation and similarity which relate to other deficiencies are, however, treated specifioally).
Defioienoies regarding aging qualification of equipment were discussed generically at the meeting and are therefore treated generioally herein.
No deficiencies were identified in test sequence, duration margin, or margins for those equipment items in NRC Categories I.B. and II.A.
All other deficiencies (e.g.,
chemical spray, submergence, etc.)
are treated specifically.
The number of deficiencies identified has not been ad)usted to account for certain equipment items which have been deleted from the scope of our environmental qualification program (e.g.,
Mercoid pressure switches, etc.).
Tables I and II, furthermore,
do not identify all equipment items within the scope of the environmental qualification final rule, 10 CFR 50.49.
- Rather, the two Tables summarize deficiencies identified by the FRC TER.
Our position regarding replacement equipment upgr ading is presented in Section V. of this Attachment.
In order to prove environmental qualification of electric equipment important to safety, certain documentation requirements must be met by power reactor licensees.
For the Cook Plant, NRC guidance with regard to these documentation requirements is contained in Section 8.0 of the DOR Guidelines, in NUREG-0588, in Plant Technical Specification 6.13.2, and in 10 CFR 50.49.
These documents basically require that complete and auditable records exist to prove environmental qualification, and that such records should describe the qualification method in sufficient detail to allow an independent reviewer to conclude that all applicable qualification criteria have been adequately addressed.
Mith regard to similarity between tested and installed electrical equipment, Section 5.2.2 of the DOR Guidelines indicates that the preferred method of proving similarity is to install equipment identical in design and material construction to a type test specimen.
Any deviations from this method are to be evaluated as part of the associated qualification documentation.
For the Cook Plant, thirty-seven documentation and forty-five similarity deficiencies were identified for equipment items in NRC Categories I.B. and II.A. (see Tables I and II).
Although the bulk of
'hese defioiencies were disoussed on a generic basW certa1n defioiencies and other documentation ooncerns were discussed speoifioally.
The speoifio items disoussed are as follows:
a)
Mobilux EP2 lubrioant (PRC TER Item No. 24, Unit Nos.
1 and 2).
We were-informed by the NRC staff that the data sheets provided by the vendor as proof of qualifioation would not be accepted without supporting dooumentation showing where the data sheet values oame from.
The NRC staff agreed that sinoe supporting proprietary information could be diffioult to obtain, dooumentation 1ndioating that we have audited the vendor's reoords for acceptability would constitute proof of qual1fioation.
b)
Sostman Model Nos.
11834B and 11901B and Rosemount Model Nos.
176KP and 176KS Resistanoe Temperature Detectors (RTDs)
(FRC TER Item Nos.
18, 19, 21, and 23, Units Nos.
1 and 2; FRC TER Item Nos, 111 and 122, Unit No. 2).
The FRC TER review for these RTDs indicate that the referenced test report (i.e.,
WCAP-9157, dated September 1977) did not fully address chemioal spray qualification parameters such as spray density and duration.
We agreed at the September 13,
- 1983, meeting to contact Westinghouse in order to verify and document that such testing was,
- indeed, performed adequately.
c) d)
Electrical termination located within containment (FRC TER Item No.
81, Unit Nos.
1 and 2).
The FRC TER review for this oonnection indicates that the qualification radiation dose does not envelope the speoified radiation service condition.
Although the corresponding System Component Evaluation Worksheets (SCEWs) in Attachments 4 and 5
to our Letter No. AEP:NRC:0578B, dated June 11, 1982, indicate that there is no problem with radiation qualification for this termination, we informed the NRC staff at the meeting that we would check our files once again to ensure that proof of qualification actually exists.
This review has since been completed, and the results are summarized in Attachment 2 to this submittal.
The termination has been found to be qualified for its radiation service conditions.
II The NRC staff asked us about the qualification status of Foxboro NE transmitters and Target Rock solenoid-actuated globe valves.
We informed the staff that completed test reports had not yet been received and/or reviewed for these equipment items, but that all available information to date (e.g.,
oonversations with vendors/suppliers, preliminary test results, etc.) did not indicate that problems were likely to be encountered.
We are still pursuing this topic at the present
- time, and expect to resolve these documentation problems on or before June 3,
1984.
e)
We informed the NRC staff that the documentation trail for some cables was questionable as proof of tested item versus installed item similarity.
We are essentially complete with this work now.
Work remains to organize the documentation files to assure complete traceability.
1-3
1I P
i II
On a generio basis, we informed the NRC staff hat, for many of the equipment items of concern, we believe that we already have adequate dooumentation in our files to prove environmental qualifioation.
For cases such as lubrioants, RTDs, eto.,
the Director of the Office of Nuclear Reactor Regulation has approved an extension of documentation
.- oompletion deadline to June 3, 1984 [see letter dated October 24,
- 1983, H.
R. Denton (NRC) to John E. Dolan (ZMECo)].
In oases where proof of similarity is a conoern, we will address those equipment items which have the greatest safety significance first, with eventual resolution of all ooncerns to be emphasized oommensurate to each equipment item's relative importanoe to safety.
Tracing of documentation trails will be completed via the use of Purchase Orders and vendor/supplier oorrespondence, as applicable.
The guidance provided by IEEE Standard 383-1974 Section 2.2 and Table I is also being used.
We expect to make documentation trails available at the Cook Plant site by June 30, 1984.
We also note that we are in the process of conduoting an internal audit of our environmental qualification files. If qualification summary (SCEW) sheets are updated as a result of this work, then they will be retained in our files and not transmitted to NRR for additional review.
Basic requirements relating to aging qualification of electrioal equipment are outlined in Section 7.0 of the DOR Guidelines and in NUREG-0588.
The FRC TER review of the Cook Plant environmental qualification program identified a total of one hundred fifty-seven deficiencies in the categories of aging evaluation, qualified life, aging simulation, and establishment of a surveillance/maintenance/replacement (SMR) program.
.As we explained to NRC staff at the September 13,
- 1983, meeting, we believe that we are taking adequate actions to ensure resolution of all aging qualification concerns.
In particular, as reported in our letter of March 4g 1983
[AEP:NRC:0775B, R. S. Hunter (IMECo) to H. R. Denton (NRC)],
we have contracted a consultant to perform aging analyses and provide input to an SMR program for the Cook Plant.
Furthermore, as stated in our letter dated August 10, 1983
[AEP:NRC:0775D, M. P. Alexich (IMECo) to H.
R. Denton (NRC)], if installed equipment is found to be non-qualified as the analyses are completed, then our plan is to establish an expedited program to replace that equipment and/or provide alternate measures to ensure accomplishment of the affected safety function (i.e., provide Justification for continued operation).
As a result of these commitments, we have already received approval for an extension of aging qualification deadlines to March 31,
- 1985, from the Director of the Office of Nuclear Reactor Regulation [see letter dated October 24,
- 1983, H. R. Denton (NRC) to John E. Dolan (IMECo)].
Additional aging topics discussed at the meeting are as follows:
a)
NRC staff suggested that our forthcoming SMR program also include equipment items with a forty-year qualified life.
Such items could be checked for degradation every five or ten years.
We will consider this suggestion in the development of the SMR program, but we note, 1-4
r k
~however, tha e are not committing to perfo estructive testing on oables or other long-life items at the present time.
b)
Aging simulation defioienoies are not strictly applioable to Cook
- Plant, beoause the equipment reviewed by FRC was to have been reviewed for compliance with the DOR Guidelines, not NUREG-0588.
The Cook Plant aging evaluation,and SMR program will, when completed, meet the intent of the DOR Guidelines.
IV.
In addition to the aging-related, documentation, and similarity defioiencies disoussed generically, we disoussed in detail a total of fifty-nine deficiencies identified in the FRC TER regarding requirements for peak temperature,
'peak pressure, duration, profile enveloping, steam
- exposure, chemical spray, submergence, radiation, test failures, funotional testing, and instrument aocuracies for those equipment items in NRC Categories I.B. and II.A.
The equipment items of concern and the relative breakdown of these defioienoies are presented in Tables I and II.
Proposed resolutions for these deficienoies, as discussed with NRC staff, are presented below:
a)
Limitorque Motor Valve Actuator (MVA) Model No.
SMB000 (FRC TER Item No.
1, Unit Nos.
1 and 2).
Ident1fied deficiency:
ohem1cal spray.
This item is the MVAs for the air reoirculation backdraft dampers.
Due to their location within containment, these MVAs are not subject to direct spray impingement.
Furthermore, these MVAs have a required" operat1ng time of thirty minutes.
The FRC TER indicated that the effects of chemical deposition resulting from exposure to a caustic environment must be addressed.
The NRC staff agreed with our contention, however, that this ooncern over chemical attack should be considered inapplicable to a motor operator of suoh relatively short operating time enclosed in a weather-proof housing.
b)
Limitorque MVA Model No.
SMBOO (FRC TER Item No. 2, Unit Nos.
1 and 2).
Identified deficiencies:
submergence and radiation.
W1th regard to submergence qualification, this MVA must close within 15 seconds of receipt of a Phase A containment isolation signal, and therefore performs its'function prior to being submerged.
Once submerged, spur1ous operation of the motor operated valve 1s prevented due to double breaking of the associated control circuit (see Attachment 3 for a discussion on AEPSC des1gn philosophy).
With regard to radiat1on qualif1cation of the MVA, it was noted that the Westinghouse supplied valve had Class H 1nsulation and was specified for nuclear service inside containment.
Furthermore, we noted that due to the short operating time and the use of white melamine (a
radiation resistant material) for the limit switch, we believed ther e was no rad1ation qualification concerns.
The NRC staff agreed with our contentions.
c)
Limitorque MVA Model Nos.
- SMB1, SMBOO, and SMB2 (FRC TER Item No. 3, Unit Nos.
1 and 2).
Identified deficiency:
submergence.
The ECCS infection and RHR normal cooling valves have a required operating time of 30 minutes.
As we informed NRC staff, the type test specimens were accidentally submerged during testing.
No failure was 1-5
experienced, the duration of the submerge could not be determined.
Nevertheless, as explained in Attachment 3, we believe spurious operation following submergence is prevented due to AEPSC~s design philosophy of double breaking the control circuits.
The NRC agreed with this argument, but questioned whether:
(a)
Cook Plant valves IMO-128 and ICM-111 and -129 would be submerged by a fee'dwater line break inside containment,
- and, (b) if so, could they properly function after being submerged.
We committed at the meeting to study this issue, and presently expect to resolve the NRC staff concerns by June 3, 1984.
d)
Limitorque MVA Model No.
SMBOO (FRC TER Item No. 5, Unit Nos.
1 and 2).
Identified defioiency:
chemical spray.
FRC questioned spray qualification for the Pressurizer PORV Block Valves'VAs because we did not explioitly address chemical deposition resulting from exposure to a caustic atmosphere.
As with the air recirculation backdraft damper MVA discussed in IV.a. above, we informed NRC staff that these MVAs were installed in a weather-proof housing away from direct spray impingement (i.e., in the Pressurizer doghouse),
and that they had a relatively short operating time (i.e.,
14 days) when compared with the time needed to cause significant chemical attack.
NRC staff agreed that these MVAs are considered qualified.
e)
Limitorque MVA Model No.
SMB2 (FRC TER Item No. 6, Unit Nos.
1 and 2).
Identified deficiency:
radiation.
As with the MVA discussed in IV.b. above, the MVAs for the RHR suction valves from the containment sump were supplied by Westinghouse with Class H insulation specified for nuclear service inside containment.
These MVAs are,
- however, installed outside the reactor containment, but near a maJor recirculation line.
These MVAs also utilize white melamine for the limit switch material.
Although these valves have a required operating time of 1 day, the NRC staff agreed with us that these MVAs are considered qualified.
f)
Reliance Electric Containment Spray Pump Motor Frame 05810P (FRC TER Item No.
14, Unit Nos.
1 and 2).
Identified deficiency:
radiation.
The FRC review for this pump motor indicates that it has no documented radiation withstandability.
Our plans to resolve this issue were discussed in Attachment 5 to our letter of May 20, 1983
[AEP:NRC:0775C, R. F. Hering (IMECo) to H.
R. Benton (NRC)], and reiterated at the September 13, 1983, meeting.
We have since received a radiation qualification verification test r eport from Reliance dated August 1983.
This report is currently under review, with initial findings indicating that the pump motor is fully qualified.
g)
Westinghouse Hydrogen Recombiner (FRC TER Item No.
15, Unit Nos.
1 and 2).
Identified deficiencies:
peak temperature,
- duration, profile enveloped, steam exposure.
The identification of these deficiencies in the FRC review appear to be based on the use of WCAP-7709-L, Supplement 7, in determining qualification parameters for the recombiner units.
Our qualification program,
- however, utilizes WCAP-7709-L, Supplement 2,
as a source document.
As discussed in Attachment 4 to this submittal, WCAP-7709-L, Supplement 1-6
.2, and not Slament 7,
ds appldoahle to theoh Plant.
We therefore believe that our hydrogen reoombiners are qualified.
h)
Sostman Model Nos.
11834B and 11901B and Rosemount Model Nos.
176KP and 176KS RTDs (FRC TER Item Nos.
18, 19, 20, 21, 22, and 23, Unit Nos.
1 and 2; FRC TER Item Nos.
111 and 122, Unit No. 2).
Identified deficiencies'(varies for different FRC TER Item Nos. - see Table I for speoific breakdown):
chem1cal
- spray, submergence, functional
- testing, and instrument accuracy.
The submergence deficiency identified for FRC TER Item No. 22 appears to be a case of a misplaced checkmark.
This deficiency should most likely have been identified as a chemical spray deficiency.
With regard to the identified spray deficiencies, it is noted that all of the RTDs were qualified under test report WCAP-9157, dated September 1977.
This
- report, however, does not give spray qualification parameters.
Our resolution of this issue (see II.b. above) is expected to be completed well in advance of the March 31, 1985, deadline approved by the Director of the Office of Nuclear Reactor Regulation [see letter
'ated October 24,
- 1983, H. R. Denton (NRC) to John E. Dolan (IMECo)].
Por the narrow range instruments, FRC identified concerns over functional testing and instrument accuracy are considered inapplioable since the RTDs perform their function (i.e., reactor trip) prior to "seeing" a hostile environment.
Similar concerns for the wide range instruments (used in post-accident monitoring) were identified by FRC, apparently due to the selection of calibration points for the RTDs used during the test sequence.
We committed at the September 13, 1983, meeting to document in our files that these concerns are inapplicable as long as identical calibration methodologies are used between tested and installed equipment.
The NRC staff agreed with this approach.
i)
Mobilux EP2 Lubricant (FRC TER Item No. 24, Unit Nos.
1 and 2).
Identified deficiencies:
peak temperature, peak pressure,
- duration, profile enveloped, and radiation.
All of these deficiencies were apparently identified by FRC because the reviewers would not accept vendor data sheets as a "stand alone" proof of qual1fication.
These deficiencies are expected to be resolved by June 3,
- 1984, along with resolution of the documentation deficiency (see II.a. above).
j)
Barton Model No. 763 pressure and Model No. 764 d1fferential pressure (D/P) transmitters (FRC TER Item Nos. 36, 37, 38, 39, 40, and 56, Unit Nos.
1 and 2; FRC TER Item Nos.
- 104, 107,
- 112, 115, and
- 116, Unit No. 2).
Identified deficiencies:
submergence, test failure, and instrument accuracy.
(Deficiencies vary between FRC TER Item Nos. - see Table I for specific breakdown.)
These transm1tters were originally qualified under Westinghouse test report NS-TMA-1950, dated September 1978.
The FRC identified deficiencies in test failure and instrument accuracy appear to ar1se from test conditions in which both narrow range and wide range pressure transmitters and D/P transmitters exceeded span specifications.
These supposed problems were addressed successfully by Westinghouse test reports NS-TMA-2120 and NS-TMA-2441.
These latter reports were not, however, factored into the FRC review.
With regard to submergence qual1fication on FRC TER Item Nos.
38, 40, and 56, concern arose over our application of a steam test at 75 psig (with no steam leakage 1-7
1
.into the tr~itter casing) as proof of sub~ence qualification at an effectiveWydrostatic head of 5 psig.
The~C staff did not accept this engineering
)udgment. It was therefore agreed that we would attempt to test the transmitters by March 31, 1985, or, if such testing proves to be impractioal, we would replaoe or relooate the transmitters by the same date.
k)
NAMCO Model No. EA180 limit switoh (FRC TER Item No. 48, Unit Nos, and 2).
Identified defioienoy:
steam exposure.
The identified deficiency appears to be based on the FRC observation that we did not identify the method of control cable termination at the limit switoh in the qualifioation of the limit switch.
The qualification test report indicated that the tested connections were protected from steam environment by sealing the chamber/threaded pipe interface with Teflon tape.
We informed the NRC staff that the limit switoh connections in plaoe at the Cook Plant are,
- indeed, proteoted from steam environment due'to the method of installation.
This information had been inoluded in the previously submitted qualification (SCEW) sheets for the control cable termination at the limit switoh.
Continental wire (FRC TER Item Nos.
52 and 54, Unit Nos.
1 and 2).
Identified deficiency:
test failure.
Due to difficulties encountered with the associated leads used to connect Continental wire samples through the chamber wall during testing, these samples had to be removed from the circuitry during testing.
Since the test setup was deficient, rather than the test samples involved, the NRC staff agreed with us that this deficiency could be disregarded.
m)
Electrical termination inside containment (FRC TER Item No. 81, Unit Nos.
1 and 2).
Identified deficiency:
radiation.
See II.c. above and Attachment 2 for a description of this deficiency and our proposed resolution.
n)
Boston Insulated wire (FRC TER Item No. 98, Unit Nos.
1 and 2; FRC TER Item No. 106, Unit No. 2).
Identified deficiency:
Profile enveloped.
As noted by the FRC review, the LOCA test profile for these samples did not envelope the FSAR temperature profile after the initial peak.
Although we informed NRC staff that the FSAR profile after the peak was below the normal operating rating of the cable, we were told that this approach was unacceptable since initial LOCA transient effects on cable rating could not be quantified.
At the suggestion of NRC staff, we have taken available margin from the short test and used Arrhenius analysis techniques to qualify the cable for long term response.
According to our calculations, the margin of test profile over accident profile amounts to thirty-four years.
Furthermore, although not specifically identified by FRC, certain issues relating to submergence qualification of Boston Insulated Wire Item f3075, Samuel Moore Cable Item 83075, and instrument cable splices (Raychem) have been raised.
These issues are discussed in paragraph (f) of letter No.
AEP:NRC:0775F
[M. P. Alexich (IMECo) to H. R. Denton (NRC), dated September 26, 1983].
Final resolution of submergence qualification for these items is 1-8
~ <
expected. to be compl~ea-or before March 31, 1985 [~rence letter dated October 24,
- 1983, H. K Denton (NRC) to John B. Dolan ~MECo)].
V.
In accordance with paragraph (l) of 10 CFR 50.49, replacement equipment will be upgraded unless sound reasons to the oontrary exist.
For the Donald C. Cook Nuclear Plant, conditions which reflect sound reasons why replaoement equipment need not be upgraded inolude the following:
Identical equipment is to be used as a replacement, and procurement aotivities regarding suoh replacement equipment had commenced prior to February 22, 1983.
Replaoement equipment qualified in accordance with the provisions of IEEE 323-1974 does not exist.
Replacement equipment qualified in accordanoe with the provisions of IEEE 323-1974 is not available to meet installation and operation schedules.
Equipment qualified to the DOR Guidelines or IEEE 323-1971 may be used for an interim period until upgraded equipment is obtained and an outage of sufficient duration is available for replacement.
Justification for use of replacement equipment not upgraded after this interim period has expired will be submitted to the NRC for review.
Replacement equipment qualified to IEEE 323-1974 would require significant plant modifications to accommodate its use.
Operating performance and reliability data for replacement equipment qualified to IEEE 323-1974 indicates poor overall equipment performance (e.g.,
mean time to failure is
. significantly shorter, for the replacement equipment).
The use of replacement equipment qualified to IEEE 323-1974 has a significant probability of creating human factor problems that may negatively affect plant safety and performance, for example:
(a) knowledge, skills, and ability of existing plant staff will require significant upgrading to operate or maintain the specific replacement equipment; (b) the use of the replacement equipment creates a one-of-a-kind application; or (c) maintenance, surveillance, and/or calibration activities are unnecessarily complex for the replacement equipment.
Furthermore, equipment components that are routinely replaced as part of normal equipment maintenance (e.g.,
- gaskets, O-rings, coils, etc.),
and components which are part of an equipment item qualified as an assembly (e.g.,
a resistor which is part of a transmitter),
may be replaced with components of identical design.
For those cases where an equipment item is replaced with
, upgraded equipment, the baseline environmental parameters to which the replacement equipment is qualified may be identical to the qualification parameters for the original equipment.
1-9
Table I EQUIPMENT QUALIFICATION DEFICIENCIES IDENTIFIED IN THE FRANKLIN RESEARCH CENTER TECHNICAL EVALUATION REPORTS DATED OCTOBER 28, 1982 Donald C. Cook Unit No (s).
FRC TER Item No.
Equipment Item Description Deficiencies 1 2 3
4 5 6 7A 7B 7C 7D 7E 8 9 10 11 12 13 14 15 16 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1/2 1,2 1,2 1,2 1,2 1,2 1/2 1,2 1,2 1,2 1/2 1,2 1,2 1,2 2
1,2 1/2 1
2 3
4 5
6 7
8 9
10ll 12 13 14 15 16 17 18 19 20 21 22 23 24 36 37 38 39 40 43 44 48 Limitorque MVA SMBOOO Limitorque MVA SMBOO Limitorque MVA SMB/1,00,2 Limitorque MVA SMB1 Limitorque MVA SMBOO Limitorque MVA SMB2 Limitorque MVA Limitorque MVA Limitorque MVA W 5009P24 W 5009H W 5808Z W TBDP Reliance Frame 05810P W
Conax EP2 through EP14 Conax EP1 Sostman 11901B Rosemount 176KS Sostman 11834B Rosemount 176KF Sostman 11834B Rosemount 176KF Mobilux EP2 Barton 763 Barton 763 Barton 764 Barton 764 Barton 764 Foxboro E13DHHSAH1 MCA Foxboro E13DMHSAH1 Namco EA180 x x x x x x x
x x x x x x x xxx xxx x x x x XXX XXX XXX x x x
x x
x x x x x x x x.x x x x x x
x x
x x
x x
x x
x x
x x
x x
x x
xxxxxxxxx x x x x x x x x x x XXX x x xxxxxxx xxx x x x
x x
x x
x x
x x
x x
x x
x
Table I (continued)
Donald C. Cook Unit No(s).
FRC TER Item No.
Equipment Item Description Deficiencies 1 2 3 4 5 6 7A 7B 7C 7D 7E 8 9 10 11 12 13 14 15 16 1,2 1,2 1,2 2
1,2 1,2 1,2 1,2 1,2 1,2 1/2 li2 1,2 1,2 2
1,2 1i2 1,2 1,2 1,2 li2 1,2 1,2 1,2 1I2 1,2 1I2 1,2 1,2 1,2 2
'1,2 1,2 1,2 1,2 51 52 54 55 56 57 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 78 79-81 82 83 84 85 86 89 90 95 96 Fisher 546 Continental Wire Continental Wire Cyprus Power Cable Barton 763 Asco HP/HT 8300 Series Asco HT8316 W 1101 Mercoid DA7031153 Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termi'nation Electrical Termination Electrical Termination Penetration Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Electrical Termination Anaconda Power Cable Anaconda Power Cable Okonite Power Cable Okonite Power Cable Cyprus Power Cable Continental Wire x x x
x x x
x x x
x x
x x x x x '
x x x x x
.x xxxx x
x xxxx xxxx x x x x
x x x XXX xxx x x x x
Sc x xxx
0
Table I (continued)
Donald C. Cook Unit No(s).
FRC TER Item No.
Equipment Item Description Deficiencies 1 2 3 4 5 6 7A 7B 7C 7D 7E 8 9 10 11 12 13 14 15 16 1,2 1,2 1,2 1,2 2
2 2
2 2
2 2
2 2
2 2
2 2
2 2
2 2
2 97 98 99 100 101 102 104 105 106
.107 108 110ill 112 115 116 117 118 119 120 121 122 43 55 68 86 101 102 Samuel Moore Cable Boston Insulated Wire Boston Insulated Wire Cerro Cable Rockbestos Cable Anaconda Power Cable Baiton 763 Continental Wire Boston Insulated Wire Barton 764 Barton 368 W 1101 Rosemount 176KF Barton 764 Barton 763 Barton 763 Anaconda Control Cable Continental Wire Raychem Cable Raychem Cable Cable Termination.
Sostman 11834B Foxboro E13DMHIMID Okonite Power Cable Electrical Termination Anaconda Power Cable Rockbestos Cable Anaconda Power Cable XXX x x x
x x x x x x x x x x x x x x x x x x x x
x x x
x x x x x x x x xxx XXX x x x xxx x
x x x x
xXXXX xxx XXX x x x x
x x
x
SUMMARY
OF D~EP Q~Eggg~DgggggggD 1N THEMBh55555 RXMMMMSX~SQKNML Zfkk3JhTXQH~JQETB DATEDM~RSB 2Lc&292 FOR >SJGX555X ~I9 XKShXSQQRXXB 5~3h M2~M~
DEFICIENCY
~SQQSX REMX3MMX NUMBER OF DEFICIENCIES 2HXX 3 25LX 2 MXHM~B*
1 Documentation 31 "36 37 2
Similarity 35 40 45 3
Aging Evaluation 52 4
Qualified Life 52 5
Aging Program 6
6 Aging Simulation ll 7A Peak Temperature 2
7B Peak Pressure 1
7C Duration 2
7D Prof ile Enveloped 3
7E Steam Exposure 2
8 Spray 9
Submer ge nce 10 Radiation ll Test Sequence 12 Test Failure 2
13 Functional Testing 6
14 Instrument Accuracy 12 15 Duration Margin 0
16 Margins 65 65 12 14 69 69 13 14.
Total Number of Deficiencies:
234 283 298
- NOTE:
Number of deficiencies identified for "Both Units" is adjusted to account for equipment items and Franklin Research Center reviews which are identical between the Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2.
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AMERICAN ELECTRIC POWER SERVICE CORPORATION ~ KQ OWER S YS~Stb December 6,
1983 Franklin Technical Evaluation Report Item 81 (SCEW 8TP-2, TP-3)
FROM>
L. F.
Caso TOt D.
Medek SCEW sheets TP-2 and TP-3 in our AEP-NRC-0578B submittal describe respectively the penetration extension wire splice to the field cable inside a floodup box and the field cable termination at the hydrogen recombiners and motor terminals.
Both these terminations were qualified to 150 mrads under Westinghouse Canada test report CWAPD-332, test samples D3 and D4 (TP-2) and E3, E4, F3, F4 (TP-3).
Both Test Items D3 and D4 did not remain energized throughout the entire test period.
Test Item D3 failed to remain energized after approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and Test Item D4 after approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of exposure to the test conditions.
Post-test examination of both test items revealed that the Kapton insulation on the lead wires was severely damaged at various locations.
The spliced connections of Test Item D3 successfully passed the post-test hi-pot tests in water.
- However, the Kapton insulation of Test Item D4 was damaged at a location very close to the splices and the post-test hi-pot test could not be performed on Test Item D4.
No portions of these test items were protected from the chemical spray by floodup tubes and there was no apparent damage to either the hypalon-jacketed cables or the associated field splices.
It should be noted that the exposed Kapton and its field splice as tested, is not representative of the present installations at the D.
C.
Cook plant for the long term instrumentation circuits installed in floodup tubes and all safety related power and control circuits.
For these circuits the Kapton wire is routed inside a stainless steel floodup tube from the penetration to the floodup box.
The whole assembly is constructed to avoid direct spray impingement on the splice or Kapton wire, and also to prevent water filling up the floodup tube (the floodup box is installed at a location above the expected flood level).
IMTRA-SYSTEM
Splice connection TP-2 is a
Raychem Corporation splice that has also been qualified by Raychem Corporation under test report F-C4033-3 (EQ central file Ref iI14).
The total radiation dose used during this test was of 200 mrads (see attached test profile).
There were no eventualities in connection with the TP3 test samples.
SCEW sheets TP-2 for both DC Cook units will be revised to reference both the Westinghouse Canada and the Raychem test reports.
Please let me know if you have any 'questi ns.
L. F.
Caso LFC/ris/2 APPROVED C. Carruth CC:
H.
N. Scherer, Jr.
S.
H. Horowitz R.
C. Carruth T.
E. King J.
A. Pria R.
Shoberg J.
G. Feinstein R.
F. Kroeger
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PA'fgi April 5, 1983
~4uyJKCfc Spurious Operation of Motor and Air Operated VaLves AO~R S'YS<~
FROMM B. Lee T. E. King r
The following is a description of AEP's "Double Break,"
philosophy and its impact on the 'question of s urious of motor and air operated valves.
The possibility of spurious operation of mot air operated valves due to h
philosophy of double breaking of the control circuit.
The do ue o s ort circuits is minimized b AEP's freak concept, as illustrated in encl d
h control contacts be located in both ol e
n enc osed schematic re n
o polari s of the ac t ng t
s configuration, a short circuit bet A to B or between points C t D
i between points of the valve.
A short circuit b t o
cannot cause a s urious
~
~
i e ween points A to D will p
operaton protective devices upstream to isol t result i iso a
e t e short.
This vill u
in motor operated valves remainin in 'the prior to the short c
t their de-energized position circus and air o crated r
A.spurious operation can only occur with a si short circuit.from poi t A t wx a s multaneous path maintained from points B to C.
oin s to B and C to D and a hi h i g
impedance located in It should be noted that points A
.B C
d n very close proximity of each other.
The selective an D are all short circuits without a circuit"fault is extremely unlikely when e
rea crxterzon is applied.
r To further illustrate the double break hiloso h
attached cable schematic 84851-12 cable run fo xndxcating the control circuit e run for pressuirzer power relief valve NRV-151.
cable 9705R-l, a
12 co d
The aforementioned points A
B C,
and D are indicated on a
12 conductor cable vith each conductor indivi-ual y insulated and surrounded by a common jacket.
A spurious opening of the pressurizer relief valve vould require a simultaneous short circuit between two (2'e pairs of conductors o
t s located xn the same cable jacket, vithout th wo
(
) seperate vo (2) pairs shorting to each other.
o e
1
)HTRA SYSTEM
Apri1 5, 1983 The likelihood of such discriminating short circuits occuring in a single multiconductor cable is remote at best.
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