ML18219D097
| ML18219D097 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/13/1978 |
| From: | Tillinghast J Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC: 00077 | |
| Download: ML18219D097 (32) | |
Text
Q zq /stag REGULATORY INFORMATION DISTRIBUTION SYSTEM (BIDS)
DISTRIBUTION FOR INCOMING MATERIAL 50-3
/316 REC:
DENTON H R
NRC ORG:
TILLINGHAST J IN 5 MI PWR DOCDATE: 09/13/78 DATE RCVD: 09/14/78 DOCTYPE:
LETTER NOTARIZED:
YES COPIES RECEIVED
SUBJECT:
LTR 1
ENCL FORWARDING INFO CONCERNING MOVEMENT OF HEAVY LOADS OVLR SPENT FUEL AREAS AT SUBJECT FACILITY...WlATT DRAWINGS.
PLANT NAME: COOK UNIT 1
COOK UNIT 2 REVIEWER INITIAL:
XJM DISTRIBUTER INITIAL:
DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS NOTES:
I
- 8. E 3 CYS ALL MATERIAL GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION'ODE AOOi)'OR ACTION:
INTERNAL:
EXTERNAL:
BR
-F ORBOi BC~~W/7 ENCL G FI E W/ENCL W/Q ENCL HANAUER~~W/ENCL AD'OR SYS 0 PROJ++W/ENCL REACTOR SAFETY BR44W/ENCL EEB>+W/ENCL J
MCGOUGH++W/ENCL LPDR S ST.
JOSEPH'I++W/ENCL TERA4+W/ENCL NS IC++WlENCL ACRS CAT B+4W/16 ENCL g~~ b~5 NRC PDR+4W/ENCL OELD+>LTR ONLY CORE PERFORMANCE BR<+W/EhfCL ENGINEERING BR>+W/ENCL PLANT SYSTEMS BR++W/ENCL EFFLUENT TREAT SYS++W/ENCL DISTRIBUTION:
LTR 40 ENCL 39 SIZE:
2P+13P+4P CONTROL NBR:
780720288 T}fE END we++.>~we +xw++ww++w~wwwwwww++~k4A+c
tOq P
KQlL805 IN f<H RK Ill"t INDIANA IIr MICHIGAN POWER COMPANY P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 September 13, 1978 AEP:NRC:
00077 Donald C. Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos.
50-315 and 50-316 License Nos.
DPR-58 and DPR-74 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S.. tVuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Denton:
By letter dated August 25,
- 1978, AEP:NRC: 00017, AEPSC informed the Commission that, we would not. be able to supply the information requested by Mr. V. Stello, Jr. of your staff, 'concerning the movement of heavy loads over spent fuel areas at the Donald C. Cook Nuclear Plant, until September 12, 1978.
The enclosure to this letter contains the information requested by Mr. Stello.
Very truly yours, ice President Sworn and subscribed to before me this 'day of September 1978 in New York County, New York cc (attached)
Notary Public KrhxEIL~sP Br's&RY NOTARY iUPLIC, Strr'.o ci New York sio. 41.-1rs0o'r'82
'ust;lisd in queens Ccunty.
Certificaru hied in ttaw Yerk County Cc,nunasIcu wp>res rnurcts 3ii, l>~
7~07 i0288
I'arold R.. Denton cc:
R. C. Callen G'. Charnoff P.
W. Steketee R. J. Vollen R. Walsh R.
W Jurgensen D. V. Shaller Bridgman
~
~
ENCLOSURE QUESTION/REQUEST FOR INFORMATION 1.
Provide a diagram which'illustrates the physical relation between the reactor core, the fuel transfer canal, the spent fuel storage pool, and the set down, receiving or storage areas for any heavy loads moved on the refueling floor.
RESPONSE
1 Included,.
as Attachment No 1, to this submittal are four (4} design. drawings, two (2) for each unit of the Donald C. Cook Nuclear Plant, depicting the physical layouts requested;.
Please note that the spent fuel storage pool proper is a. shared, 'system't the Donald'.
Cook Nuclear Plant.
.QUESTION/REQUEST FOR INFOK'LOTION 2.
Provide a list of all objects that are required to be moved over the reactor core (du. ing refueling) or the spent fuel storage pool.
For 'each object listed, provide its approximate weight and siz, a diagram of the movement path utilized {including carrying height) and the frequency of movement.
RESPONSE
The following equipment, is required to be moved over the reactor core during. refueling '.
Oescri tion Approximate
'~the h'h.)'rimary Oimens i on 15 x 15 Spent Fuel Handling Tool*
Upper I'nternals.
Burnable Poison Rod Assembly
'andling Tool Upper Internals Guide Tube Cover Tool radiation Sample Hand'Ling Tool 15 x 15 New Fuel Handling Tool*
Rod Control Cluster Thimble Plug Handling Tool Shaft Tool*
17 x 17 Spent Fuel Handling Tool*
- 17 x 17 New Fuel Handling Tool * "
Manipulator Crane Fuel Assembly Polar Crane:
Trolly Polar Crane:
Large Hook and Block 397 116 000*/125 000+*
800*/634 '*
185
23Q 75 235 /270**
145 412 85 42,700 1,400 (max.)
240,000 7I 600 35 Ft.
38.0 Ft.
34,00 Ft 48.0 Ft 24.875" in.
445* in/43C** in 214 in.
35 ft 24.5 in.
12 Ft.
Polar Crane:
Small Hook and Block 1, 344
- Unit 1 Specific
- Unit 2 Specific
RESPONSE
Cont'd The specific movement paths of the objects listed above are dependent on the nature of the operation being performed.
Hence it is not possibe to supply exact movement paths as requested.
The specific object being removed from the reactor core is raised to a.height sufficient to assure that the reactor vessel has.been cleared.
- The object is then moved away from the reactor core and transported. to its 'final'estination as shown in the drawings included as attachment No.
1 to this submittal.
-2a-
QUESTION/REQUEST FOR INFORMATION 3.
What are the dimensions and weights of the spent fuel casks that are or will be used at your'facility?
RESPONSE
3.
The Cask Drop Protection System (CDPS)- to be installed at the Donald, C. Cook Nuclear Plant is designed to handle the largest, practicable spent fuel cask currently available namely the National Lead Industries, Inc.,
(NL),
No. 10/24 one hundred ton rail cask The nominal parameters for the NP No. 10/24 rai3 cask, and the corresponding CDPS design parameters.
are as follows=
Parameter NL No. 10/24 Cask CDPS Desi n Weight (Empty)
Length Diameter (Ovezall 179-,500 lbs 170 ft.
88: in.'00,000 lbs.
170 88 in.
The CDPS to be installed, at the'Donald C. Cook Nuclear Plant
's capable of handling different cask designs with one or more of the above parameters being smaller than the corresponding.
NL No 10/24 rail cask parameter valve.
The effects of a hypothetical, inadvertent cask drop are con-tained in our response ta FSAR Question. 14.15.
The analysis'=.
contained therein is based on the inadvertent dropping of a spent fuel cask, the dimensions of. which are given above under
'CDPS Design'.
These results are bounding (worst case) results for a hypothetical cask drop accident, and the effects of a cask drop accident with a smaller cask (one or more design parameters smaller than the CDPS design parameters),
would be correspondingly less.
QUESTION/REQUEST FOR INFORMATION
.4.
'Identify any heavy load or cask drop analyses performed to date for your facility.
Provide a copy of all such analyses not previously submitted to the NRC.
RESPONSE
The cask. drop analyses and questions pertaining to an inadvertent cask drop over the spent fuel pool can be found in our responses to FSAR Questions 14..15.1 through 14.15.21..
QUESTION/'REQUEST FOR INFORMATION 5.
Identify any heavy loads that are carried over equipment required 'for the safe shutdown of a plant that is operating
~ at the time the load is moved.
Identify what equipment could be affected in the event of a heavy load handling accident (piping, cabling, pumps, etc.)
and discuss the
.feasibility of such an accident affecting this equipment.
Describe the basis for your conclusions.
f
RESPONSE
5..
No heavy loads are carried over equipment required for the safe shutdown of either unit of the Donald' Cook Nuclear Plant
QUESTION/REQUEST FOR INFORMATION 6.
If heavy loads are required to be carried over the spent fuel storage pool or fuel transfer canal at your facility, discuss the feasibility of a handling accident which could result in water leakage severe enough to uncover the spent fuel..
Describe the basis for your conclusion.
'ESPONSE 1
. 6 No heavy loads are required to be carried over the spent fuel storage pool or the fuel transfer canal at the Donald C
Cook. Nuclear'lant.,
Ta Prevent damage to the spent fuel pool which would result in water release severe enough to uncover spent
- fuel,
- a. system of limit switches, interlocks and the necessary admininstrative controls will be installed..
When this system is finalized, movement of the spent fuel cask over. the spent. fuel will, be limited to the critical path as described in our responses to FSAR Question 14.15.4.
'ovement of the spent fuel cask, while lowering or raising the cask to the top of; the spent fuel pool.'
immediately adjacent'. to the outside edge of the pool wall, will be restricted to. a. vertical corridor directly above a knock-out section of the floor, which. is designed to give way in the event of a cask drop event, allowing the cask, to come to rest on the ground level without affecting the integrity of the spent fuel pool wall.
At present, any movement of the overhead (auxiliary building) crane over the spent fuel pool is under administrative control and is limited by means of limit switches and inter-locks to a portion of the end of the pool away from stored fuel assemblies.
ENCLOSURE QUESTION/REQUEST FOR INFORMATION 7.
Describe any design features of your facility which affect the potential for a heavy'oad handling accident involving spent fuel, e.g., utilization of a single failure-proof crane.
RESPONSE
7.
The auxiliary building crane will be interlocked using limit'witches and'. relay logic so that the crane hook can never pass over the spent fuel, cells and. can only pass over the east end of. the spent fuel pit proper during cask'. loading.
II Two zones of protection wiIL-be provided.
Zone 1, over the spent fuel cells, can never be entered by the crane hook and if an attempt is made to do so,. the crane: will. automatically shut. down.
Zone 2, over the local cask storage area can only be entered
'by the crane hook by by-passing a crane bridge limit switch
.located. on the crane runway rail just east of the local cask storage area..
To avoid unauthorized by-passing of this limit
- switch, a keyed by-pass control will be provided on the crane control cabinets located upon the crane walkway.
ENCLOSURE QUESTION/REQUEST FOR INFORMATION 8.
Provide copies of all procedures currently in
',facility for the movement of heavy loads over core during refueling, the spent fuel storage equipment required for the safe shutdown of a is operating at the time the move occurs.
effect at your the reactor pool, or plant that
RESPONSE
8.
No'eavy loads. are moved over the reactor core during refueling operations.'
Movement of heavy loads over the spent fuel storage pool is discussed in our. response to Question 6
of this submittal.
ENCLOSURE QUESTION/REQUEST'OR. XNFORMATlON
- 9. 'iscuss the degree to which your facility complies with the: eight. (8), regulatory positions delineated in Regulatory Guide 1.13 (Revision 1,'ecember, 1975)'egarding Spent Fuel. Storage Facility Design Basis.
RESPONSE
9.;.
Re lato
. Guide Position. No., I; The: spent'ueJ, storage facility. including structures',
fue3= racks, cask drop protection'ystem,,
and. cranes which traverse'ver: the'pool are Category. I'eismic. Design.
~,
R ato Guide Position No 2
The. spent fuel pit'is wholly contained: in the Category T
Seismicly.Designed Auxiliary Building which is designed to withstand., the-effects of tornado winds arid missiles; generated.
W.bj" these winds from entering the spent, fuel, pool.
Re ula'tor Guide. Position-No..
3.'he.
response. to this question is given in response to Questions No. 6'nd': No
-7 above..
- Re ulator Guide Position No 4'-
The auxiliary buiIding, which. encloses the spent fuel pool is designed'. to'contain. all liquid leakage which occurs in this building to-'he environs of this. structure prior to processing and to, normally maintain
- a. negative pressu're.
within; the building to control gaseous removal from the building.
The design of the ventilation and filtration systems meet Regulatory Guide 1.25 assumptions and the resulting radiation dose consequences from a fuel handling accident based on these assumptions are present in Chapter 14, Section
- 14. 21 of the FSAR.
Regulator Guide Position No.
5 The installation of the Cask Drop Protection System and restriction of cask movement to the critical path discussed in responses to Questions No 6 and No.
7 above in conjunction with our responses to FSAR Questions 14.15.1
- through 1'4.15.21 limits cask movement over the pool to an area that will not affect the stored spent, fuel or uncover it in the inadvertent event of a cask drop (Regulatory Po'sition C'nder Section C..5 in Regulatory Guide 1.13)
Re ulatoz Guide Position No.
6'mall diameter drains-located within the spent fuel pool stuctuze directly under the pool floor liner do not breach the. li;ner where installed solely to detect any small leakage through the 1'incr. All piping which enters the pool is terminated at an elevation approximately six (6) feet above the stored spent fuel. assemblies so that syphoning of water below this. elevation will not occur: should an inadvertent
- leak, occur at any point in this: piping (either in oz outside the pool) or in. the. systems connected to it.
Re ulatoz Guide Position No. 7 Two (2) water
'~ evel indicators are installed, in. the spent fuel pit which alarm both at a local control panel and in the control zoom in the event that the water level falls six (6), inches'elov.
nominal pool, level.
As mentioned in response to Regulatory Guide 1.13 Position. No.
6 above, the pool water level cannot be inadvertently drained below an elevation six (6) ft. above stored fuel assemblies due to a break in any piping.
A radiation monitor located in the exhaust path of air sweeping over the pool activates the mechanism to channel the exhaust. air entraining any radioactive gases through charcoal filters in the event of a High-Radiation-Level Alarm..
Re ulator Guide Position =No.
8 Since the spent fuel pit is wholly enclosed within the Auxiliary Building which is designed to withstand earthquakes, missil.es originating in high winds, 'and turbine missiles, and the Cask Drop Protection System@
controlled movement, of the spent fuel cask.,
and restriction of the A'uxiliary Building overhead'rane.
movement over the pool except along the critical path during spent fuel
- shipment, are the design, features to prevent lo'ss of water
'from the pool in the. inadvertent event of a heavy load; drop Hence,.
no. water'make-up system to add coolant is required ATTACHMENT TO AEP:NRC 00077
Imerltj
~~
"~
REGULATORY INFOR M I
N SYSTEM (RIBS >
DISTRIBUTION FOR INCOMING MATERIAL 50-316 REC:
DENTON H R NRC ORG:
MALONEY= G P IN 5 MI PWR DOCDATE: 09/ii/78 DATE RCVD: 09/19/78 DOCTYPE:
LETTER NOTARIZED:
YES COPIES RECEIVED
SUBJECT:
LTR 1
ENCL 1
FURNISHING SUPPLEMENT INFO TO APPLICANT LTRS OF 04/27/78 0 08/11/78 CONCERNING REQUEST TO APPENDIX "A" TECH SPEC CHANGE TO ALLOW UNRESTRICTED CONTAINMENT PURGE OPERATION IN BOTH THE UPPFR AND LONER VOLUMES O-SUBJECT FACILITY... NOTARIZED 09/1 1/78... W/AT PLANT NAME: COOK UNIT '2 NOTES:
1.
SEND 3 COPIES OF ALL MATERIAL TO ISE GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE AOOi)
REVIEWER INITIAL:
XJM DISTRIBUTOR INITIAL:
44>~4+44++44++4++
DISTRIBUTION OF Tl-IIS MATERIAL IS AS FOLLOWS FOR-ACTION:
INTERNAL:.
BR CHIEF ORBSi BC+wW/7 ENCL REG FILETS
/ENCL
/~ ENCL HANAUER44W/ENCL AD FOR SYS 5 PROJ~4N/ENCL REACTOR SAFETY BR++W/ENCL EEB~~W/ENCL J
MCGOUGH4+W/ENCL NRC PDR4+W/ENCL OELD+4LTR ONLY CORE PERFORMANCE BR+<W/ENCL ENGINEERING BR+>W/ENCL PLANT SYSTEMS BR<4W/ENCL EFFLUENT TREAT SYS>+W/ENCL EXTERNAL:
LPDR S ST.
JOSEPHr MI++M/ENCL TERA+4W/ENCL NSIC+>W/ENCL ACRS CAT B44W/16 ENCL DISTRIBUTION:
LTR 40 ENCL 39 SIZE:
2P+3P
%% %% 4 4.%08%%%%%%%%%%%%% %%%%%% %%%4 4%%%
CONTROL NBR:
780870338 THE END
%%4'4%%%%44%%%%%%%%%%%%%%%%%%%%
PM I
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INDIANA 5 MICHIGAN POWER COMPANY P. O. BOX 18 BOY/LING GREEN STATION NEW YORK, N. Y. 10004 September ll, 1978 AEP:NRC:
00082 Donald C. Cook Nuclear Plant Unit No.
2 Docket No. )0-3l6 License No.
DPR-74 Mr. i Harold R. Denton, Director Office o'f Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C.
20555 err r l
~I rn <>
r 'rr
~Qs Ic"
Dear Mr. Denton:
The purpose of this letter is to supplement our letters of April 27, l978 and August, ll, l978 concerning our request for an Appendix "A" Technical Specification change to allow unrestricted containment purge operation in both the upper and lower volumes of the Donald C. Cook Nuclear Plant Unit No. 2.
Attachment l to this letter provides the additional information rectuested by members of your staff to further, support our request to delete Unit 2 Technical Specification 3/4.6.l.7.
In short, we were requested to supply sensitivity analyses of the resistance coefficients for the elbows and debris screens of the containment purge system test valve, and provide the dependence on those coefficients of the re-sulting torque.
As can be seen from the results shown in Attachment, l, the various conclusions shown in our previous letters on the matter remain unchanged.
Very truly yours, P.
M one Vice Pr side t
, Sworn and subscribed to before me this I1~ day of in New York County, New York Wu Notary Pubis.c TIILIsE gJsI RY NOTARY PU8UC, Steto of New'ork cc:
(attached) gQ 4I.4606792 Q
I'I'ed in Ctuccns County, Cert'i'ceto filed in New York CoUnly
" 't0 197m Contiirission Expires church 3,
. e
,780870338
Harold R. Denton
~
~
AEP:NRC:
00082 cc:
G.
R.
P.
R.
R.
R.
D.
Charnoff C. Callen W. Steketee Walsh J. Vollen W. Jurgensen V. Shaller Bridgman
p
Attachment 1
D.C.
COOK NU0LEAR PLANT UNIT 2 DOCKET NO. 50-316 LICENSE NO. DPR-74 CONTAINMENT PURGE VALVES Attached are two tables comparing the effects of assuming high or low resistance coefficients for the original inlet debris screens and elbows on the containment purge system test at D. C.
Cook Unit ¹2.
The original resistance coefficients for the debris screens installed during the test and the elbows were 1.8 and 0.6, respectively.
These were used since they produced a more con-servative estimate of the maximum torque that the valves experienced during the 8.6 psig test.
The resistance coefficients used in the analysis are high and could be as low as 1.5 and 0.2 for the screen and elbows, respectively.
The four different combinations of high and low resistance coefficients for the screen and elbows are shown in the tables together with the resulting torques they would produce on the closing valve.
The highest estimate of the torque on the valve is obtained by using the lowest values for the resistance of the debris screen and elbows (Run ¹4).
The corresponding torque produced at 15 psig using the value for the new debris screen and the lower value of the resistance of the elbow is shown in Run ¹5.
D. C.
COOK NUCLEAR PLANT UNIT NO.
2 DOCKET NO.
50-316 LICENSE NO.
DPR-74 TABLE 1 RUN>>COMBINATIONS 1
2 3
5 6
Inlet Press.
PSIA 23 ~ 3 23 ~ 3 23 e3 23.3 29.7 29e7
. Inlet Screen 1.8 1.8 1.5 1.5 7.0 7,0 Elbow K
.6
~ 2
.6
~ 2
~ 2
.6 Outlet Flow K
1.0 1.0 1.0 1.0 leO 1.0 Outlet Elbow K
.6
~ 2.6
~ 2
~ 2
.6 TAB OF PROGRAM ESU Run g Inlet Condition PSIG Pipe Diameter Inch Flow SCFH Pl P2 P3 P4 P5 Inlet Resis-tance Outlet Resis-tance 1
2 43 5
6 8.6 8.6 8.6 8.6 15.
15.
30 II 30
'II 30 II 30 II 30 II 30 II 8.15x1066 8.8gxl06 8.35xl06 9.12xl06 8.45xl06 8'5xlo 23 ~ 3 23 '
23 ~ 3 23 ~ 3 29 7 29-7 19.43 19.46 19 73
'9.83 19.06 19.4.
18.86 18.77 19.14 19 1 18.44 18.84
- 18. 26 18.04 18'2 18.33 17 77 18.25 14 ~ 66 14..72 14.72 14 75 14.7 14.6 2.4.
2.0 F 1 1 ~ 7 7 '7.6 1.6 1.2 1.6 1.2 1 ~ 2 F 6
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D.C.
COOK NUCLEAR PLANT UNI t0.2 TABLE 2 DOCKET NO. 50-316 LICENSE NO. DPR-74 CALCULATED PRESSURES AND TOR UES CONTAINMENT PURGE VALVES 30" Upper Comp. Purge Supply - Containment Pressure 23.3 psia.
Run Opening-Deg.
Pin, psia
~Pact Opening Torque Closing Torque, in-lb.
in-lb.
90 90 90 90 70 18.86 18.77 19.14 19.1 19 57 w
w
.60
~ 73
.62
~ 77 1.66 3430 4177 3391 7115 2706 3415 2858 3617 6099;04) 30" Lower Comp. Purge Exh. - Containment Pressure 29.7 psia.
Outboard Valve only closing 5
90 18.44.
5*
70 19 09 6
90 18.84.
6*
70 19.53
.67 1 53 0 59 1 35 3111 5599'~
2706 4906 30" Lower Comp. Purge Exh; - Containment Pressure 29.7 psia.
Inboard Valve only closing 5
90 19 06 5*
70 19.67 6
90 19.4.
70 20+03
.62 1.46
.56 1 ~ 29 2858:,<<",
5330;.0'554 4677
- Balance with valve at 70'pening (Highest torque range)
il J
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P 4
4 I
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< RIDS)
DISTRIBUTION FGR INCOMING MATERIAL 50-316 REC:
DENTON I-I R NRC ORG:
MALONEY G P IN 8( MI PWR DOCDATE: 08/31/78 DATE RCVD: 09/07/78 DOCTYPE:
LETTER NQTARIZFD:, YES COPIES RECEIVED
SUBJECT:
LTR 1
ENCL 0 f.URNISHING ACKNOWLEDGMENT TO NRC" LTR OF 08/16/78 AND REQUEST AN ADDL THREE WEEK EXTENSION PERIOD UNTIL 10/04/78 FOR... NOTARIZED 03/31/78.
REVIEWER INITIAL:
XJM DISTRIBUTOR INITIAL:~
PLANT NAME: COOK UNIT 2 GENERAL DI TRIBUTIGN FQR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE A001>
h 4~~~~k>H~sHH~>'~sesl<+
DISTRIBUTION OF TI.IIS MATERIAL IS*AS FOLLOWS NOTES:
i.
SEND 3 COPIES GF ALL MATERIAL TO ISE FOR ~CTIO~:
INTERNAL:
EXTERNAL:
BR CIIIEF ORB81 BC>+LTR ONLY(7>
I LTR ONLY(i>
I h E~~~LTR ONLY<+)
~
IIANAUER++LTR ONLY(i)
AD FOR SYS 8c PROJ4~LTR ONLY( 1)
REACTOR SAFETY BR4~LTR ONLY(i>
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NSIC<~I TR ONLY(i)
ACRS CAT B++LTR ONLY(16>
NRC PDR++LTR ONLY(i)
GELD~~LTR ONLY<i)
CORF PERFORMANCE BR++LTR ONLY(
ENGINEERING BR+%LTR ONLY< 1)
PLANT SYSTEMS BR++LTR ONLY(i)
EFFLUENT TREAT SYS<<LTR ONLY<i DISTRIBUTION:
2P
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THE FND CONTROL NBR:
782490107 eh)@
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4
INDIANA ti MICHIGAN P. 0. BOX 18 BO WL I N G G R E EN ST A'6 ION NEW YORK, N. Y. 10004
- f. IlOP7 C3
<n
(
MK@IIII'<MlIKTI:II OWER COMPANY nv J'ugust 31,"lgT8 AEP:NRC:00081 Donald C.
Cook Nuclear Plant Unit No.
2
,Docket No.
50-316 License No.
DPR-74 Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commis. sion Washington, D.C..
20555
Dear Mr. Denton:
The purpose of this letter is to. acknowledge the receipt on August 24, 1978 of Mr.
N.
C. Moseley's letter dated August 16,
- 1978, and to request an additional three (3) week extension period to reply to the aforementioned letter.
As you are
- aware, AEP management is taking steps to im-prove operating activities" at the Donald C.
Cook Nuclear Plant.
- Indeed, Mr. John Tillinghast's letter of June 30, 1978 to Mr. J.
G. Keppler set forth some plans and policies to achieve this.
We are fully aware of the importance of the points brought forward by Mr. Moseley's letter and wish to provide a satis-factory answer..
Unfortunately, Mr. Tillinghast is outside the country at the present time and he is not scheduled to return until after the response to the letter is due.
Furthermore, he is personally interested in the development of the response to Mr. Moseley's letter.
~~2490i07
AEP:NRC:00081 In view of this status, we trust you will approve our request to extend the time for our response to October 4,
1978.
Very truly yours, GPM:em P,.
Maloney Vice President Sworn agd subscribed to before me this3~
day of August, 1978 in New York County, New York Notary Publ 1c DAVID G. HUME NO7ARY PUSLIC, State of New York No. 31-4BOB113 Quafified in New York County Commission Expires Merch 30, 1979 cc:
R P
G R
R 0
C.
W.
Cha Wal W.
V.
Cal 1 en Steketee rnoff sh Jurgensen Shaller-Bridgman