ML17317B344
| ML17317B344 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/22/1979 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Tillinghast J INDIANA MICHIGAN POWER CO. |
| References | |
| NUDOCS 7907240605 | |
| Download: ML17317B344 (26) | |
Text
liay 22, 1979 t
Docket thos.
and
-316 Ib'. John Til'iinghast, Vice President Indiana and Hichigan Electric Company Indiana and Hichigan Po)ver Company Post Office Box 18 Bowling Green Station Ne>v York, Wee> York 10004
Dear Hr. Tillinghast:
SUBJECT:
NRC STAFF REVIEW OF RESPONSES TO IBE BULLETItlS 79-06 AtlD P~~ so -30s j4I He have completed a preliminary review of the licensee responses to IEE Bulletins 79-06,79-06A, and amendment 1 to 79-06A.
The pur pose of this letter is to advise you of the preliminary results of that review, with particular emphasis on potential problem areas, and to identify related concerns
~ihich
~ve believe require your further examination.
1 Ile have scheduled a meeting ~ith the os~ners of all operating plants having Westinghouse de'signed nuclear supply systems.
This mieeting v>ill be held on Hay 30, 1979, in rooms P-110/114 at our Phillips Building office in Bethesda, Haryland.
You are expected to attend the meeting and, be prepared to discuss those matters identified belovI along with a schedule and procedure for providing the information needed by tlRC to complete the review of these issues.
1 (1)
Our pre'liminary review of the Bulletin responses indicates that
.a number of the Bulletin items are not yet satisfactorily resolved.
Enclosure 1 provides a summary of our current. assessment of the responses to the Bulletins issued on Mestinghouse plants.
(2)
In certain instances, licensee responses differ, viithout apparent justification, from the Westinghouse recommendations for individual Bulletin items.
Me expect to resolve each such differelice, as <cell as licensee exceptions to specific Bulletin responses, prior to our approva't of the Bulletin responses.
A copy of the Westinghouse 7/d 7
. 'ecosssendations is provided as Enclosure 2.
(3)
The, Westinghouse advice is prescriptive on resetting of the high pressure injection system and incomplete as to the need for keeping 1e r'a orrscst~
gURSeAldlt+
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
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Hr. John Tfllinghast Indiana and Hi'chigan Electric Co.
. Indiana and Nfchigan Power Co. (4)
We are finalizing a generic repor t on THI-2 matters related to Westinghouse operating plants.
Although this report fs not yet,
- complete, among other things, we expect that it will recommend further analyses of transients and small reactor coolant system
- breaks, the development of appropriate written procedure, guidance to operators in. the use of these neu procedures.
(5)
In certain instances, licensees are using fuel and relying on, safety analyses,.
which were not provided by Westinghouse.
As a result, it is not clear to us what the respective roles~of the-licensees, Westinghouse, the fuel suppliers, and/or other parties should be fn implementing those requirements described in item (4) above.
Ue need a clear and concise definition of their respective roles in these cases.
(6)
The Advisory Committee on Reactor Safeguards (ACRS) has issued five letters to the Commfssion as a result of their examination of 'the TtlI-2 accident.
Me need a clear and concise position from all licensees with respect to. each of the recommendations contained in these letters.
A summary of the ACRS recommendations is provided as Enclosure 3-(7)
Individual licensees have indicated an interest in meeting directly with the staff regarding the Bulletin items for their fa'cflftfes.
Experience to date has demonstrated that the staff does not have time to meet indfvfdually with each licensee to resolve these items.
It fs clear that there are a significant number of technical issues yet to be resolved for a large number of Westinghouse operating plants.
There are limited resour ces available within the HRC staff to perform the necessary wolk.
This situation fs exacerbated by the need to conduct similar and concurrent activities with those owners of B8M, C-E, and GE designed operating plants.
At the same time, there is a need to resolve these matters promptly.
To resolve the fssues described above fn a prompt and expeditious
- manner, we believe there is a compelling need to establish an owner's group for Mestfnghouse operating plants.-
Me expect that such a group would be needed for the remainder of calendar year 1979.
Owner's groups have worked effectively. in the past in minimizing staff and industry resource requirements to resolve other generic problems.
He strongly urge you.to meet with other owners of Westinghouse operating plants to consider the formation of such a group prior to our meeting on May 30.
This will be one of the principal agenda items at that neetinge orrscc~
CgfSHAllSW DATE~
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Hr. Dohn 'Allinghast Indiana and tHchfgan E'lectrfc Co.
Indfana and tffchfgan Pointer Co. Please note that investigation of a number-of areas related to the THI-2 accident, including the long-tern~
ACRS recommendations and long-tera action items from HUREG-0560, vifll be specifically, included as part of the future "Lessons Learned" staff actfvfty.
You can expect additional cor respondence fn the future on these items.
If you require any clarification of the vtatters di'scussed f'n this letter please contact Patrick D. O'Reflly, the staff's assigned project manager for these activities on tlestinghouse plants.
ttr. O'Reflly may be reached on (301) 492-7745.
Sincerely,
Enclosures:
As Stated A. Schwencer, Chief Operating Reactors Branch 8'1 Division of Operating Reactors ce:
w/enclosures See next page
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0 PS UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 2II555 May 22, 1979 Docket Nos.
50-315 and 50-316 Mr. John Tillinghast, Yice President Indiana and Michigan Electric Company Indiana and Michigan Power Company Post Office Box 18 Bowling Green Station New York, New York 10004
Dear Mr. Tillinghast:
SUBJECT:
NRC STAFF REVIEW OF RESPONSES TO I&E BULLETINS 79-06 AND 79-06A We have completed a preliminary review of the licensee responses to I&E Bulletins 79-06,79-06A, and amendment 1 to 79-06A.
The purpose of this letter is to advise you of the preliminary results of that review, with particular emphasis on potential problem areas, and to identify related concerns which we believe require your further examination.
Me have scheduled a meeting with the owners of all operating plants having Westinghouse designed nuclear supply systems.
This meeting will be held on May 30, 1979, in rooms P-110/114 at our Phillips Building office in Bethesda, Maryland.
You are expected to attend the meeting and be prepared to discuss those matters identified below along with a schedule and procedure for providing the information needed by NRC to complete the review of these issues.
(1)
Our preliminary review of the Bulletin responses indicates that a number of the Bulletin items are not yet satisfactorily resolved.
Enclosure 1 provides a
summary of our current assessment of the responses to the Bulletins issued on Westinghouse plants.
(2)
In certain instances, licensee responses differ, without apparent justification, from the Westinghouse recommendations for individual Bulletin items.
Me expect to resolve each such difference, as well as licensee exceptions to specific Bulletin responses, prior to our approval of the Bulletin responses.
A copy of the Westinghouse recommendations is provided as Enclosure 2.
(3)
The Mestinghouse advice is prescriptive on resetting of the high pressure injection system and incomplete as to the need for keeping the reactor coolant pumps running.
~l
Mr. John Tillinghast Indiana and Michigan Electric Co.
Indiana and Michigan Power Co.
(4)
Me are finalizing a generic report on TllI-2 matters related to Mestinghouse operating plants.
Although this report is not yet
- complete, among other things, we expect that it will recommend further analyses of transients and small reactor coolant system
- breaks, the development of appropriate written procedure guidance to operators in the use of these new procedures.
(5)
In certain instances, licensees are using fuel and relying on safety analyses, which were not provided by Westinghouse.
As a
result, it is not clear to us what the respective roles of the licensees, Mestinghouse, the fuel suppli'ers, and/or other parties should be in implementing those requirements described in item (4) above.
We need a clear and concise definition of their respective roles in these cases.
(6)
The Advisory Committee on Reactor Safeguards (ACRS) has issued five letters to the Commission as a result of their examination of the TMI-2 accident.
Me need a clear and concise position from all licensees with respect to each of the recommendations contained in these letters.
A summary of the ACRS recommendations is provided as Enclosure 3.
(7)
Individual licensees have indicated an interest in meeting directly with the staff regarding the Bulletin items for their facilities.
Experience to date has demonstrated that the staff does not have time to meet individually with each licensee to resolve these items.
It is clear that there are a significant number of technical issues yet to be resolved for a large number of Westinghouse operating plants.
There are limited resources available within the NRC staff to perform the necessary work.
This situation is exacerbated by the need to conduct similar and concurrent activities with those owners of BSW, C-E, and GE designed operating plants.
At the same time, there is a need to resolve these matters promptly.
'To resolve the issues described above in a prompt and expeditious
- manner, we believe there is a compelling need to establish an owner's group for Westinghouse operating plants.
Me expect that such a group would be needed for the remainder of calendar year 1979.
Owner's groups have worked effectively in the past in minimizing staff and industry resource requirements to resolve other generic problems.
We strongly urge you to meet with other owners of Westinghouse operating plants to consider the formation of such a group prior to our meeting on May 30.
This will be one of the principal agenda items at that meeting.
1 It
Mr. John Tillinghast Indiana and Michigan Electric Co.
Indiana and Michigan Power Co. Please note that investigation of a number of areas related to the TMI-2
- accident, including the long-term ACRS recommendations and long-term action items from NUREG-0560, will be specifically included as part of the future "Lessons Learned" staff activity.
You can expect additional correspondence in the future on these items.
If you require any clarification of the matters discussed in this letter please contact Patrick D. O'Reilly, the staff's assigned project manager for these activities on Westinghouse plants.
Mr. O'Reilly may be reached on (301) 492-7745.
Sincerely,
Enclosures:
As Stated A. Schwencer, Chief Operating Reactors Branch
$ 1 Division of Operating Reactors cc:
w/enclosures See next page
1~
Mr. John Tillinghast Indiana and Michigan Electric Company Indiana and Michigan Power Company cc:
.Mr. Robert W. Jurgensen Chief Nuclear Engineer American Electric Power Service Corporation 2 Broadway New York, New York 10004 Gerald Charnoff, Esquire
- Shaw, Pi ttman, Potts and Trowbridge 1800 M Street, N.M.
Mashington, O.
C.
20036 David Oinsmore Comey Executive Director Citizens for a Better Environment 59 East Van Buren Street Chicago, Illinois 60605 Maude Reston Palenske Memorial Library 500 Market Street St. Joseph, Michigan 49085 Mr. D. Shaller, Plant Manager Donald C.
Cook Nuclear Plant P. 0.
Box 458 Bridgman, Michigan 49106 Kenneth R. Baker 2874 Robin Hood Drive Stevensville, t1ichigan 49127
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oMan lootHo INDIANA& MICHIGANEIECTRIC COMPANY DONALDC. COOK NUCLEAR PLANT P.O. Box 458, Bridgman, Michigan 49106 (616) 465-5901 June 18, 1981 Mr. J.G.
- Keppler, Regional Director Office of Inspection and Enforcement United States Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Operating License DPR-74 Docket No. 50-316
Dear Mr. Keppler:
Enclosed please find a Licensee Event Report that exceeds the 10 day time limit reporting requirement.
On June 18, 1981, we advised your Senior Resident Inspector, Mr. E.R.
- Swanson, of this delay.
Sincerely, D.V. Shaller Plant Manager
/bab cc:
J.E.
Dolan R.S.
Hunter R.t<. Jurgensen R.F.
Kroeger K.J. Vehstedt E.
Swanson/N.
DuBry RO:III R.C. Callen MPSC G. Charnoff, Esq.
J.M. Hennigan ll. Lavallee EPRI PNSRC J.F. Stietzel E.L. Townley Dir., IE (30 copies)
Dir., MIPC (3 copies)
Jun: a2
~SBO
I C
'1
NRC FORM 366 U S NUCLEAR REGULATORY COMMISSION (7-77)
LICENSEE EVENT REPORT CONTROL BLOCK:
Q1 (PLEASE PRINT OR TYPE ALI.REQUIRED INFORMATION)
H I
6
~87N10CCZQ200000000 000Q341111QE~QE 7
8 9
LICENSEE'CODE 14 15 LICENSE NUMBER 25 26 LICENSE TYPE 30 57 CAT 58 CON'T'01 soURGE LJQ6 0
5 0
0 0
3 1
6 Q 0
6 0
4 8
1 QB 0
6 8
Q 7
8 60 61 DOCKET NUMBER 68 69 EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES Q10
~02 DURING NORMAL OPERATION HE S
~O3 SERVICE TO REDUCE SECONDARY SYSTEM CHEMICAL CONTAMINATION FROM CONDENSER INLEAKAGE.
ON JUNE 4 IT WAS DISCOVERED THAT THERE WAS NO FLOW THRU RAD MONITOR R-19 FROM ANY OF THE FOUR STEAM GENERATORS ~
THIS IS NON-CONSERVATIVE IN RESPECT TO APPENDIX "B"
~06
~O7 T.S ~
2 ~ 4.2 ~ G ~
THIS UNIT IS EXPERIENCING A SMALL PRIMARY TO SECONDARY LEAK.
THE TOTAL POSSIBLE TIME FOR THE UNMONITORED RELEASE WAS 35 MINUTES ~
PREVIOUS OCCURRENCE OF A SIMILAR NATURE INCLUDES 050-316 81-04.
~OB 80 7
8 7
8 SYSTEM CAUSE CAUSE COMP.
VALVE CODE CODE SUBCODE COMPONENT CODE SUBCODE SUBCODE C
Q11 X
Q12
~XGE I
N S
T R
0 QEE
~XQis LZJQ 9
10 11 12 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LFRIRP EVENT YEAR REPORT NO.
CODE TYPE NO.
Q17 RERoR1
~OI Q
~Z0 Qr
~04 QT Q
Qo 21 22 23 24 26 27 28 29 30 31 32 ACTION FUTURE EFFECT SHUTDOWN ATTACHMENT NPRDQ PRIME COMP.
COMPONENT TAKEN ACTION ON PLANT METHOD HOURS Q22 SUBMITTED FORM SUB.
SUPPLIER MANUFACTURER LXJQ18 ~FQ79
~ZQ25
~ZQ21 0
0 0
0
~YQ23
~YQ24
~NQ25 0
4 4
0 Q28 33 34 35 36 37 40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS Q27 o
HE LOSS OF FLOW FROM R-19 FR M ALL F F
CRUD AT THE ROTO-METER FLOW REGULATOR ~
CYCLI G
F THE F
W DISLODGED THE CRUD AND FLOW WAS RE-ESTABLISHED.
DESIGN CHANGE RFC 12-1825 WAS WRITTEN 3
TO PROVIDE TO CONTROL ROOM ANNUNCIATION DURING LOSS OF FLOW.
THE ATTACHED SUPPLEMENT EXPLAINS THE EVENT IN DETAIL~
7 8
9 FACILITY STATUS
% POWER OTHER STATUS Q 5
E Q28 ~10 0
Q29 NA 7
8 9
10 12 13 44 ACTIVITY CONTENT RELEASED OF RELEASE AMOUNTOF ACTIVITYQ 6
M Q33 M
Q34 SEE CAUSE DESCRIPTION 7
8 9
10 11 44 PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION Q39
[ll7] ~00 0
Q37 ~ZQ38 NA 7
8 9
11 12 13 PERSONNEL INJURIES
~oo" o""QEo "'VNA' 8
9 11 12 LOSS OF OR DAMAGETO FACILITYQ43 TYPE DESCRIPTION 9
Z Q42 NA
'7 8
9 10 PUBLICITY ISSUED DESCRIPTIONQ
~20
~NQ44 NA 7
8 9
10 NAME OF PREPARER R. A.
PALMER 80 80 80 80 80 80 NRC USE ONLY Es 80 0
68 69
- PHDNE, 616 465 5901 METHOD OF DISCOVERY DISCOVERY DESCRIPTION
~BQ31 CHEMICAL TECHNICIAN OBSERVATION 45 46 LOCATION OF RELEASE Q36 S.G.
BLOWDOWN STARTUP FLASH TANK TO ATMOS.
45
l I
IP
SUPPLEMENT TO RO 50-316 81-020 04T-0 During normal operation the steam generator startup blowdown flash tank was in service to reduce secondary system chemical contamination.
Flow through radiation monitor R-19 had been previously verified.
Routine technician surveillance at 1415 hrs noted the absence of flow.
By rapidly cycling the valves full open and shut~ the lines were blown out and flow reestablished immediately.
Leaving the area to notify control room of the situation would have resulted in a longer time period of unmonitored release.
Blowdown had been on the startup flash tank since 1340 hrs.
The maximum time the unmonitored release could have existed was 35 minutes.
Daily analysis to monitor the primary to secondary leak on Unit 2 had been completed and all parameters were within limits for release via the startup flash-tank.
A grab sample at the blowdown flash tank was taken at 1415 hrs and analyzed and showed secondary system activities had not significantly changed since the previous analysis.
Time
~Sam le Tritium pCi/cc Gross 8-pCi/cc)
Iodine-131 uCi/cc) 0015 0015 0015 0015 1415 f21 S/G 822 S/G 823 S/G 824 S/G 2-DSX-350.
(flash tank composite) 252x106 2.74 x 10 3.58 x 10 6 2.42 x 10 6 5.21 x 10-6 4.38 x 10 1.67 x 10 5
1.54 x 10 5
1.06 x 10 5
4.46 x 10 6
6.43 x 10 7
2.89 x 10 "
N.D.
2.75 x 10 7
5.81 x 10 7
Using the 1'415 steam generator blowdown flash tank composite analysis of the following releases were calculated:
Tritium Gross B-y Iodine-131 2.48 x 10 pCi/cc/sec 2.12 x 10 9gCi/cc/sec 2.77 x 10 1"pCi/cc/sec Release Rate 2.29 x 10 11 Ci/sec 1.96 x 10 11 Ci/sec 2.55 x 10 12 Ci/sec Total Release 4 81 x 10 8 Ci 4.11 x 10 8 Ci 5.36 x 10 9 Ci A design
- change, RFC-12-1825, is currently being engineered which will provide an immediate loss;of'flow warning to the Operations control room, whenever flow to R-19 is interrupted.
~
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/'Qleccrlc WnIAea & MICHIGANELECTRIC CDMPANY DONALDC. COOK NUCLEAR PLANT P.O. Box 468, Bridgman, Michigan 49106 (616) 465-6901 June 19, 1981 Mr. J.G.
- Keppler, Regional Director Office of Inspection and Enforcement United States Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Operating License DPR-74 Docket No. 50-316
Dear Mr. Keppler:
Pursuant to the requirements of the Appendix A Technical Specifications, the following report/s are submitted:
RO 81-021/03L-O.
Sincerely, D.V. Shaller Plant Manager
/bab cc:
J.E.
Dolan R.S.
Hunter R.ll. Jurgensen R.F.
Kroeger K.J. Vehstedt E.
Swanson/N.
DuBry RO:III R.C. Callen MPSC G. Charnoff, Esq.
J.M. Hennigan W. Lavallee EPRI PNSRC J.F. Stietzel E.L. Townley Dir., IE (30 copies)
Dir., MIPC (3 copies)
-"'*~2 1981
V l
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\\
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~
~
NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSION (7 77)
LICENSEE EVENT REPORT
~ CONTROL BLOCK:
QI
{PLEASE PRINT OR TYPE ALLREQUIRED INFORMATION)
I 6
~01 M
I 0
C C
2 Qs 0
0 0
0 Qs 4
1 1
1 I Q~ Q 7
8 9
LICENSEE CODE 14 15 LICENSE NUMBER 25 26 LICENSE TYPE 30 57 CAT 58 CON'T
~01
"'""'LQB 0
5 0
0 0
3 1
6 Q7 0
5 2
3 8
1 QB 0
1 9
8 I 0 7
8 60 61 DOCKET NUMBER 68 69 EVENT DATE 74
'75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES Q10 FOLLOWING A TRIP OF THE UNIT 2 REACTOR FEEDWATER ISOLATION VALVES FOR ¹1 AND ¹4
~03 STEAM GENERATOR
( FMO-201 AND FMO-204 ),
FAILED TO CLOSE UPON RECE IPT OF A FEE DWATER 4
ISOLATION SIGNAL.
THIS CONDITION WAS NON-CONSERVATIVE WITH REGARDS TO TECHNICAL SPECIFICATION 3.3.2.1.
~06
~OB 80 7
8 9
7 8
SYSTEM CAUSE CAUSE COMP.
VALVf CODE CODE SUBCODE COMPONENT CODE SUBCODE SUBCODE Q1>
[X~ QTs
~BQ13 0
K T
Q glQ"
~Z Qis 9
10 11 12 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LFRIRP EVENTYEAR REPORT NO.
CODE TYPE NO.
Q REPQRT ~91
~
~02 I
~w
~03
~L
~0 21 22 23 24 26 27 28 29 30 31 32 ACTION FUTURE EFFECT SHUTDOWN ATTACHMENT NPRDX PRIME COMP.
COMPONENT TAKEN ACTION ON PLANT METHOD HOURS Q22 SUBMITTED FORM SUB.
SUPPLIER MANUFACTURER LaJQlsLZJQls
<ZQss LZ II' O
O O
<VQss
~V@
~NQss N
I 2
O Qs 33 34 35 36 37 40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS 27 0
THE CAUSE FOR THE FAILURE OF THE FEEDWATER ISOLATION VALVES, FMO 201 and 204 WAS DUE TO THE MISALIGNMENT OF THE SWITCH ACTUATING ARM ON REACTOR BYPASS BREAKER A, CAUSING THE SWITCH TO STAY IN THE OPEN POSITION.
THIS SWITCH, IN SERIES WITH SWITCHES ON
~i 4
FEEDWATER ISOLATION'SEE ATTACHED SUPPLEMENT 7
8 9
FACILITY
~30 METHOD OF STATUS BB POWER OTHER STATUS ~
DISCOVERY DISCOVERY DESCRIPTION s
0 Qss
~QQ 0
Qss NA
~AQsl OPERATOR OBSERVATION 8
9 10 12 13 44 45 ACTIVITY CONTENT RELEASED OF RELEASE AMOUNTOF ACTIVITYQ LOCATION OF RELEASE Q 6
Z Q33 Z
Q34 NA NA 7
8 9
10 44 PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION
~00 o
Q>>~zQss NA 7
8 9
11 12 13 PERSONNEL INJURIES 45 3 REACTOR TRIP BREAKER A,SUPPLY THE "REACTOR ISTRIPPED" SIGNAL TO TRAIN ASSPS FOR 80 80 80 80 7
8 7
8 O
7 8
9 11 12 LOSS OF OR DAMAGETO FACILITYQ43 TYPE DESCRIPTION Z
Q42 NA 9
10 PUBLICITY ISSUED DESCRIPTION~
MNQ44 NA 9
10 MAIS/IE OF PREPARER R. A.
PALMER PHONE 80 68 69 616-465-5901 80 NRC USE ONI.Y ss 80 ss 0
I I
ATTACHMENT TO LER ¹ SUPPLEMENT TO CAUSE DESCRIPTION THE SWITCH LINKAGE WAS REALIGNED, AND FUNCTIONALLY TESTED AND THE SYSTEM RETURNED TO OPERATION.
NO FURTHER ACTION IS PLANNED.
'I
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