ML20054G622
| ML20054G622 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/16/1982 |
| From: | Fay A WISCONSIN ELECTRIC POWER CO. |
| To: | Clark R, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR TAC-07727, TAC-08074, TAC-7727, TAC-8074, NUDOCS 8206220143 | |
| Download: ML20054G622 (6) | |
Text
WISCONSIN Electnc a coum 231 W. MICHlGAN, P.O. BOX 2046. MILWAUKEE, WI 5320s June 16, 1982 Mr.
H.
R.
Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555 Attention:
Mr.
R. A. Clark, Chief Operating Reactors Branch 3 Gentlemen:
DOCKET NOS. 50-266 AND 50-301 SUBMITTAL OF OUTSTANDING INFORMATION NUREG-0612, CONTROL OF HEAVY LOADS POINT BEACll NUCLEAR PLANT, UNITS 1 AND 2 Your letters dated December 22, 1980 and February 3, 1981 requested that Wisconsin Electric Power Company review the handling of heavy loads at the Point Beach Nuclear Plant and provide information as requested in Enclosure 2 to the December 22 letter.
Our transmittals of September 30, 1981 and January ll, 1982 submitted our six and nine-month responses, respectively, which included the majority of the information requested in your letters.
Our February 25, 1982 letter provided a proposed schedule for the completion of those outstanding information items.
Enclosed for your review is Wisconsin Electric's response to NRC question Attachment 1-5,
" Interfacing Lift Point Evaluation".
This information is provided in the form of revised pages 7 and 21 and a new Appendix D,
" Interface Lifting Lug Analysis Summary", for inclusion in our nine-month response.
We are unable at this time to submit our response to NRC question 2.1.3.D,
" Evaluation of Lifting Rig Compliance with ANSI N14.6-1978".
The analysis necessary to supply this information is not yet complete.
It is our intent to complete and submit this information by July 12, 1982.
D33 8206220143 820616 PDR ADOCK 05000266 P
Mr. H. R. Denton June 16, 1982 In addition, we are hereby modifying the scheduled submittal date for our response to NRC question 2.3.4.b,
" Reactor Vessel Head Drop Analysis".
This analysis is being performed for Wisconsin Electric by Westinghouse Electric Corporation.
Westinghouse progress to date indicates that November 15, 1982 is a more realistic submittal date for this information.
We will continue to keep you apprised of this situation.
Please contact us if you.have any questions.
Very truly yours, hY Assistant Vice President C. W. Fay Enclosure Copy to NRC Resident Inspector Subscribed and sworn to before me this /GJ(, day of June 1982.
Y't h b _ ~ -_,
Notary PuMZic, State of Wisconsin
/
g,,/ 4,KV..
My Commission expires
RESPONSE TO HRC REQUEST FOR I!1 FORMATION 011 CONTROL OF IIEAW LOADS NIllE MOIITII REPORT FOR THE POIllT BEACII NUCLEAR PLANT UNITS 1& 2 WISCOt1 sit 1 ELECTRIC POWER COMPANY Prepared by Bechtel Power Corporation San Francisco Revision 1 Cali fornia 94119 May 1982 112/6 s
Rcv. l
+
The spent f uel shipping cask lif t points evaluation will be deferred until a shipping cask that is licensed is chosen for l
use at the Point Beach Nuclear Plant.
No shipping cask move-I ment over the spent fuel or safe shutdown equipment will be l
permitted until the evaluation is completed and compliance with UUREG-0612, Section 5.1.6(3) or its equivalent is con-firmed or justified.
Modifications, if required, will be completed prior to cask use.
An evaluation of the lugs for the concrete hatch covers, the large and small filter cask, the resin cask and watergate has l
been performed and is summarized in Appendix D.
I 2.4 NRC Ouestion 2.2-4 For cranes identified in 2.2-1, above, not categorized accord-ing to 2.2-3, demonstrate that the criteria of NUREG 0612, Section 5.1, are satisfied.
Compliance with Criterion IV will be demonstrated in response to Section 2.4 of this re-quest.
With respect to Criteria I through III, provide a discussion of your evaluation of crane operation in the spent fuel area and your determination of compliance.
Response
The spent fuel pool crane was identified in 2.2-1 above and was not categorized according to 2.2-3.
As stated in the response to 2.2-2, this device carries spent fuel elements which weigh less than the defined heavy load of 1750 lbs. and therefore is excluded from further consideration.
109/11.
_-.__=
5.
Rev. 1 i
i 1
I i
t APPE!!DIC ES Appendix A -
Load Drop Analysis of Unit 2, B Reactor Coolar>t Pump Flywheel
' Appendix B -
Ioad Drop Analysis of 17,000 lb. Main Feed Pu.np Motor in the Control Building.
Appendix C -
Technical, Specification 15.3.8 Refueling'and Spent Fuel Assembly Storage j!
I Appendix D -
Interface Lifting Lug Analysis Summary 4
4 h
s h
l J
i j
109/11 '
t k
Rev. -l a
APPEllDIX'D Interface lif t points were evaluated in accordance to NUREG-0612, Section 5.1.6 and the results are tabulated below in Table Dl.
TABLE D1 I
I I
I EQUIPMEllT l WEIGIIT l
MATERIAL /
- l. REQUIRED l CALCULATED l
l UTS (KSI) l SAFETY l
SAFETY l
l l
FACTOR l
FACTOR I
I I
I I
I I
I Concrete !!atch l
6,250 l
A-36/58 l
10 l
11.52 Covers l
l l
l l
l 1
l Large Filter Cask l 7,700 l
A-36/58 l
10 l
11.30 l
l l
l Small Filter Cask l 4,000 l
A-36/58 l
10 l
21.75 l
I I
I Resin Cask l 96,000 l
A240 Type 3041 10 l
6.53 l
l
/75 l
. l (Note 1)
I I
I l
Spent-Fuel Pool l
6,000 l
A-36/58 l
10 l
14.50 Gate (Watergate) l l
l l
l l
l
' ~
110 t e s :
(1)
The Resin Cask lif ting lugs will be upgraded by reinforcing with a 5/8" stainless steel plate.
This reinforcement will increase the calculated safety factor to 10.04, which exceeds the required safety factor per tlUREG 0612.
L k
i l
136/51
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