ML20032A551
| ML20032A551 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/26/1981 |
| From: | Hukill H METROPOLITAN EDISON CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR L1L-280, LIL-280, NUDOCS 8110300334 | |
| Download: ML20032A551 (5) | |
Text
Metropolitan Edison Company Post Of fice Box 480 I l.
L-Middletown, Pennsylvania 17057 Writer's Direct Dial Number October 26, 1981 L1L 280 O
h Office of Nuclear Reactor Regulation
\\
f Attn:
D. C. Eisenhut, Director 6,.
Nj(Ijjg )
y Division of Licensing f
U. S. Nuclear Regulatory Commission
,F' 0 C T 2 9 193y m 9 Washington, D.C.
20555
" gajamurou
Dear Sir:
z g,-
Three Mile Island Nuclear Station, Unit 1 (TF -O
{>,
Operating License No. DPR-50 va Docket No. 50-289 Control of Heavy Loads Enclosed please find our response to Sections 2.2, 2.3 and 2.4 to your letter of June 26, 1980, concerning the handling and contcol of heavy loads. This letter supplements our response of February 17, 1981 (TLL 474)
Sincerely, b.D. lukill Director, TMI-l HDH:LWH:vj f cc:
R. C. Haynes R. Jacobs J. F. Stolz L. Barrett
/
6 go3ih 8110300334 811026 DR ADOCK 05000280 PDR roeuupu,,ian Edison Company is a Member of the General Pubhc Utilities System g
.q 2.2i SPECIFIC l REQUIREMENTS ' FOR OVERHEAD' HANDLING SYSTEMS. 0PERATING : IN THE. VICINITY OF FUEL STORAGE POOLS ~
.NUREGLO612, Sectien 5.1.2, provides guidelines :concerning the ' des i gn and operation of7 load-handiing1 systems zin the vicinity.of stored, spent fuel..
c-
-Information provided11n. response ~to this section shou t d. demonstrate that adequate measures-have.been-taken to ensure that in this area, eitherEthe a likelihood of.a load drop which might, damage spent fue'l. is extremeIy smalI,. or
~
that the : estimated consequences' of such.a drop wii1 Lnot exceed the Iimits -set by..
the evaluation criteria of NUREG 0612, Sectic.4 5.1, Criteria'l.through lil.
Item ~2.2-1.
Identify by.name, type, capacity,.and equipment designator, any'
-cranes' physically-capable (i.e., _ ignoring. interlocks,. moveable.
mechanical stops, or operating procedures) of. carrying loads which-could,. if - dropped, -land or - f al 1.into 'the spent fuel pool.
Response?
Fuel Handling Buiiding 110. ton overhead crane.
. See Section 9.7.of
(
TMi.1 FS AR f or-deta il s).
Fuel Handling Bridges (3 ton /5 ton) as documented ~in our February 17, 1981 (TLL 474) response.
These bridges are limited in use (i.e.-1 fuel assembly lift) and therefore not considered.
ltem 2.2-2 Justify the exclusion of any cranes in this area.from the above category by verifying that they are incapable of carrying heavy loads or are permanently prevented f rom movement of -the hook centerline closer than 15 feet to the pool. boundary, or by pro-viding -a suitable analysis demonstrating that for any f ailure
~ mode, no heavy load can f al I-into the fueI-storage pooi.
Resoonse:
Tech.' Spec. 3.11 " Handling of irradiated Fuel" and TDR-142 " Cask Drop Analysis
'for Fuel Handling Building" both of which were included as Attachment 1 and 2 in the February 17 Met-Ed/GPU letter to the NRC, demorstrate that the 110 ton overnead crane need not be considered based on alternate but equivalent methods since procedu'res have b e er, imposed' limiting the movement of activated fuel, interlocks have oeen added te limit.the travel area of the crane, and the probability of a load drop tnat might damage spent fuei is extremely smal'.
Item 2.2-3 Identify any cranes listec in 2.2-1, above, which you have evaluated as having sufficient design features to make the likelihood of a load drop extremely small for all loads to be carried and the basis for this evaluation (i.e., complete compliance with NUREG 0612, Section 5.1.6 or partial com-pliance supplemented by suitable alternative or additional design features). For each crane so evaluated, provide the load-handling-system (i.e., crane-load-combination) Infor-mation specif ied in Attachment 1.
Response
The Fuel Handling Building 110 ton crane is categorized in 2.2.4.
Iterc 2.2-4 For cranes identified in 2.2-1, above, not categorized ac '
s cording to 2.2-3, demonstrate that the criteria of NUREG 5
0612, Section 5.1, are satisfied...
5
Response
This information is contained in Tech.: Spec 3-11 and TDR-1.42 (attachment 1 and g
2 respectively) (See also FSAR Supplement 2, Part VII.of the MET-Ed/GPU letter g
uto the NRC'of_ February 17,;1981 (Ti.L 474).
2.3 SPE( 'FIC.. REQUIREMENTS OF OVERHEAD HANDLING SYSTEMS OPERATING IN THE CONTAINIC.NT
'HUREG 0612, Section'5.1.3, provides guidelines concerning the design and operation of-loa d-handl ing systems-in the vicinity of the reactor. core.
informati_on provided in response to this section should be. suf f ici ent. to demonstrate that adequate measures have been taken to ensure that in this area, either the likelihoot of a load - drop which might damage spent fuel is_ extremely smal1, or that the estimated consequences of such a drop' wiii not exceed the limits set by the evaluation criteria of NUREG 0612, Section 5.1, Criteria i though ill.
Item 2.3-1.
identify by name, type, capacity, and equipment designator, any cran _es physcially capable (i.e., taking no credit for any inter-locks or operating procedures) of carrying heavy loads over the reactor vessel.
Response
Reactor BuiIding 185 Ton PoIar Crane.
Item 2.3-2Justify the exclusion of any cranes in this area f rom the above category by verifying that they 'are incapable of carrying heavy loads, or are permanently prevented from the movement of any load either directly over the reactor vessel or to such a loca-tion where in the event of any load-handling-system f ailure, the load may land in or on the reactor vessel.
Response
The Reactor Building 185 Ton Polar Crane is the 2nly applicable crane and is limited as discused in response to item 2.1.1 and :.1.1(b) of our February 17, 1981 letter.
Item 2.3-3.
Identify any cranes Iisted in 2.3-1, above, which you have evaluated as having suf ficient _ design features to make the likelihood of a load drop extremely small f or al l loads to be carried and the basis for this evaluation (i.e., complete comp Iiance wIth NUREG 0612, Sect ton 5.1.6, or pertial comp-liance supplemented by suitable siternative or additional design' features). For each crane so evaluated, provide the load-handting-system'(i.e., crane-load-combination) infor-
~
- /
4 mation 'speci fled 'in' Attachment l'.
Response
'The Reactor Building 185. Ton Polar Crane. is categorized in '2.3.4L beiow.
Jitem2.3-4.
For. cranes::I dentified in 2.3-1, above, notfcategorized ac -
. cording to 2.3-3, demonstrat'e that the evaluation' criteria of. NUREG 10612, -Section 5.1, are satisf ied. Compliance with-Ctiterion IV will'be-demonstrated in'your-. response to Section>
2.4;of this' request. With' respect to' Criteria 1. through lil,
pro.ide a discussion ~of your.evaluetion~of~ crane operation inL the' contaiment 'and your determination of compliance. This '
response should include the following information for each
~
. crane:
Where reliance is placed on the-installation.and use a.
of electrical interlocks or mechanical stops, indi -
cate the circumstances under wnich ~.t'ese protective devices can-be removed.ck-bypassed nd the'admints-trative procedures invokea to ensure proper authori--
zation of such action. Discuss any. elatec or proposed technical specification concerning -the bypassing of.
such interlocks.
Response
Our analysis. places no reliance on these factors (i.e., electrical interlocks or
. mechanical stops.
b.
Where reliance is placed on other, site-specific con-siderations (e.g., refueling sequencing), provide present or proposed technical specifications and dis-cuss administrative or physical controls provided to ensure the continued validity of such considerations.
Response
I
. Our. anal ysis p laces no reliance ~ on site specific considerations, tech specs or administrati.- and physical contro l s.
c.
Analyses performed to demonstrate compliance with Criteria i through til should conform with the guide-lines of Attachment 5.
Justify any exception taken to these guidelines, and provide the specific information requested in Atte:hment 2, 3, or 4, as appropriate, for each analysis performed.
. Response:
Our analysis satisfies this section by using an alternate but logical engineering evaluation similar to the approach proposed in Attachment (5) in of the ' June 26,1980 NRC letter by:
- Examining the' worst' credi t'le - case drops.(assuming 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after' reactor
. shutdown f or ref uel ing) near or on tne Reactor Vessel invoising both_tne Reactor Head '(155 ton) 1and tne upper Internals (P.lenum) assembly.
- In. no case would a crop result;in over stress or water leakaae throughout the entire Reactor Vessel.
Therefore, this precluces any 'potentially significant
. rel ease' of contamination / radiation + rom the Reactor Vessel.
We base-this calculation on energy balance theory that concluces that the maximum impact forces are less than tne critical buckling load.
- We Take no specific exceptions to~ Attachment 5 guicelinas.
However, we believe that tne rest acoroariate manner of dealing with the i. ore extensive /
specific information request of your Attachments 2, 3, and 4 is through a
- peneric approach by tne S&W Owner's Group. -We are currently stucying a proposal S&W on this subject.
4 SPECIFIC REQUIREMENTS FOR OVERHEAD HANDLING SYSTEMS OPERATING IN PLANT AREAE CONTAINING EQUIPMENT REQUIRED FOR REACTOR SHUTDOWN, CORE DECAY i
HEAT REMOVAL, OR 53ENT FLEL POOL COOLING
Response
i
- 1. Ir conjunction with the information presented in tne February 17, l'981 NRC-te -Ec/GDUN submittal, tne-following information applies:
Fo- -he Fue l Handl inc Sui ld ing 110 Ton ove-head crane, the information in response.To 2.2(1) (of: tnis attacnment) ape!!es.
- c-the Reactor Buildinc 185 ton Pola-Crane, tne only.
cotential area of concern, is cecay heat removal (DHR).
Close analysis reveals tnat no single loac crop woulc nave tne capabi l ity to a versely af f ect tne ab i l ity to mainTai n the Reactor in a safe shutcown condition.
Lpecific-informa-tior app l i cab le is as follows:
DHR Takes a suction from in (RC outlet) and discherges back to Reector Vessel inrough Core looding injection noccles (2 noccles - 180 (ciametrically) opposec on ococsite sices of~ tne Reactor core.
I Assuming refueling operations (i.e. hea: Removal) can commenca ne sooner than 72 nours af ter Reactor Shutcown.
The design basis for DHR:
OHR can main ain RC tem; I
s 140 F or belo. 10 nrs. a'ter shutcowr @ 1 (of'2)
DHR pumps 6 3000 gpm throug-i tot 2) ce:ay neat emoval cociers i 3000 ger cocling.arer recai e:
(or less) cepencing on co e nistcry*
efi cesigr 'io.
rates-are 6000 gom).
It has been' deduced from the above information that the single failure proof concept is fulfilled, because in no case, can one sincie load c cc adversely af f ect tne acility to obtain o mairtain Safe Reac cr snutcown or decay heat removal
. capability.
- 800 full power. days of operation.