ML17319A897
| ML17319A897 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/10/1980 |
| From: | Jablonski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17319A894 | List: |
| References | |
| IEB-79-01B, IEB-79-1B, NUDOCS 8106010437 | |
| Download: ML17319A897 (75) | |
Text
ENVIRONMENTAL QUALIFICATIONOF SAFETY-RELATED ELECTRICAL EQUIPMENT IEB 79-01B TECHNICAL EVALUATION REPORT DOCKET NO. 50-316 DATED:
November 10, 1980 Licensee:
Indiana and Michigan Electric Co.
Type Reactor:
V PWH P1ant:
D. C. Cook, Unit 2 Prepared.
by:
F. J. Jablonski Engineering Support Section Reactor Construction and Engineering Support Branch, RIII
k CONTENTS
~Pa e
Introduction
Background
and Discussion Summary of Licensee Actions/Statements System Comparison
~ 2
~
l Equipment Evaluation 2-4-8 Caveat Conclusion 2-.3 Attachments:
1.-
Referenced Test Reports 2.
Onsite Inspection Report 3a.
Generic Issues 3b.
Site Specific Issues 4.
Licensee System List 5.
NRR's System List 6.
Category Criteria 7.
LER's 8.
Unresolved Generic - Specific Issues 9.
Concurrence Code
Introduction This report is submitted in accordance with TI 2515/41-for use as input to 1/
the Safety Evaluation Report on qualification of* Class 1E electrical equipment installed in potentially "harsh" environmental areas at this facility.
Back round and Discussion 2/
IE Bulletin No. 79 required the licensee to perform a detailed review of the environmental qualification of Class lE equipment to ensure that the equipment would function under (i.e. during and following) postulated accident conditions.
The Technical Evaluation Report (TER) is based on IE's review of the li-
. censee's submittal for conformance. with the DOR guildelines or NUREG-0588, a'ite inspection of selected system components, to vegjfy accuracy of the submittal, and EQB's review of component test reports.
Licensee submittals were received on March 7, 1980 and October 31, 1980.
The site inspection was completed on June 17, 1980. Geyric and site 4/
specific guidance was requested from IE/NRR headquarters.
Summa of Licensee Actions/Statements Aging is a generic industry issue whose resolution is not clear at this time.
As relevant information becomes available, the licensee intends to continue his evaluation of the sensitivity to aging of materials and components.
Limit switches used only for indication of valve position will not be replaced.
The limit switches, along with their cable and their cable terminations, for the Ice Condensor Refrigerant supply (VCR's) Main Feedwater Regulating Valve, and Steam Generator Stop Valves, have been deleted from the submittal.
For the valves outside containment, valve position can be readily verified.
Those inside containment (VCR's) are containment isolation valves; a redundant con-tainment isolation valve located outside the containment, in series with the one inside the containment, will serve as a backup device.
The position of the valve outside containment can be verified.
Equipment required due to changes in emergency operating procedures after January 14, 1980 (the issue date of IE Bulleitn 79-01B) is not always included in this submittal.
The cut-off date was agreed to at the February 7, 1980 clarification meeting in Glen Ellyn, Illinois.
1/
Technical Evaluation Report (TER) On Results Of Staff Actions Taken To Verify Reactor Licensee
Response
To IEB 79-01B And Supplemental Information.
2/
Environmental Qualification of Class 1E Equipment.
3/
Attachment l.
4/.
5/
Attachements 3a and 3b.
~.
~
In the equipment qualification charts, the specified radiation dose is the calculated bounding dose from gamma radiation as per the licensing basis of the Cook Plant.
(Not the guidelines)
S stem Com arison A comparison was made between the system~ list provided by the licensee-'/
and a similar list provided to IE by NRR-during a meeting in Bethesda, MD
/
on September 30,- 1980.
The following systems were not included in the li-censee's submittal.
Auxiliary Steam Isolation
'Auxiliary'eedw'ater Isolation
'ontainment Air Purification/Cleanup Accumulator Pressurizer Spray Power Operated Relief. Valves Steam Dump Containment Radiation/Sampling Containment Sump Safety Equipment Area Ventilation g
1 I
R E ui ment Evaluation Class lE eqgjpment was evaluated, that is, placed into five separate categories. Result of the evaluation follows:
(See pages following)
Caveat Test reports and other documentation which licensees referenced as estab-lishing environmental qualification were reviewed for acceptability by NRR, Environmental qualification Branch.
(Reference a, memorandum dated June 20, 1980 Hayes to Jordan.)
This TER does not include information about seismic or. fire withstand capability. It should therefore not be inferred that Category I equipment meets all necessary qualification requirements.
Conclusion Based on IE's review of the licensee's submittal, the site inspection, and licensee's proposed actions, it cannot be concluded that there is reasonable 6/.
7/.
8/.
assurance all components installed at the D. C. Cook Nuclear Plant, Unit 2 are environmentally qualified and *installation methods of environmentally qualified components would not contribute to the failure of such components during a potential accident.
A positive conclusion cannot be made until:
1.
All matters referred to IEH(}S/NRR have been satisfied.9/
2.
The 10 systems missing from the licensee's s'ubmittal have been evaluated by NRR.
(Page 2)
. 'The negative'equ'ipment evaluations have been re'viewed'by MR.
(Pages 4, 5, 6, and 8.)
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toclaving Power Cable Cable est Hezorts AT~pACHPZI'lT 4 IPS-234 Penetration Blank IPS-137 Penetration Blank F-C 3341 Anaconda Cable E-C.3694.
Okonite. Power Cable:
Kerite Report on Effects of Gamma Rad and Au Conax IPS-348 BIW Cable F"C4033 1'er minations F-C3683 Sam Noore Cable
'somedix 5/76 Sam Noore Cable Cerro Cable 5/76 CWAPD-332 Terminations F-C 4033-3 Term Kapton to XLP Blank MOY 600456 IPS-326 IPS-327 IPS"329 Term WCAP 7709-L Supp 2 H
Recombiner WCAP 7829 Fan Motors TBDP 2
NOV-600198 MOV 600376A Limitorque 600461 Isomedix Corp. T.R. 11/7S Essex Power Cable Foxboro T R. No. TE-1013 Letter W to Case 4/26/78 No. NS-PLC-5023 WCAP 9157 RTOs ASCO AQS 21678 Barton XNTRS 764 W Letter NS-TNA-1950 BIW 73C212 Cable BIW 75C008 Cable F"C 2935 Continental Cable Rockbestos Firewall III FC3016 Cyprus Power Cable Blank Qua.l. of Okoguard EPR Insul. (Okonite) Inst Cyprus Statement of 6/16/76 Cyprus Statement of 9/14/76 Cyprus Rep'ort 03525 Cyprus Report 03658 Cyprus Statement of 6/21/76 NANCO Limit SNS 9/5/78 F"C 3271 Control Cable MOVs IPS-339 Conax Cab le Term IPS-349 Conax Cable Term Grease Mobilux EP"2 Letter Grease Nobi lux 'Letter Okonite Form N-1 Power Cable
~o egg PG 4
~
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UNITED STATES NUCLF<R REGULATORY COMMISSION REGION III 769 ROOSEVELT ROAD GLEN ELLYN,ILLINOIS60137 July 1, 1980 MEMORANDUM FOR:
. E. L. Jordan, Assistant Director, Division of Reactor Operations Inspection, IE:HQ G. Fiorelli, Chief, Reactor Construction and Engl'neerl ngupport Branch
'ROM:
SUBJECT:
D. W..Hayes, Chief, Engineering. Support Section. l.
SCREENING REVIEW OF LICENSEE
RESPONSE
TO IEB 79-01B AND
SUMMARY
OF INSPECTION OF INSTALLED SYSTEMS'AT D.C.
COOK UNITS 1
AND 2 " DOCKET NOS. 50-315; 50-316 Frank Jablonski has completed his initial review of the D.C.
Cook facility response to IEB 79-01B, and the inspection phase of the system audit.
There were some basic di fferences between Units 1
and 2, e.g.,
main steam flow and narrow range RCS temperature transmitters were not part of Unit 2's engineered safeguards actuation; Unit 1 did not have a "dedicated" post accident monitoring system.
A telephone conversation with the licensee indicated the matters were considered by NRR during 1 icensing of Uni t 2.
The licensee has taken exception to aging,'eplacement of certain limit switches (indication only), and to protec)ion of certain components.
from adverse environments including flood Test reports referenced by the 1 icensee are Included in Attachment 2.
The reports were not evaluated as part of this review.
Review of all test reports and other documentation referenced by the 1 jcensee shall be accomplished by NRR (Environmental Qualification Branch)
A walkdown was conducted on June 16 and 17, 1980 to inspect installed components associated with Unit 1 Engineered Safety Features actuation and Contain@ant Isolation, and Unit 2 Containment Spray Systems.
Prior to the walkdown, wiring diagrams were reviewed.
Components included on Attachment 3 were observed.
1 See Attachment 1
(Excerpts from Mena Hayes to Jordon A/I F03067180 dated June 20, 1980)
Onside Inspectio"..
ATTACEKIIT 2 I
~
~
P
E. L. Jordan 2
July I, 1980 Unit 1 Observations:
Isolation Valves Four solenoid valves were observed; all were ASCO catalog number 2063812RVU or NP83l654V.
(The control cables and penetrations for the solenoids inside containment were not located above flood level nor were cables..installed in flood-up tubes)..
Limit switches, associated with valves controlled by the solenoid valves, were'HAMCO 'numbe'r "'.'
EA740"6000-0.
)
'I Two 1imitorque motor operated valves'ere observed.
One was type SMB-00" with a Reliance motor, insulation Class RH, the other was type SMB-2 with a Reliance
- motor, no insulation class stamp.
Both valves were located below flood level.
One valve had its cable installed in flood" up tubes, the other did not.
Instruments Four transmitters were observed, including flow and= pressure.-=-Temperature.
measuring elements could not be observed because of high radiation.. With the exception of the containment pressure transmitter (outside containment) all Barton and Foxboro instruments included a hermetic seal.
Only the cables for the Pressurizer Pressure (long term monitoring) transmitter were installed in flood-up tubes.
The main steam pressure transmitters
{outside containment) were protected by missile shields.
Terminations (See Attachment 4)
All terminations at instruments inside containment were covered with Raychem WCSF heat shrink tubing.
Solenoid terminations for valves VCR-103 and VCR-104 were covered by electrical tape in a junction box with gasketed cover.
The terminations had just previously been made and were supposedly in a manner qualified in ASCO test report AQS 21978/TR.
Terminations at penetrations were various.
Those penetrations which had cables in flood<<up tubes were terminated in large terminal boxes at elevations above the max'imum flood level.
Termiaating points were covered by Raychem WCSF heat shrink tubing.
For penetrations without cable in flood-up tubes,
{non safety-related) control cable terminations were made to,terminal blocks in large terminal boxes at elevations below maximum flood level.
Instrument cable was terminated to the penetration pigtails by multi pin connectors or directly A'TACK'~iZ2
E.
L. Jordan July 1, 1980 with butt splices covered by Raychem WCSF tubing.
Power, cable (both safety and non safety-related) was terminated directlv to the.pene-tration pigtails with the termination point covered by Raychem heat shrink tubing.
Power, control, and instrument penetration pigtail insulation was spirally wound Kapton, and in many cases, was exposed directly to the
- potential.,containment.
accident envir'onment of radia'tion,,
chemical spray; and flood.
The concern is, even though a component may not be safety-related or required to operate. subsequent to flooding, could short circuits cause gross fai'lure'f 'a penetrati'on and breach.
co'ntainment'ntegrityT (Failure of Kapton insulation'is discussed in licensee reference No.
13; refer to IGH letter to Case dated April 21,
- 1978, last page).
Cables (See Attachment 4)
(Refer to discussion on terminations).
The only cable which had visible markings was BiW's Bostrad 7.
- Cables, like terminations, were installed In various ways including below flood level.
Since none of the cable was tested for long term sub-
- mergence, the same concern exists for cable as does for terminations, i.e., short circuit effects on penetrations.
Host of the power and control cable had an asbestos braid jacket.
NOTE:
Even though beyond the scope of IEB 79-01B, it is unclear whether in Unit 1 justification has been made for use of cables which did not pass the
!PCEA"19-81 vertical flame r'esistance test.
Reference:
November 25, 1977 NRR letter to licensee, subject:
"Summary of Heetings Held on November 3 and 4, 1977, Reference-Fire Protection" (Question
- 16).
Penetrations (See Attachment 4)
Penetrations were located below flood level.
Penetrations were manufactured by Conax and consisted of power, control, and instrument types.
Data sheets indicated that test o eratin times varied be>ween 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for instrument control t e
and s for ower t e.
Penetrations are s ecif'ed to rerrain operable, l.e., integrit
, for up to one year.
As reported
- above, penetration feed-throughs (pigtails) were terminated to plant cables in ~ three distinct ways:
multi-pin connectors; butt spliced directly and covered; through flood-up tubes via a separate terminal cabinet.
Except for those pigtails in flood-up tubes, the Kapton insulated pene-tration feed-through conductors (pigtails) were exposed directly to potential harsh environments which could cause short circuits.
AT'ACHIKHT 2
E. L. Jordan July 1,
1980 Fan Motor A Westlnahouse motor was coupled to fan unit HV-CEQ-2.
The following pertinent nameplate data was o~~alnt=u:
Serial No. 7206, HP-75, Insulation F,
Frame 449TZ, Type TBDP, Style 71C20763-04, Bearings 4109D41G05, Grease 53701RW - Style 773A773G05.
Page Fl-1 of the licensee's June 5,
1980 submittal references Westinghouse Corp. test. report WCAP-7829 as the. qualification. document..
A cursory...
review of the repor't indicates it does"not identify the specific motor type which was tested;
- however, the report speci'fically states that Ther'malastlc.Epoxy'nsulation'is' "key -component" in the motor, yet. the specifies the o eratln time to be one ear however 'he " uglification" co umn o blank.
The WCAP-29 test was for "short term'hich was not defined althou h inferred to be less than one year.
Unit 2 Components for the containment spray system were all located outside containment.
Redundant components were located
.i.n separate rooms.
A list of equipment observed and model numbers is included on Attachment 3.
No adverse environments were apparent.
Conclusion Except as noted above, cables - fan motor type - exceptions to aging-replacement of limit switches - and protection from adverse environment, equipment desciptions provided by the licensee on the system component evaluation work sheets were complete and accurate.
As stated previously, all test reports listed by the licensee will be evaluated by the Environmental Qualification Branch of NRR.
D.
W. Hayes, Chief Engineering Support Section 1
Attachments:
1.
Excerpts from Memo Hayes to Jordan (A/I F03067180) dtd 6/20/80 2.
List of iest Reports 3.
Components Inspected 4.
Figures lA and 1B cc w/attachments:
J.
G.
Kepp I e r G. Fiore 1 I i R. Heishman R. Masse, Res..Insp.
V. Thomas, IE:HQ
ATTACHMENT I EXCERPTS FROM MEMO HAYES TO JORDON (A/I F03067180}
DATED JUNE 20, 1980 item 3, Attachment I to Memo
.;Reference,No..
3,'. even th'ough a'corn'ponent may. not'be required to ope'rate.
'subsequent to flooding, what effect wl.ll short cl rcults have on containment electrical penetrationsT Was this considered by NRR7 items 2, 4, 5, Attachment 21 to Memo 2.
Licensees maintain that aging is a generic industry Issue whose resolution is not clear; therefore, evaluation has not been made or wi I I continue to be made-as relevant Information Is made available.
4.
Lim1t switches used for valve position indication only have been deleted from the submittal.
Licensees maintain that a valve outside contai nment ln series with one inside can have its position verified visually following an accident.
Is this acceptable?
The licensees maint'aln that neither valve position limit switches, solenoid valves, nor control cables for air operated'ontainment 1solation valves need be replaced or protected from the adverse environment, including flood, because all postulated fai,lures.wi.ll result in the Isolation valve assuming Its fall-saFe pos ltlon.
Is this acceptable?
AF"ACHPZii'Z 2:
ATTACHE'lENT 2 LIST OF TEST REPORTS 20 3
4 ~
5 6.,
70 8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22.
23.
24.
25.
26.
27.
28.
29.
30.
31.
32.
33.
34.
35.
36.
37 38.
39.
40.
41.
42.
43.
44
'5.
46.
47.
- 48.
49.
IPS-234 Penetration Blank IPS-137 Penetration BLank F-C 3341 Anaconda Cabt.e F-C,3694.-. Okonite Power. Cable Kerite Report on Effects of Gamma Rad and Autoclaving Power Conax IPS-.348 BIW Cable
'F-'.C4033"1
'Te rminati ons F-C3683 Sam Moore Cable Isomedix 5/76 Sam Noore Cable Ce rro Cab Le 5/76 CWAP D-332 Te rminat ions F-C 4033-3 Term Kapton to XLP Blank NOV 600456 IPS-326 IPS-327 (527)
IPS-329 Term
'WCAP 7709-t Supp 2 H
Recombiner WCAP 7829 Fan Motors TBDP 2
NOV 600198 NOV 600376A Limitorque 600461 Isomedix Corp. T.R. 11/75 Essex Power Cable Foxboro T.R. No. TE-1013 Letter W to Case 4/26/78 No. NS-PLC-5023 WCAP 9157 RTDs ASCO AQS 21678 Barton XNTRS 764 W Letter NS-TNA-1950 BIW 73C212 Cable BIW 75C008 Cable F-C 2935 Continental Cable Rockbestos Firewall III FC3016 Cyprus Power Cable BLank QuaL of Okoguard EPR !nsul. (Okonite) Inst. Cable Cyprus Statement of 6/16/76 Cyprus Statement of 9/14/76 Cyprus R'eport 03525 Cyprus Report 03658 Cyprus Statement of 6/21/76 NAYiCO Limit SNS 9/5/78 F"C 3271 Control Cable NOVs IPS-339 Conax Cable Term IPS-349 Conax Cable Term Grease Nobi Lux EP-2 Letter Grease Nobilux Letter..-
Okonite Form N-1 Power Cable Cable
ATTACHMENT 3 COHPONENTS INSPECTED DE CD COOK UNITS 1 AND 2 CY'CTC'Q ESS1 ESS1 ESS ESS1 ESf" CI2 CI2 CI CI2 2
1 1
.1 1
~
~
'1 1
1 1
1 2
2 2
2 2
1 1
1 1
PLANT ID NO.
NFC-11010 NPP-220 NTP-111 NPP-151 PPP-300710 OCM"2508
.ICN-1115
'FRV"'210 VCR".101 6 VCR-10366 VCR"1(T)~
PP-009 12 INO-215 INO-210 INO-22215 ILS-950 Various Various Var ious HV-CEQ-2 r~M~+~C NANE Transmitter Transmitter Transmitter Transmitter
'Transmitter Notorized Valve
.Notori.zed Valve, Solenoi'd Valve Solenoid Valve Solenoid Valve Solenoid Valve Pump/Motor Notorized Valve Motorized Valve Notorized Valve Transmitter Penetration Cable Terminations Fan Notor
. IN OUT X
X X
X X
X X
X X
X X
X X
X X
X 1 Engineered Safety Features 2 Cont ai nment Iso I.at ion 3 Containment Spray 4 Air-reci rculati on 5 Catalog No.
2063812RVU 6 Catalog No.
NP831654V
'7 SI'IB, Class RH~ no flood-up tubes 8
k SNB-2~ Class
~ flood-up tubes 9 ITT Barton 764 10 Foxboro E11GN 11 ITT Barton 763 12Limitorque~ SNB"0. Rel.iance Motor~ Ins.
Class HR 13Limitorque~ SMB-00~ Ins.
Class B
14 Reliance Type P~ Ins.
Class B
15 Foxboro E13DM without hermetic seal
l NQ ItK.
FLOOD LE.V;L-Ir g IJ /= p. r S r R4Y&lgh( MC'rP 0Jul t' POW E.R IN 57.
R4VCIIEJ4 WCSF~
7Q FIELD ggeiR~
~ //). I:J'7 LC Cc'-
'C tlat P'LlLTI Pl/l ca.i:iwTC32
~
/
COPJ VFO EXP'/"/ iC//PTO'/ //./SULQ7fb../'/l6 TI-'Lz~
SPwc~F FIG.LA.. SuOZT Teem Wvea<e WCSF Al07K 'u rt h e r /// forma 9/cn iega~djng thea maf$ ev is egg )ggnFd lrf SE znspzc)yii So-~(I) ~S-oq j/
I ~
IQS, I~E.
7LQOD LEVEL~~
om.tet POWE. R WWW
~~+>+
~ 'l/WlAv% ~ %IV~~v~~
4'i>>i;i ~i%%A~
CON T$;9 1
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NUCLEAR REGULATQlisY COiViViiSSION RiE GIOV III 799 ROOSE VELT ROAD GLEN ELLYN.ILLINOIS60131 aa
~ 0
<<s July 23, 1980
~
~
MEMORANDUM FOR:
E. L. Jordan, Assistant Director, Division of Reactor Operations Inspection,,
IE:HQ THRU:
G. Fiorelli, Chief, Reactor Construction and Engineering Support Branch FROM':: ' '. V. Hayes, Chief, Engineering Support Section 2
SUBJECT:
. iEB 79"0)B':(A/I'03067180')
e
~,
5 ~
v Attached is a copy of a memorandum dated July 17, 1980 received from Frank Jabionskl relative to IEB 79-01B.
It is being forwarded for your information and solicited guidance.
The question of identi f1 cation of safety related, systems and components (paragraph No.
1 of the menu) ls an old one.
I disagree with Frank ln that I feel that this identification ls a responsibi1ity of the
- licensee, not the NRC.
He must know his plant.
I do agree,
- however, that more guidance ls needed for our inspectors in this area.
This is especially important'or those Inspectors that have not had reactor operating experience.
i s
The significant di fferences in master lists that Frank discusses-, 'in paragraph two does raise questions.
We can only compare these lihts against the SAR.
Review and evaiuation beyond this is assumed tq ba an NRR function.
In regard to Frank's question - should we assume the iicensee's response to IEB 79-01B to be complete an'd correct -
I have told him yes.
- Further, that if he identifies signlflcant incompleteness in the response, or Incorrect information during his reviews, to bring these to my attention so appropriate action can be recomnended.
COImnents and further guidance Is requested concerning matters discussed in paragraphs 3 and 4 of Frank's memo.
D.
W. Hayes, Chief Engineering Support Section 2
Generic Issues ATTACHI';ENT 5a
E. L. Jordan July 23, 1980
Attachment:
F.
J.
J ab 1 on s k i Memo to D.W. Hayes dtd 7/lj/80 CC J.
V.
A.
R.
D.
J.
R.
0 F.
w/attachment:
G. Keppler, Rl I I
D. Thomas, IE:HQ Finke 1, R I Ha rdw i ck, R I I
- McDonald, RIV
- Elin, RV F. Heishman, RIII J.
Jab lonski, RI I I ATTACHMENT 3a
~p,it Rggt
+~
~~o
++*++
UNITED STATES NUCLEAR REGULATORY COMMJSSION REGION III 799 ROOSEVELT ROAD GLEN ELLYN,ILLINOIS60137 July )7, 1980
~
~
~
MEMORANDUM FOR:
D..W. Hayes, Chi ef, Eng inee" ing Support Sect ion 1
s FROM:
F. J. Jablonski, Reactor Inspector
~'r
's ~ '
s
~
~
~.
~,
..., s'-
SUBJECT:
. " FORMULATING TECHNI CAL EVALUATION REPORTS
. (TER)
REV I EW OF I EB 79-01B RE:-
MEMO TO.YOU'DATED 'JUNE.16'; 's1980 - SAME SUBJECT Since the'eview of IEB 79-018 ls continual, new discrepancies continue to show up; discrepancies are not necessarily the licensees'.
As you know, 'there is no specific nuclear power plant design required by NRC.
Further, the designation of safety related systems is somewhat arbi trary and inconsistent.
In fact, the NRC places responsib i 1 i ty for classi fying safety related systems on the 1 icensee.
Action Item No.
1 of 79-018 requested each 1 tcensee to provide a "master.
list" of all ESF systems in their respective plant required to function during a postulated accident."
Appendix A to 79-01B lists "typical" equipment/functions needed for mitigation of an accident.
A comparison of master lists was made of four licensees with similar Westinghouse-PWRs (see Attachment 1).
Arbitrary selection and non-standard nomenclature of systems makes evaluation of the master lists extremely difficult.
NRC requested each licensee to submit the information under oath.
Should the information therefore be assumed complete and correct2 It is extreme'Iy frustrating to review" responses which vary so much in attention to detail, depth of review, etc.
As stated previously in the draft TER for D.C. Cook, because I as a principal reviewer lack detailed systems/operations experience, further gui dance is requested.
I Another TER related matter is motorized valves equipped with Limitorque operators (see Attachment 2).
As can be seen, each test report is for a
~s ecific unit type including motor type end insu'ietion ciess.
Almost all licensees refer to the various test reports as qualification
.documentation for all series of=operator types; never is name plate data provided.
For example, test report No.
600II56 (SMB-O-IIO, Rel lance Motor with Class RH insulation) may be listed for all operators from series SMB-000 to SMB-5; motor name plate data not provided.
Without the name plate data and the basis for extrapolation, a meaningful evaluation cannot b'e made.
I r
e ATTACHMENT 3a
~
~
~
~
~
~
D.W, Hayes 2"
July lj, 1980 It ls requested that this memorandum be forwarded to IE:HQS as an addition to A/I F03067180, with the same copy distribution.
3-,'
"'"F. J. Jablonskl Reactor Inspector Attachments:
1.
Comparison of Haster Lists 2.
Hotor Operated Valve Tests CC:
J.
G. Keppler G. Fiorel 1 I 0 ~
ATTACHNENT 3a
ATTACHMENT 1
~SYSTEM Aux.
F.W.
Chem.
8 Vol. Cont.
Cntmt. Air Hndlg.
Cntmt. H2'ont.
Cntmt.
Sp.
. Main Stm.
Aiix'. Stm.
Stm.
Dump
,Rx. Clnt.
~
~
Res. Ht. )m.
Saf. 'Inj.
CLg. Water Esnt'L. Serv.
Wat.
Comp.
C lg. Wat. 2 Emerg.
Corp CLg.
Aux. Clnt.
Cntmt. Purge Rx. BLdg. Vent Inst.
5 Prot.
Rx. Trip. Act.
Rx. Cont.
8 Prot.
Rad. Nonit.
Rx. Hot Samp.
Stn.
8 Inst. Air Stm.
Gen.BD Post Acc. Nonit.
Rem. Sht.
dn. Nonit.
Cntmt. Isol.
Nn. Stm. Isol.
Nn.
FW Isol.
P.I.
X X
X X
X
.X, X
X
~
X X
X' QA X
X X
X X'2 X
X X
X,...
X 2
2 X
X X
X X
X X
X KEW.
~ X X
X PT.
BCH X
X X
1
~ '; X'"
X:'
X
~t ATTACHMENT 3a
ATTACHYiENT 2 NOTOR OPERATED VAlVES NOV's 1.
There are basically two type series of limitorque operators:
SYB and SB.
The operators are sized from 000 (smallest) -to 5 (Largest) as foLLows:
SYB-000
. SNB.-OO
~
1 SNB/SB-0 SYB/SB-1 Sm/SB-2 SNB/SB"3 SMB/SB-4 SNB-5 This series may also
.include SB r ~
This'eries
'may also include WB
'his series may be suffixed "T" 2.
Test Reports include:
Report No.
a.
600198 Date 1-2"69 Motor Type Insulation-SNB-0-15*
~
Re Liance No Radiation Speci a l Hi Temp Unit Type Environment
- b. 600426 4-30-76 SNBW-25*
(B-0009)
- c. 600376A 5-15"76 SMB-0-25*
FIRl F-C 3441 BWR 1x10 R
340 BWR 2x10 Peerless DC Reliance H
RH d.
600456 12-9-75 SNB"0-40*
PWRB 2x10 Re Liance RH e.
600461 6.-7-76 SMB"0-25*
Outside Cntmt7 2x10 Re Liance
- f. WCAP7410L 12"70 7744 8-71 SNB&0
- denotes foot pounds of torque only SNB-0 has been tested seismically Re: a, b, c
ATTACHNENT 3a
tg gS R "KC) 4 O
~~i V
+0
+0 UNlTED STATES NUCLEAR REGULATORY COMMlSS)ON WASHINGTON, D. C. 20555 JUL.3 1980 SSINS j/6820
.HBlORANOUM FOR:
Z. R. Rosztoczy, Branch Chief, Equipment gualification Branch, Division of Engineering,
'NRR THRU:
@~+
E. L. Jordan, Assistant Director. for Technical Programs,.
~
, Division of. Reactor Operations'nspection','ZE FROM:
D. Thomas, Task Manager, Review Group, IEB 79-.01B,.
~ Division of Reactor Operations '.Inspection; IE
SUBJECT:
REQUEST'OR NRC POSITIONS ON REVIEW QUESTIONS OF IEB-79-01B LIt;ENSEE RESPONSES In accordance to our verbal agreement, we would be happy if you would provide positions on the questions noted in'the enclosed memoranda.
Since it is essential to establish a uniform approach to the review effort to obviate the questions being generated in the on-going review of licensee responses,.we will be happy to meet with your staff to.discuss 'these concerns to expedite resolution of the issues.
'nclosures:
l.
Nemo D.
dated 2.
Yemo F.
dated 3.
Nemo F.
DATED
~
~
Yincent D. Thomas, Task Manager Review Group, IEB 79-01B M. Hayes to G. Fiorelli, RIII June 20, 1980.
Jablonski to D. Hayes, RIII Jun 16, 1980.
Jablonski to D. Hayes, RIII June 10, 1980.
CC:
w/enclosures E. L. Jordan, IE Y. S.
- Noonan, NRR G.'iorelli, 'RIII D.
M. Hayes, RIII A. Finkel, RI R. Hardwick, RII E. Jablonski, RIII D. NcDonald, RIV J. Elin, RV
'UL 71980 ATTACHMENT 3a
~nn ~a~~,
~c+
4**y4 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 799 ROOSEVELT ROAD GLEN ELLYN.ILLINOIS60137 June 20, 1980 MEMORANDUM FOR:
E. L. Jordan, Assistant Director, Division of Reactor Operations Inspection, IE:HQ
~
THRU:
~. -Fiorell1, Chief, Reactor Construction and Engineering Support Branch FROM:
SUBJECT:
D.
W. Hayes, Chief, Engineering Support Section I
I EB 79-01B (A/I F03067180)
Attached are two memorandums from one of my inspectors, Frank Jablonskl.
The first is dated June 10, 1980. and the second June 16, 1980.
Both memos raise basic questions for which we require.guidance to complete our review of-responses to IEB 79-01B.
By this memo I also would like to confirm our understanding that NRR (Environmental Qualification Branch) will review. for acceptabll ity ail test reports and other documentation which licensees reference as establ i shing environmental qual Ification of instrument/electrical equipment.
In connection with this, we are sending under separate cover test reports, etc.
In our possession to be forwarded to the Environmental Qualification Branch..
(We further understand that the IEB 79-OIB task group; on a volunteer basis, may agree to review some of these documents).
The status or schedule for site inspections and review/evaluation of the final reports is also attached.
Please note that every licensee has asked for some sort of time extension to submit their first report.
We understand that the other regions have had s Imllar reporting problems.
Assuming 'that-:all our 'licensees meet their extended submittal dates, we should complete our site inspections,
- reviews, and technical evaluation ATTACHMENT 3a
E.
L. Jordan June 20, 1980 reports by the end of December 1980.
Further delays in the submittals or any unforeseen events wil,l hamper our ability to meet the new February 1,
1981 deadline.
D.
W. Haye, Chief Engineering Support Section 1
Attachments:
1.
Memo F. Jablonski to D. Hayes 6/10/80 2.
Memo F. Jablonskl to D. Hayes 6/16/80 3.
Inspection Sta'tus/Schedule 4.
"Separate Cover" List (Test Reports Sent to IE:HQ)
- Separate'over:
See Attachment 4
cc w/attachments 1,
3, 6
4 only:
J.
G.
K ppl er G.
Fi ore 1 1 i V. 'D. Thomas, IE:HQ A. Finke I, Rl, R. Hardwick, Rl I D. McDonald, Rl V J.
El in, RV R.
F. 'Heishman ATTACHMENT 3a
UNITED STATES NUCLEAR REGULATORY COMMISSION RE G ION I I I 799 ROOSEVELT ROAD GLEN ELLYN,ILLINOIS60137 June 10, 1980 NENORANDUN FOR:
D.
W. Hayes, Chief~ Engineering Support Section 1
FROM:
F. J. Jablonski, Reactor Inspector
SUBJECT:
EFFECT OF PREVIOUS NRR'EVIEW ON'A'TTERS'RELATING
'O IEB 79"01B In almost every I.icensee response to IEB 79-01B there is a subtle or direct reference to matters apparently reviewed by NRR.
'Because ot the referenced dates it is assumed by me that NRR has given either tacit or direct approval to the refer'ences; examples foLLow:
1.
ALL licensees refer to their FSARs for establishing the List of engineered:safety feature systems and environmental data such as temperature~
pressure~
radiation~ etc.
2.
One Licensee~
Wisconsin Public Service Corporation~ states that "The AEC, in their "Safety Evaluation of the Kewaunee Plant", Section 7.5, issued July 24~ 1972, conct.uded that our criteria and testing program for environmentaL qualification were adequate".
It js further stated that "Our FSAR~ which was approved by the AEC~ discusses at Length the post accident conditions and required qualifi-cations for appl.icable equipment.
(See Section 7.5 of the Kewaunee FSAR.)"
3.
Two licensees, American El.ectric Power and Wisconsin Publ.ic Service Corporation~
have discussed the effect of-components beLow flood LeveL simply by referencing Letters previously submitted to the NRC~ or FSAR questions/answers as foll.ows:
AEP Letter dated 9-29-75 from Tillinghast (AEP) to
, Kniel (NRC);
FSAR question 40.10 Appendix Q.
WPSC Letter dated 2-2"76 from James (WPSC) to Purp'Le (NRC).
ATTACHNENT 3a
D.
W. Hayes June 10, 1980 My specific concerns are:
Is it.to be assumed that the referenced FSAR parame'ters~'o.
1 above~
are correct~ i.e.
reviewed by NRR?
If the'answer is yes~ 'then should't also be assumed that No. 2 above is Likewise adequate?
(If the answer is no~ then none of the Licensee responses which reference the FSAR can be assumed to be correct.)
Reference No.
3~ even though a component may not be required to operate subsequent to flooding~ what effect will short circuits have on containment eLectricaL penetrations?
Was this considered by NRR?
I am requesting that these questions/concerns be forwarded to the Assistant Director~ Division of Reactor Operations'nspection for resoluti on.
F. J. Jablonski
'eactor Inspector
~ N CC J.
G. Keppler G. FioreLLi
~
~
ATTACHMENT 3a
I g
~
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 799 ROOSEVELT ROAO GLEN ELLYN.ILLINOIS60137 June 16, 1980 HENORANDUH FOR:
D.
W. Hayes, Chief, Engineering Support Section 1
FROH:
SUBJECT:
F. J. Jablonski, Reactor Inspector
. FORMULATING TECHNI CAL EVALUATION.REPORTS
.(TER).-
REVIEW OF IEB 79"OIB In accordance with IEB 79-01B, an overall conclusion relative to the qualification of instrument electrical equipment is to be made for each operating plant based on a screening review of all plant systems, and by a detailed review and observation of specific system components.
Unresolved concerns previously 'identified by RII I inspectors during reviews of IEC 78-08 and IEB 79-'1 along with subsequently identified concerns make It difficult for us to formulate meaningful TERs for certain plants.
The previous unresolved concerns are documented In the memorandums listed below (1,2,3) and are reiterated In Attachment A to this memo.
Subsequently identified concerns are listed in Attachments B, C, and D.
To assure uniform evaluation, guidance is needed for these Items.
Please forward these concerns to IE:HQ.
1.
TI'515/13 - Qualification of Safety Related Electrical Equipment Fiorelll to Sniezek, 10/13/78 2.
Same title as 1., Fiorell I to Klinger, 12/78 3.
Review Status of Responses to IEB 79-01, Hayes to Jordan, 9/5/79 9 g (- KkM;LS(~.
F. J. Jablonski Reactor Inspector
Enclosures:
As Stated CC:
J.
G. Keppler G. Fiorelli V. D. Thomas, IE:HQ A. Finkel, RI R. Hardwick,. RII D. McDonal'd,. RIV J. Elin, RV ATTACHNENT 3a
ATTACHMENT A 1.
Foxboro Models E11GM and 611/613 transmitters with MCA modification are believed by RIII, to be under a generic review by HRR.
It I
II's furthe'r. belie'f that the "MCA" modification does not make the transmitters suitable for use ln a radiation environment.
Is Region.II.I's unders.tandlng correct7.'.'.
Several licensees have declined replacement of limit switches which provide position indication of valves used for primary containment isolation.
Are"these switches required to be qualified2 GE cable type Si-57275 is used on penetratlons manufactured by GE.
Penetrations with this. type cable are instal.led, at Monticello Dresden 1 and 2, Quad Cities 1 and 2, and Duane Arnold.
The cables w thstood LOCA tests performed by Wyle Laboratories, Report No.
44114-2;
- however, the cable did not pass the IPCEA S-19-81 vertical flame test.
- Further, In the same Wyle test,6GE cable Si-58136 failed at radiation levels In excess of 5xlO rads.
We recognize that in regard to GE cable type Sl-57275 flame tests are not part of the environmental qualifications addressed in IEB 79-01B, but it makes no sense to find these penetrations acceptable per IEB 79-01B knowing that they may not meet other requirements.
Concerning GE type Sl-58136 cable, this item should be evaluated on a generic basis since many of the early GE plants use this cable.
4.
One licensee, American Electric Power, lists a letter Ho. NS-TMA-1950, W to HRR, as technical reference for qualification of ITT.Barton
- differential pressure transmitters.
Please supply us with the disposition and status of the letter.
ATTACHMENT 3a
ATTACHHENT B The following questions are based on our review of some I ic'ensee submlttals to IEB 79"OIB:
1.
Licensees maintain that aging is not a required consideration
.for
. components that are Included ln a routine.periodic Inspection and ca'libration'rog'ram.s this a'cceptable?",'
2.
Licensees maintain that aging ls a generic'industry issue whose
. resolution is not clear; therefore, evaluation has not been made or will continue to be made as relevant Information is made available.
3.
Licensees are referencing manufacturers'etters as establishing the qualification of ancillary parts such as lubricants,
- tapes, etc.
Is this acceptable or are manufacturers'est reports requi red' e
\\
Limit switches used for valve position indication only have been deleted from the submittal.
Licensees maintain th'at a valve outside containment in series with one inside can have I ts position veri fied visually following an accl dent.
Is this acceptable?
5.
The licensees maintain that neither valve position limit swi tches, solenoid valves, nor control cables for air operated containment isolation valves'eed be replaced or protected from the adverse environment, including flood, because all postulated failures will result in the isolation valve assuming Its fail-safe posi tlon.
Is this accept'able.
6.
Some fan cooler motors do not me~t FSAR requirement of 1.5xl0 rads.
8 Qualification test was to I.kxIO rads.
Licensee states radiation
~
g I'
~
~ t I
ATTACHNENT 3a
ATTACHMENT B P
level is."close enough" to expected accident radiation level to.
be acceptable.
ls this acceptable to the NRC2
. 7..'ttachment D is. a.summary:of problems incurred during a,.one yeai':.:
opera'tio'n test of' containment fan'ooler. unit.
Mould you consider the test to be a success7
'TTACHMENT 3a
ATTACHHENT C l.
In lieu of a test report, what constitutes an acceptable Certificate of Compliance2
~
~
I
~2.
What if the test specimen and installed component differ, e.g
- model, type, etc2 3.
What, as a minimum, must be Included for an analysis to be accept-s abl e2 c]
,j 1$
The guidance provided in Enclosure 4 of 79-01B allows analysis (evaluation) for service conditions such as radiation and chemical sprays.
Is analysis (evaluation) and "engineering Judgment" the same,thing2 5.
Since effects of radiation and chemical spray are "allowed" to be analyzed (evaluated) for Important components such as containment
'lectrical penetratlons, Is it prudent to require a licensee to prepare a ful I b Iown analysis to qua I I fy a 7/C 12 AWG cable when a similar 5/C 14 AMG cable was actually tested and shown to be qualified2 6.
Provide us with + limits for evaluation of test data such'as
- pressure, temperature, radiation, duration, chemical
- spray, and aging.
7.
Host tests include only single components and the reports do not include any acceptance criteria.
Test conclusions are that usually, no matter what happens during the test, the component Is accepted.
This is comnonIy referred to as a "dead bug" test.
Provide us with minimum acceptance criteria requirements for a test and its report to be acceptable.
ATTACHMENT 3a
4J 4
~
~
)Ag Q
q
'A~y ~ q A
PAPA.~ CI 4 AA'PAAAO I Y K c~acato o
3 T
- 'L 'Ler 4
4 ~
s..)."all-~i"..3 is a brief 'cry.of the C ii"J i") (..) PP:ir w~ A..
3 A.
~ erst Lon r,mliiicati:n tost.
T...3 6:(
L::: rilL CF 1.;51:=s Lr.curred J)1 of t 40
. -A4CA l
i 12 13 75 12-17-75 12-25-75 1-13-76 2-17-76 2-21-76 3-3-76 3-10-76 3-19-76 4-12-75 4-22-76 4-2)-76 4-:0-75 75 '
1-"0. 2 373. 8 G27. S 164a 1741.5 2211. 5 2352.7 2 54 2"4'7. 5 2"A 3C07.9 3022.6 1-3/4 ]:=urs 2-1/4 t:"urs 2-1/4 hours 15 in.
9 1 "urs 15 min.
40 min.
4 ?~rs 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 10 Cays S-l/2 Lwurs 9 taurs 2-1/2 Cays Ihintcnance prob]its eau~cd lo)s of,"s~er.
Tmnsior"cr coil turned out.
Loss of plant pouer.
Electrical stors caused loss of plant poQdrP "pray rings pluE3ed.
Rig"A bypass or ccoling eater.'s"..ut~~m required to drill 14olcs in test ch"='her.
'lcnt
-aintenance r~ired cut-oH of pc~or.
Eloctrical storm c"used t~
short shutdowns because of tc=~rary poMer interruption.
Yri lcd on overload Faul ty soles'.oid caused condensate to tack up in chamber.
Unkr~.
hearing problem; see Appendix C.
Unknown.
LAnkroix - installed recording c,,ui:nt-"as ad voltage to C)tcc t rui s ".co tr Lps.
Luisarce trip " ~ ~ to overlcad hcatcr f"Lluro;: l-r,th of's'.ut-
~e ula AD]lure occur,.
L late Fr '.] ".-.d...-t) nat
~
~ ~
L)u.ly ATTACHNENT 3a
C I
)
~
"<<< <<r
)<<
f$ <<
a I )<<Q)'~) ~.
t
~ 4 ~ ~ ~ <<
<<a.r
(
C.
'C h
r
~
~
~ c.
~
(
~
I
~~ ma%
I
't'..E 5 9 76
"-15 76 TLk:- ON
)':)'.", ).
H~R
$140 4759.5 lE..~iV, OF
)
'.: )')!
Ko Tzip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> "F.):LC
)'roblen vith d~p valves hut corrected ~ithout ~hutdovn.
Fo~er failure 'due to
<<lcctrical stor 7 $0-76 5109. 8
'2-1/2, days
,.Tersiint.l Ward'upt'ured
'causir. loss of prcssure ctw.btr <<
'lnis board mes a sc=l re-~izod for testis" and mes not a part of equip"cnt being qualified.
S-14<<76 C-19-76 I) 19-76 5369. 2 5487 ~ $
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Trip 5502.6 6-l/2 hours Povcr failure.
Slight problc= in the controls c u-in" a slifht cycling of tc=pcr-ture.
Froblc= corrected
.vitI)out sLatdoav.
Fcu)ty solenoid zesu)tin>> in cc.",den"atc b"cLin" up in ch':cr ar.d c u"ir.,". rotor to trip o.. ovczlc,"d 4 28 76 C-$0-76 C-$1-76 5451.4 5725 5752 2-1/2 hours kb Trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Solenoid did not function properly end override cizcuit Kid not op".rate.
Notor trip;ed cn overload.
Tc"perxture vcs doun slightly.
Solenoid Lcd failed to operate but override circuit res vcntinq ci:t"."r.
Frob}c= v=s corrcctc3 Nithcut shutdo:~.
Rotor triF,"ed on ovcrlo"d.
Float.valve -stu=k cxusing con'cnsxto to tuild up in d~'.cr.
~ r AZTACWENT >>
va n ii v r h.v
> v p~ s p g eCo Fv)Lac ill-la, C';:9 T1ME ON DATE K%JR h!ETER 9-8-7e SQS4.".
'I LEH~ni OP toa>. TIM l-l/2 days REASON Srcakdoiw in electrical tape and other insulation -appl ied by JQY at-the ends of the. lc d
caused a short which burned off one iced resulting in a single based condition.:.
Les'd end
=i red.
11-".-76 7384; 3 2 days Lead separation at tezainal bocrd caused by brcakdo~ of insulation, lead c=brittle=ent and vibration.
Le 5 vns re~ircd.
11-30-76 7840. S 20 hou"s nsulction failure sinilcr to that of 9-8-76. caused unit to trip.
.'=:12-24-76 8590.9
.-~
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
?rproperly assc=bled
~~lcnoids cruscd condense c to beel; up in chc=bcr. resultint; in an trip.
12-30-76 SSS0 No Tril Lost phase sl; continued to operate in a single phase condition.
3-9-77 10145. 9 1-11
~ 27 '814. 2 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />
, Poser surge ccuscd unit to trip.
Because of r.in"lc p!=se conditio:.'nit could rot b" restarted..
Rcpcircd the lccd r.t the tc.=in l board ard rcsu=cd tcstir.S.
%his brc.': ur.s si"ilcr to that of -11.-9-76.
]
A~:.rent short in motor caused tc~irction of.,test I
~
ATTACHMENT 3a
1'Icu.
c~
P e~
~+g o
lf**44 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 799 ROOSEVELT ROAD GLEN ELLYN. ILLINOIS60197 July 10, 1980 THRU:
FROM:.:
MEMORANDUM FOR:
E. L. Jordan, Assistant Director, Division of Reactor Operations Inspection, IE:HQ Fiorelli, Chief, Reactor Construction'nd Engineering Support Branch D; W."Hayes", 'Chief; Engin'eering Support Section 1
SUBJECT:
FINAL REVIEW OF LICENSEE RESPONSE TO IE 79-01B-DRAFT TECHNICAL EVALUATIONREPORT FOR D.C.
COOK UNITS 1 AND 2 - DOCKET NOS. 50-315; 50-316 Attached to this memorandum is the draft technical evaluation report (TER) for the D.C.
Cook plant.
The format is submitted for comment b the task group; the content is submitted for review and resolution by NRR.
D.
W. Hayes, Chief Engineering Support Section 1
Attachment:
As Stated cc: w/attachment J.
G. Keppler G. Fiorelli R. Heishman Resident Insp.
V. Thomas, IE:HQ A. Finkel, RI R. Hardwick, RII D. McDonald, RIV J. Elin, RU Spec'ic Issues A'
AC:"""I"!'tT 31
DRAFT TECHNICAL EVALUATION REPORT OF THE ENVIRONHENTAL QUALIFICATION OF CLASS IE EQUIPNENT INSTALLED AT THE D,C.
COOK NU"'At( ruwER PLANT -,
RE:
IEB 79"OIB INTRODUCTION This Technical Evaluation Report (TER) is the result of reviews and inspection necessitated by IEB 79-OIB; however, it is also the culmination of IE inspection activities which started in 1975.
Attachment I includes a synopsis of speci fi c I E matters relating to the development of'his TER.
In addition to IE activities relative to environmental qualification, NRR has also accumulated several years of past experience, expended resources, performed reviews, held meetings and various telephone conversations, issued and made amendments to licenses, etc.
It is for these reasons that NRR performed the detailed review of licensee submitted test/qualification documentation.
This TER does not include information about seismic or fire withstand capability.
It should therefore not be inferred that equipment which is "environmentally" qualified meets all necessary.qualification requirements.
It should be noted that environmental data may be extrapolated directly to similar materials;
- however, because of significant differences in physical size, direct correlation may not be justified for seismic considerations.
~Por ose The purpose of this TER is to document IE's role and effort in the matter of IEB 79-0IB, and forward the results to NRR's Equipment Qualification Branch for their use in developing the Safety Evaluation r
e Report (SER) for the D.C.
Cook plant.
~
I
~
~ l Heehod On June 5, 1980 the licensee submitted the final response to IEB 79-OIB describing in detail thei r review of equipment used to mitigate the consequence of an accident.
Equipment installed in potentially harsh environments was of primary interest.
On June 16-17,
- 1980, IE-RIII performed a site inspection of specific system components in order to verify the accuracy of the licensee's response.
(Refer to Attachment I, item B5).
- Further, a complete review of the submittal was made to identify discrepancies and determine if the submittal references actually supported equipment qualifications.
(Refer to IE:RIII Review Results below).
IE:HQ and 'NRR were to review and eva'luate IE-RIII 'inspection results and other potential
- problems, provide guidance for uniform evaluation of licensee submittals, perform detailed reviews of all licensee referenced test/quali fi cation documentation submitted by I E: RI I I.
(Refer to Attachment I, items 81-4).
NOTE:
No response has been received by IE:RI I I
>r e
"3-to the above referenced inquiries.
Licensee Results In the response letter, the licensee stated the list of equipment was complete and proof of qualification to withstand adverse environment existed in all cases except where noted in the letter, attachment 4,
5, and 6 to the response, and in various worksheets.
A synopsis of those exceptions follows:
- Aging is a generic industry Issue whose resolution is not clear at this time.
As relevant information becomes available, the licensee intends to continue its evaluation of the sensitivity to aging of materials and components.
+Limit switches used only for indication of valve position will not be replaced.
+Components for the pressurizer relief valves, Including solenoid and motorized valves, limit switches, and terminations are being considered for upgrade as a result of TMI lessons learned.
- Instrument cable is located below flood level for several instruments.
+Control cable is located below flood 'level for several containment
isolation valves.
+Auxiliary feedwater flow transmitters located outside containment will be replaced.
- Emergency service water pressure swiches located outside containment will be replaced.
+Hotorized valves and terminations in the RHR System may not be qualified for exposure to radiation.
(Outstanding item).
IE:RIII Review Results In addition to the discrepancies noted during the site inspection, the following potential discrepancies were noted during review of the entire submittal.
Due to lack of guidance from NRR's Environmental Qualification Branch and lack of detailed systems/operations experience of the principal
- reviewer, resolotion was not possible.
- Continental instrument cable failed a
LOCA test as documented in Conax test report IPS-348.
Continental control cable, made of what is believed to be the same insulating and jacketing material, "passed" the same test.
Because only one sample of each specimen was tested and no real acceptance criteria except "survival" was used, it is the principal reviewer's opinion that the Continental control cable should not have been deemed acceptable.
Continental control, cable, is, installed inside containment and below flood level.
- No correlation can be made between the test reports and instrument terminations on submittal page Ti-l-l, Tl-2-1, and Tl-4-1.
None of the tests included submergence, yet the w'orksheets ind'icate terminations are qualified for submergence.
The terminations are below flood level.
- Worksheets Ti-3-1 and TP-1-1 infer all penetration terminations are made in flood up tubes; not true.
- Penetration terminatlons for motorized valves, hydrogen re"combiner and fan motors are located below flood level.
(Worksheet TP"3-1).
~Core cooling and containment isolation valves are located below flood level.
(Worksheet VI-1).
- Long term monitoring steam generator instruments are located below f1 ood 1 eve 1.
(Works hect
- 1).
- Long term monitoring boron injection tank discharge flow transmitters are located below flood level; referenced by emergency procedures.
(work-sheets 1-7-8).
+None of the referenced lubricants/greases had been "qualified" for LOCA by test.,'Letters from suppliers and/or manufacturer's specifications were referenced.
(Worksheets GI-Gll).
- In many cases, the component worksheet specified operating time is one year; in only limited cases has that time been qualified.
~Summa r Licensee submitted response to IEB 79-OIB.
IE:Rl I I performed review of submittal, made site Inspection and forwarded to NRR via IE:HQS inspection results and requests for specific guidance.
Conclusion It cannot be concluded, at this time, that there is reasonable assurance all components installed at the D.C.
Cook nuclear power plant are environmentally qualified, nor can it be concluded that installation methods of environmentally qualified components would not contribute to the failure of such components during a potential accident.
A positive conclusion cannot be made until all matters referred to IE:HQS/NRR have been satisfied, and NRR has reported the results of their review of all test reports.
Principal Reviewer
ATTACHMENT I
MATTERS RELATING TO DEVELOPMENT OF THE TER FOR D.C.
COOK A.
IE INSPECTION REPORTS 1.
50-316/75"04 2.
50"316/75-07 3.
50"316/76"01 4.
50-316/76-04 5.
50-316/77-09 6.
50-316/77-38 7.
50-315/78-08; 50-316/78-03 8.
50"316/78-09 9.
50-315/78-24; 50-316/78-22 10.
50-315/78-31; 50-316/78-29 11.
50"315/79"16; 50-316/79-09 B.
MEMORANDA 1.
2.
3 ~
5.
12/28/77 Hei shman to Reinmuth 1/12/78 Hei shman to Sniezek 1/16/78 Sniezek to Kniel 6/20/80 Hayes to Jordan 7/1/80 Hayes to Jordan C.
AMENDMENT 4 TO LICENSE DPR-74 D.
DRAFT TECHNICAL EVALUATION REPORT OF THE ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT INSTALLED AT THE D.C.
COOK NUCLEAR POWER PLANT.
ATTACiQZII7,3b
l.
2.
3.
4.
5 0 6.
7 ~
8.
'9:
10.
'll.
12.
13 14.
15.
16.
17.
18.
19.
20.
Auxiliary Feed Water Chemical and Volume Contr@,
Containment Air Handling Containment H2 Qnntrol
-Containment Spray Main Steam Reactor Coolant Residual Heat Removal Safety'lrigection'ssential Service Water.
Component Coolin'g Water Emergency Core Cooling Reactor Trip Actuation Post Accident Monitoring Remote Shut Down Monitoring Containment Isolation Main Steam Isolation Main Feed~ater Isolation Engineered Safeguards Actuation Main Feedvater Control E
~
<<Emergency Core Cooling PPressurizer and. Relief Valves
<<"Containment Radiation Monitor 1
. 'Licensee '.-"System '.List
'TTACK~~IPi 4
1.
Reactor Protection 2.
Safeguards Actuation co.
Safety Equipment Area Ventilation 27.
Containment Isolation 28.
Main and Auxiliary Steam Isolation 5.
Main and Auxiliary Feedwater Isolation 6.
Feedwater Control 7.
Containment Air Purification/Cleanup 9.
Containment Ventilation/Cooling 10.
Containment Combustible Gas Control a
High Pressure In)ection 12.
Low Pressure Infection 13.
Accumulators 14.
Chemical and Volume Control 16.
Pressurizer Spray 17.
Power Operated. Relief Valves 18.
Steam Dump 19.
Containment Radiation/Monitor 20.
Containment Radiation/Sample 21.
Component Cooling Water 22.
Service Water 23.
Emergency Power
~
24.:
Containment Surrp
'5;:'Control Roori Hab-'tability HRR Systems List ATTAC~>Zi'
CATEGORY I.
Zauiement is uglified for Plant Life a.
Equipment meets all applicable requirements of DOR Guidelines or IKREG-0588.
b.
Qualification by Judgement may be acceptable with sufficient basis.
II.
E ui ment is alified with Restrictions Equipment meets all applicable requirements of DOR Guidelines or NUREG-0588 with the folgowing limitations:
a.
Equipment Qualification for service life less than the plant life.
b.
Equipment requires modification to meet qualification requirements, such as relocation or shielding.'II.
E uimment is Exempted from Qualification Equipment where safety related function can be accomplished, by redundant fully qualified equipment which meets single failure criteria.
V.
alification of E uioment Unresolved a.
Qualification resting scheduled, but not complete.
b.
Qualification Records search still in progress.
E ui ment Not uglified Categories AT hCEZ"T 6
LFR' LER' ATTACH."KNT 7
UNB SOLV"0 G"N"RIC SP"CIFIC ISSU:"S 1.
No answer was ever recieved to the Generic Issues, Attachment 3a, discussed in Attachment 2 of memorandum Hayes
.o Jordan dated June 20, 1980.
2.
No answer was ever received to the Specizi<
~ i~~u~~, Attachment 31, discussed in the attachment of memorandum Hayes to Jordan dated July 10, 1980.
Unresolved Generic Specific Issues A PACESdT 8
A.
Pending review and approval
'by NRR/EQB of test reports or other documentation.
B.
Reviewer not qualified.
C.
Test specimen different from installed equipment.
D.
To be replaced; when not specified.
E. Fails Safe.
F; G.
h
\\,
~
'Operates'efore fl'ood.
Effe'ct of 'short circuits on containment'lectrical penetrations not consi'dered.>>
r I
Analysis Engineering, review.
H.
Limit switches used for valve position only.
X.
Long term testing not performed.
<<>> 1, 2
J.
Ancilliary components being considered for upgrade.
K.
Calculated containment temperature 2.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> (10,000 seconds) after a 0
LOCA is 185 F and. decreasin5.
The electrical cable is rated for continuous operation at 194 F (90 C).
Therefore, the containment environmental temperature after 2.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> does not represent a
challenge to the mechanical or electrical quality of the cable.
(Margin of 9 F; EEL 323 is 15 F)
L.
230 F for ten seconds and 11.5 psig for 0.1 seconds will not 0
challenge the mechanical or electrical quality of the termination; M.
No details on aging.
N.
Aging:
250 F/7 days; 300 F/54 days; 390 F/20 days; 302 F/7 days; 360 F/4 days; 165 F/8 days; 200 F/8 days.
Vhich one is acceptable?
1 0.
Referenced by emergency procedures.
P.
Materials listed in Table C-1 of ZEB 79-013 indicate a potential for significant thermal aging when exposed to normal ope ating conditions.
(XLPE/Hypalon 40 years; EPR/Neoprene 10,years)
Not included in routine replacement program.
R.
Radia ion specification determined by ArPSC NSEL calculation DC-N-6420-2.
(Not reviewed.).
S.
Shielded; not in direct line of stated accident.
>>Letter 1 inghaus to <niel datec Septe. ber ques ions 40.6, 40.10, anc 40.12.
,(At ached.)
"~ '. r, 29, 1975; FSAR Appendix Q
h Concurrence Code g+,h rh
~ ~
Q 9
'h
~
~
T.
No credit for assumed operation.
U.
Limited test length.
V.
Main steam line break plus failure of another steam line to isolate is acc~p~~~l~.
~o steam generator stop valves can be affected by any one break.
W.
Unable to evaluate; nine test reports listed.
. X...; Even though. beyond.;the scope of.IEB 79-01B; it, is unclear. whether in Unit 1 justification has been made for use of cables which did
.not,pass the.IPCEA-19-8l.vert;ical flame resistance test.
~
Reference:
November 25, 1'977 NRR.letter to licensee, subject:
"Summary of Meetings Held on November 3 and 4, 1977, Reference Fire Protection" (Question 49).
Y.
Below flood level.>>
Z.
Generic Issue The issue of possible short circuit interactions with safety related.
equipment including potential degradation is beyond the 79-03'eview scope.
However, this issue was raised as part of the Unit 2 license review under Reg.
Guide 1.63 conformance (Q40.6)
(a copy of Q40.6 and its response are also included for your information and use),
and AEP responded to this issue by installing redundant circuit breakers for all equipment inside containment (both safety and non-safety-related) to alleviate this sort of problem.
Installation of this system was a
Unit 2 license condition and is 100fo installed in that Unit.
We are also installing the same protection in Unit 1 even though no committment was made to the NRC to do so.
The work in Unit 1 is 955 complete.
"<<1 The penetrations which are specified for long term operation are qualified for more than one year's expected radiation dose.
In addition the penetrations have been tested for submergence (test ran as long as 6 days) and performed properly after the testing.
Documentation of this can be found on pages 8 and A5 of IPS-326 and pages 4 and A4 of IPS-327,
"<<2 The subject WCAP ac'dually contains two tests.
A "short tenn" test was run in which the fan motor was operated. without its associated.
heat exchanger which provides component cooling.
The second "long term" test involved. operation of the motor with the heat exchanger operable as it would be during the long term post-accident configuration.
We Westinghouse gas determined that the long term post-accident operation of the motor is qualified for indefinite long term operation.
ln our upcoming suomittal, the quali¹cation chart for this fan motor (Fl) will be formally updated to reflect this position.
The fan motor in question was carefully identified by AEP.
As per our
- request, the supplier (Westinghouse) nas explicitly stated. that this fan motor is qualified by the subject report.
(A copy of the letter documenting this qualification (A"P-79-583, Noon to Caso) is also attached for your information).
List A Engineered Safeguards Actuation
~
. Containmept Phase A,Isolation,.Actuation ;".;...,:
Main Steam Isolation Actuation Reactor Trig Actuation Main Steam Normal System Monitoring.
List 3 Engineered Safeguards Actuation Containment Phase A Isolation Actuation Post Accident Monitoring Reactor Trip Actuation, Main Steam Normal System Monitoring Remote Shut-down Monitoring List C iist D Engineered Safeguards Actuation Containment Phase A Isolation Actuation Main Steam Isolation Actuation Main Feedwater Isolation Actuation Reactor Trip Actuation Reactor Coolant Normal System Monitoring iist E Engineered Sa eguards Actuation Containment Phase A Isolation Actuation Remote Shut-down Iionitoring Reactor Coolant No al System Mon'tortng
\\
r r
Engineered Safeguards Actuation Containment Phase A Isolation Actuation Post Accident Monitoring Containment Phase B isolation Actuation Containment Spray Actuation Main Steam isolation Ac uation List G
Containment Phase B isolation Actuation Containment Spray Actuation Main Steam isolation Actuation.
List H Post Accident Monitoring Main Feedwater isolation Actuation Reactor Trip Actuation Main Feedwater Normal System Monitoring list J Reactor Trip Actuation Post Accident Monitoring Remote Shut-dwon Monitoring Reactor Coolant Normal System Monitoring
~.
Y(estinghous Bectric Corporation Nater Rc";lor Divisions Huctear Service Oivision Box 2728 Pittsbnrgtt Pennsy! vania 15230 AEP-79-583 July 3, 1979 Mr. L. F.
Caso Hectri cal Genera tion Section...
American Electric 'Power Service'orporation 2 Broadway New York, New York 10004
Dear Mr. paso:
Subject:
Containment Air Recirculation Fan Motor Environmental qualification
Reference:
AEP-5479 dated March 30, 1979 and AEP-5501 dated June 18, 1979 This letter is to advise American Electric Power that the containment air recirculation fan motors installed in D.
C.
Cook Units 1
and 2 have been environmentally qualified by Westinghouse.
The environm ntal qualification of these motors is documented in the following Westinghouse MCAPs, topical reports, and drawing:
- 1) WCAP-7829 "Fan Cooler Motor Unit Test"
- 2) MCAP-7343-L "Irradiation Testing of Reactor Containment Fan Cooler Motor Insulation"
- 3) MCAP-7396-L "Safety Related Research and Development for Westinghouse Pressurized Mater Reactors, A Program Out'.ine Fall 1969"
- 4) MCAP-7498-L "Safety Related Research and D velopmeni for Westinghouse Pressurized Mater Reactors, Program Surmaries Spring 1970"
- 5) MCAP-7722 "Safety Related Research and Development for Westinghouse Pressurized Mater Reactors, Program Sumnaries Spring 1971"
- 6) WCAP-9003 "Fan Cooler Motor Unit Test"
- 7) Westinghouse Drawing 206C391 "Safeguards Containm nt Motor Cable Connector Insulation"
- 8) Containriient Air Recirculation Fan Motor Seismic Analysis MCAP-7829 was transmitted to Ms. Joan;Etzweiler on April 9, 1979 by my letter AEP-79-539.
MCAPs-7343-L, 7396-L, 7498-L, 7722, 9003, and Westinghouse drawing 206C391 were transmitted to Ms. Joan Etzweiler on Hay 4, 1979 by my letter AEP-79-Shl.
The fan motor seismic analysis was transtttitied to you on May 8, 1979 by my letter AEP-79-555.
AEP-79-583 Julv 3 1%70
%ease be aware that Mestinghouse drawing 206C391 provides details on the aethod of connection of power to the motor qualified by Westinghouse and does not necessarily apply to the connections at D.
C. Cook.
It is 6'."'"
msponsibility to.ensure that the method of connection and the insulation materials utilized at the D. C.
Cook units is equal to or better than that shown on drawing 206C391.
The information contained in the subject HCAPs, topical reports, and draring
"'illprovide documentation of 'Mesti'nghouseualificati'on testir'g 'for seismic,'LB and LOCA.
Also included are the lubricant used and its condition after the test, the Mestinghouse nethod of connecting the motor leads to the power
- cables, the insulating materials used by >lestinghouse on the motor pigtails with details of the motor qualified in a simulated accident test.
I trust that this information will allow you to respond to IKE Bulletin 79-0'l.
incerely,,
I 'o"~
Noon,-. flanager Eastern Region 5 MNI Support GGP:cam cc:
D. Y. Shaller J.
G. Kern
E
Response
to Additio ml LMor~tion
'n ECCL A~ysis requested by..'BC
=',. Eonald C.
Coo'.:.'nuclear'Plant
~
. Boche.'.. y.0-315
$0 316
. DPR 58
'. CPPli Zo. 6l, E+
E
'E September 29, 1975
~
t E
~ '
E
~ ~
IQ ~ Karl ~el, Chic iJ4 Sate reactors
=': an h;lo. 2-2 "rision of Beactor Licensin"-.
J ~
~
J:u lear Zc(ulr~ very Coz-"ission
'-.;asM~tc-,
D.C.
2C-"55
~
~ E
==Dear
. ~el:==
~
E The folio=ing is in response to your let e date=
Ju y 1G, 1975 recuesting additio.".al infor"atiorJ in o de to allo+ the staff to co-delete its eve'tion of the Donald C.
Coo'c
.'boucle. r Plan
> 0 Dt -.'o.-l
"-CC3 re-analysi's.
Ne conside" tx~s additional in>or ation only as
.'e believe.
om July 9 1975 su" ">> ttal of A"enK~ent 6-: to the OoJal" Coos.
> "ear Plant Final
~ feiy Amlysis.-;effort fulfills
" q~ire."..ents of the Dece=ber 27 1974 Order for
'adificat on of License and reets the requiremen.s of, Apyen"x iC to 10CZA50.
Tteu 1 - Break
'~ectru~
an". Pa"tiaL L~oa G~erat'~-..
The provisions of'he
.'L C memorandum entitled "FJinimu Zea'mre ents for -CC3 Break ~cectrm.>ubmttals" have been co olied --its a of our s"b'"it al of A;"enaent 64 to the Oonald C. Coo~.'Juclear Plant rdA-.
info -ation on, 3-looo operation is currently boir" evaluate".
J. til thi rc-rie'r i-co-deleted, ve ~a.ll a'Mr.-
istrati':4>>l.- res4=ict 3-'co" ooerat'on in accordance
~ ith the re
~ i=e~-ents of th= P-8 set"oint fo four loca o"erat'on as yrovid..:"in Table 3.3-l on ",="e 3/I+ 3-7 of the Tecum'cal -oecifications o:
t;".e boa ld C.
Coos.uclear
- P.lant.
I
Hr Karl. ~el.
September 29, 1975 2 ~ >
a+
9 yl tea D sy4q4~
Wn~
<e have u~~U.e"
."..= '=".:.~tiong~u have req'ster D our let+or of A.
3.1 14 1";7y.
~3."co teat "uM.ttx" ve have "=cci;.sQ aQh i'd& I~a":.ztiaa "rc~:i~stir-"sos re~ardir;-.-. "own preci~itatioz folio;diag a hot le,"; cram an'ot 1e= reci'::1"xioa.
TMs ir"-or~tiozx.:Zs t".
M~ted to the
~..Q ia. '..'e=-".
hpas" le"t,~ -'-'.'-3,"+,.rioted
.My,..2'., 1975'...
~ cozy"of t~~s' sti'~ house 3.etter is attic>ed to sugrlc"ent a" " -~on" of April 1+
1975.
p he desicn oz the Dom~ld C. Coos ".>clea
&la.-'.t U"~t
~h~ 1 ir."sr~orates an analysis oz cossible=failure "o~es
'ooth active an." ~a."s v, o
'C~'a'z" p=ent a d theU effec s on
~ C.i gerfoz~nce
~
"'1e =CCD sp si nU iI~Dtalle" on the 'ol'~ld mo'.: "uclea.
Plan+ satisfies
+..e Ge". rat-i=si.".n Crite ia
'~":;ucl"=a=-
&~:re>> ilants {A-..yon"ix A of 10 Cz.< >0) as;.cl3.
ue ~ro isioas of Je tio~ 50.46 a.".Q Ap"erdi-.. K to 10 C9'3 0
" l
.'Leo'ci..~lollse VC evaluat oQ '
el {s~a X78d
~t~+ ~;.CD3'-i "3o) revie';red tne rarge of go~"ib'e ai1ure roc'.es
~C~ eau'~:.mt
~".8 assume~
the asst dm"-i.""- si~,le
-"ml"re o
'CC3 eqMa~mt The Do&'. Coo~..'aclear P'a"+'e ipse Critorka clu~es an aI'a g'siD 0 va1ve fa>> ldzes Nt coed a4'Je selt a =ect 0 ~.ce
~
As re=u<steJ i. -.ou =8'~; '0 3.ettcr, "e ~ve -a-e a ~a'ther s ~.'y ox'n~
=CD~ valves
."=';;.e '.nve rot heea aole to dete ~we a failure;o".e o~
electrical
~@~to~ cxporent.i'~t co'~lo ca~zsc aa
> ".desire~0
..echanical ation, of a v@3.vo or o+':ver flu'd spate;". cor"ionmt.
5.".iis "s esoeciQly true on t.'".e ~.=";3." C.
Coo=:. clo"r Plan
~e=ause a 'o;~trol ci=cuits to these valves a' provic'.ed cor acts on bo
".. t'".e go~.er s'de an" ti>e garou a,
ii'e oZ M con rol ~o;.or> thus a am~i.
- t.: t a sih.-le ~rouaD could not res"'t in x~'vertvat valve ac".mtior..
o ~>rese t sg stn ~ ov a" tomcat co 2 "r".'Qtgrg valve @os't"orna" s'-'.mals folio:rin:; a ~CA oyerati-..~
yroce""~cs,:~Z'>.:> a'".'ac'" cat-on, era o orator rualifica io.".s aa5 pe iedi" rot=a" r~'.~ prcclu"es the likelihoo" of a vis"ositio.". d
.~@1 e t'wt co'ad "esu1t ia t;;e loss of s-ste~
safet=
-" "".ct o
-ore liDti:i~ xhm tn-t i~cor,orate:1 in the desi a criteria J.or t.':e Zomld C. Goo-clear Plar;t.
~ >
I
r
~
~
~ l i ~ a 1
~'
apta.
.er 29, 1975 branch Technical Position EICSS 18 regarding ingle failure criteria of manually controlled electrically operated
. valves is inconsi tent vith past prac.ice regarding the General Design Crite"ia for l'uclear Pov r Plants.
Xt should be furth "
noted that this type of single failure vas not applied to the
.- Donald C; Coos Iiuclear Plant.
Xn our )udgment, there is no basis for a~plying this interpretation to tne Appendix 6 ECCS evalu"tion =cdel.
Since tnis int rp etaticn nas not been mad~
at the tiae the ECCS regulations
.~ere pro-ulgatea, it cannot be consider d an appropriate inter"retation of 10 CFR Part 50
. Appendix K,Section I.D.1.
Me firmly believe that Branch Technical Po ition
'.'EICSB..18.should not:be.,app&ed to'he Donald C.'ooR I:iuc'leer Plant'or both the analytical re-sons and the special design
. features described in the preceding paragraphs'e have ho@ever> defined the valves >:bien -ay be affected by Branch Technical Position EXCSB 18.
Tne e valves are listed belch;:
ve '.u, ~~r
. IHO 910 DIG 911 XMO 261 IHO 390
.IRV 110 XRV 120 IRV 130 IRV 140 IMO $1 ZHO 52 "IMO 53 zoo 54 RlllaL'ul PS Suction to charging pu ps from RbST Suction to EPB pu ps from
. RVfST Suction to Safety In)ection Pumps from P,%ST
.Accurulator drain valves to reactor coolant drain tank Valve from charging pumps to reactor coolant systea
'old legs Plot: Di"~ra"-
6.2-1 6'2.IA 6 2-1 6.2-A 6.2-1 It should be noted that valve DIG-910, 911, 261 and 390 vill be required to change position fol3.oisin"- t'".e postulated LCCA,.
EEV-110, 120>
130 and 140 are closed durin~
the entire c"ur e of the acciaent, and IlI0-51, 52, 53, and
'4 are open during tne entire course of the accident.
Item 4 -
ub. e
.ed Valve".
Ve have revieved all the valves with motor operatcrs inside of the contain "ent vitn re pect to ECCS performance-analysi
, and ccntain~ nt isolation, vhich includea the require"ent for liaiting boric acid concentration auring long term cooling.
Ka l a>gey V
ClT>
t The valve
. X'os.o' a loss of operators t?at ~~~X'he.s"-
coo~--% acc'.cnt izcl -+a the "o' J
IIqVee ) g V..
QV
~
1 II
\\
1
"~
'C '
..'YQ-llG
-': Z>Q-120
. Zi-;~-1>0.
.'. ':-.aZ-14
'.0 Q
'.0-g~
Xi'.l'-g3 p,
~T"-0-l"'8 XC.:;-3 >g.
~ 1
~ Lta,o>>>>J V r
- r
~
I
'I r
Chirgin~ ~ Dischare to reactor coola s steh les.
cols r.
I
~ IP
'.loraa3. Cooldovn -rom >mt Z.e
.... 6. '-'
i4araa3. cool"o"w to cold le=a
.6.2-"'.
Reactor coola"t yu"p sea" t~a'.2-1 Accu~ator "t"c?~pe V~~
6.='-.-':,.
Xt s'auld be ~te~ that'. a"h oz t"e z~"ve v"'. '-:
Ss z'ega ed to re.".p.
i.
8 fixed osl ~0& Z.olla%"
'.e., valve o~eratlo>> of t'.xe a~ave "alves is not.;r~~".':r".
o 1 ovmvg tale acci Qeiito I
Fu ther rev~ev has ho:;,~ t~~t =~ rate-~o=."= c. -"".-.;-.-
o~
he azve valve "otors >Cll ros"
~ =re@ t.e s:z+.:"=";".='c."..
of the c.~ve valves.
~he details o. ta's evie=- --'==-.':
belo'.T ~
. V
~"'hese valves a
e AC."uto"-o"..erato'~t~
-.= ""
s..polie at c00 V':-sta"te
'o ate" in.':ato co;."=='e~tcrs o tsi"e of the conta':e:cnt
~~"~ t serefo e,
from th-valve o". "a+or
~~a:t. ti:o floo" cori+'on.
Q.Parol o
these valves is a v 20
'v
',v a'2g, 9.s fzon co"t~l tra='sfor;".ers de "cated to sac.". va';e a.
Va3.ve ope.'a~r
.";eat rs m.e ac-merci at 1"3 A
="mc..
sepa=ate va've heater dist:icutio.
""-5'~'ts.
~ '
~
'. 'he cont ol ele e.".ts subl ct to flookin~ >t ~l~~irical conductor, "11 not result in energizin ei her of the e
coils becau e the control circuit is so designed that the control sw'tch and/or safeguards relay in the control roon isolate the e co'ls from the re. ainder of the control circui.t
. util such ti"e as the operator or the safeguards logic elect
.to change, valve position.'..
It ad.ght also be pointed out that flooding the above valves will not re ult in spurious operation
, oX'ny other valves sub erged or otherwise as there are no interlocks to oth r valves made on 'the sub'ect
. Valve position switcnes.
As a result of thi review it is concluded that no design chan=e are required to solve potential ilooding problems.
Item 5 -
ntzin. ent
. essu e
Ve are infor ed by Ve tinghouse Electric Corporation that the contain"ent pres ure used to evaluate the perfor=ance capability of the ECCS analysis is con istent with tne Branch Technical Position.CSB6-L, and that t¹ model was approved by an liBC 'etter.to Vestinghou e dated Hay 30, 1975.
~
"e
'ten 6>>
w ECCS Pelood Pate The Mestinghouse ihrch 15, 1975 ECCS evaluation model was used in the Donald C.
Coos iiuclear Plant
.Appendix K ana ysis, as noted in Appendix P to the PSAR.
Un t
- l for-..io l
The above analysis is applicable to the Donald C, Cook liuclezr Plant Unit 1 with 25
.x 15 fuel.
- However, as we have ir,formed you previously, it is our intent at
'this ti-e to use 17 x 17 fuel.
Nhen the above Xnfornation
, becomes available for operation with 17 x 17 fuel, it, will
0
~
p
~
Karl Z~iel 6
September 29, 1975 I
be. @applied on a tinsel@ basis consistent vith the initial Un't 2 core ioadin> schedule.
Very trulJ Jou"s,
\\
e
~.
~
0 lg Enclosure
- k. '.~alsh
- obert J. Volley B. C. Callen P.,
- .i. Qteketee
.. M. Jiugensen - Rf.d<an
~
rr 3, ~
s) ~ Run ver 43
ico Presi ent bc:
A. 8. Grfues~i. L. Shober 8
Z. 1Q.lioti/P. V. Daley J
G. Fe~nstein Dl-.>-666$.7 DC-2-60/9 H Z. Olden/T.F. Kir~/R.C. Carruth Xi" ni - &ot5. j 1
r L
\\ ~
'Ic',1g"
~
p E
1
a l
ea V
t State one extent of conformance of the safety'ystems to Regulatory t'uides 1 41, 1 63, 1.75, 1.93'lso for each case Mere the D. C. Cook Unit Number 2 design does not conform to a particular recommendation contained in these Regulatory Guides, state the technical bases for the acceptability of,the D
C.
Cook Unit Number 2 design II The E3.ectrical Designfor 'D.'; Cook Plant complies with Regulatory Guide 1.41.
Pre-operational verification tests demonstrating train independence will be performed, ~ 'hese tests will demonstrate the full redundancy of plant safety systems, their power supplies and control power requirements, AC and DC, in conformance with the regulatory position.
D C
Cook Technical Specifications are in compliance with Regulatory Guide 1 93.
6 The electrical containment penetrations vere purchased based on specifications developed in May, 1970.
The electrical penetrations used at D
C.
Cook Plant are in compliance with IEEE 317-1972 and Regulatory Guide 1.63 except as follows:
Section
- 4. ~ 1 Mechanical Design.
The D. C. Cook penetrations vere made in accordance with ASME Boiler and Pressure Vessel
- Code,Section III, Subsection B, for Class B vessels.
Welds were magnetic particle tested in lieu of the radiographic method as per ASME Boiler and Pressure Vessel Code,Section III, Appendix IX, Article 3X - 350 d
6 Paragraph C.l.
The prototypes of the electrical penetrations were tested for the momentary values of the fault current.
The'vercurrent protective dev'es, either safety related or balance of plant, are a11 of the same qua1ity level Redundant protective devices are not employed.
Paragraph C 4.
The penetrations vere manufactured, in accordance with Mil. Spec.
Mil-Q-9858A which meets the functional require-ments of ANSI H.45.2-1971.
Appendix Q
.Unit -2.
Amendment
.79
...November, 1.97'7'.--"
Response
to Question 40.6 (con't)...
Safet Guide 1.75.
The. design of D. C. Cook Plant complies with the separation requirements of Safety Guide 1.75 as applied to Class ZE equipment and circuitry.
The design does not: c.~p>y with the safety guide in the treatment of associated circuits.
Eon Class IE cables are routed with Class IE cables
~ in cable
+ays.
The cable numbers of these associated circuits are modified to include.,a 3.etter. designation...identi;fying the train
'ssociation.
These cables are allowed to leave the Class IE cable trays and be routed with non safety cables but are not allowed to be again routed with Class IE cables.
Power cable trays are of the open ventilated construction.
Control and instrument trays are made of steel with solid sides and bottom.
Tray covers are normally provided only where mechanical protection is requixed for the trays.
Cable tray installations where the minimum separation distances specified in 5.1.3 and 5.1.4, as'pplicable, are equalled or exceeded are not supplied with barriers.
Where the specified minimum separation distances are not maintained, fire-barriers consisting of 1/2" haranite are installed between trays for horizontal. separation or on the bottoms and sides of the upper trays for vertical separation.
The minimum vertical separation distance where barriers are employed is one foot.
The vertical arrangement of cable trays of opposite trains is permitted only where trays of one train cross trays of a different train.
The barriers installed on the bottoms of the upper tray extended 18 inches beyond the crossing.
The maranite barriers 'attached to the bottoms and sides or mounted between trays prevents direct flame imp'ngement on the trays and provides a fire proof barrier of extremely low thermal conductivity.
Heat rising to the upper tray from a fire originating in.a lower 4xay is directed around the upper-tray and heat transfer to the upper tray is. effectively reduced.
Appendix 9 Urd.t 2 4O. 6-2 Amendment 79
- November, 1977
4 Identify all electrical equipment, both safety and non-safety, that may become submerged as a re'suit of a LOCA.
For all such equipment that is not qualified for service in such an environment provide an analysis to determine the following:
- 3. ~
The safety significance of the failure of this electrical equipment (e.g.
spurious actuation or loss of actuati.on function) as a result of flooding.
The effects on Class IE electrical power sources serving this equipment as a result of such submergence, and "Any'roposed design changes 'resulting'rom"this anal'ysf s.
The lists submerged aO 4
below identify all electrical equipment that may become as a result, of a LOCA.
The motor operated valves which will be submerged as a consequence of a LOCA are the following:--
IMO-110 0 -120 IMO-130
~~ -140 Accumulator discharge valves IMO-51 n
52 1f 53 ii 5g IMO-128 ICM-129 ICM-111 QCM-250 Charging pump discharge to reactor coolant system cold legs Normal cooldown from hot legs Normal cooldown to cold legs Reactor coolant pump seal water These M. O. valves are the same as those listed in Mr. J. Tillinghast's letter to the NRC of September 29, 1975.
These valves were studied and the comments made in Mr. Tillinghast' letter -were-re-verXfied.
Further,,
the accumulator 'discharge.valves power supply breakers, in the motor control center, are kept open to avoid tne possibility of any inadvertent operation.
mo ""
mo-M S~
Damper motor for hot sleeve ventilation an 11 1t
$ 1 I1 lt If After a LOCA the above two valves limit switches and rotc"s will be flooded.
The pos'tion of th se aampers foowi"g a LOCA is not critica1 and their. misoperation
;it'ccur ~~,
will not affect the safety of the plant; Appendix Q
Unit 2 40-10-1 Amendment 79 November,. '977.
M-"ponse to Quest.~on 40.10 (con't)..
valves which willbe submerged after a IAKh, are ".e following:
Test line from Accumulator tank 51 Valve XRV 165 XRV-175 XRV-185 XRV-156 XRV-166 XRV-17.6 XRV-186 XRV-50 X6%-100 DRV-150 DRV-1 HRV-101 QRV-150 RRV-103 0 '
0 r
N g2 N
Teat, line from Accumulator tank 3 Valve 0
N N
Ie O4 u
RHR Injection Line 1 sl 0
N 0
1 2
N
, r.
..e.. ~:a...".s.:
~ -
e.....
s'."i'3.
a
~
a N
e s
N Accumulator fillline valve from Boron Injection Tank Accumulator fillline valve from Safety Xnjection Pumps Reactor drain line to drain tank valve Reactor coolant drain tank emergency drain valve Pressure relief tank drain to reactor coolant'drain tank valve Hot leg loop Ol sample valve RCP seal water bypass return valve Pressure relief tank letdown to vent header valve All these valves are operated from a 250 VDI source.
The elements subject to flooding are the operating coil and limit switches.
Study of their electrical control circuits showed that in case of a flood, shorting of the limit switches will not result in energization of the operating coil or changing in the position of the valve; this is so because control switches in the control room, that cannot be short circuited by the flood, effectively will isolate the operating coils from their power sources.
No interlock exists among these valves, or from
.them to any other equipment in the plant, therefore flooding of their limit switches and spurious coil energization, should it occur, would not result in a spurious operation of another device.
Appendix Q
Unit 2 40-10-2 Amendment P9
- November, 1977
. c.V e 'll