ML17194A953
| ML17194A953 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/03/1981 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML17194A952 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, TASK-81-07, TASK-A-36, TASK-GL, TASK-OR TAC-07727, TAC-08074, TAC-10993, TAC-46747, TAC-7727, TAC-8074, NUDOCS 8103120056 | |
| Download: ML17194A953 (6) | |
Text
' !.
UNITEO STATES NUCLEAR REGULATORY COMMISSIOI\\
WASHINGTON. D. C. 20555 February 3, 1981
- TO.fl.LL LICENSEES OF OPERATING PL.A.NTS.C\\ND APPLIC~NTS FOR OPER.C\\TING LICENSES AND HOLDERS OF CONSTRUCTION PER~ITS*
SUBJECT:
CONTROL OF HEAVY LOADS (Generi~ Letter 81-07)
Gentlemen:
By our letter d1tted December 22, 1980, you were requested to review your controls of the handling of heavy loads to determine the extent to which the guidelines of NUREG-0612 are presently satisfied at your facility and to identify the changes and modifications that.would be required in order
. to fully satisfy these guidelines.'
- To exoedite your review, three enclosures were included with the letter.
One of the enclosures was Request for Additional Information on Control of Heavy Loads (Enclosure 3).
He have found that five oaqes from were missing due to a reproduction error.
The missing pages 2re enclosed with this letter.
In addition the December 22, 1980, letter on Page 2 in Item l required that information identified in Section 2.1 throuqh 2.4 of Enclosure 3 be included in a report documenting the results of your review.
This requirement should be modified to read:
"Sections 2.1 through 2.4 for PWR olants and Sections 2.1 through 2_.3 for Bl~P plants.
11 Because of these errors we are extending the Enclosure 2 90-day implementation recuirement to May 15, 1981.
Enclosure:
"Enclos.ure 3 11 missing oages Sincerely,
- With the exception of licensees for Indian Point 2 and 3, Zion *l and 2 and Three Mile Island l
- )
.(
-\\
Attachment (4)
A~~LYSIS OF PLANT STRUCTURES The follo..,dng information shoul'cf be provided for analyses conducted to demon-*
strate co:pliance 'IJith Criteria III and IV of NUREG 0612. Section 5.1.
- 1.
INITIAL COS'DITIONS/ASSL"M?TIONS Discuss the assumptions used in the analysis, including:
- a. *weight of hea~~* load
- b.
I=pact area of load
- c.
Drop height
- d.
Drop location
- e.
AssU!:!ptio~s regarding crecit taken in the analysis for the action of inpact li~iters
- f.
Tnick~ess o: ~alls or floor :labs impacted
- g.
Ass~ptions regarding crag forces caused by th*
enviro:u:nent
- h.
Load combinations considered
- i.
~.aterial properties of steel and concrete
- 2.
~::::::rr.o~ o: P~\\~LYSIS
?rovide the :c:;ethod of ar-.al;sis used to de:nonstrate* that sufficie.-.t loecl-carrying capability exists ~ithin the ~all(s) or floor slab(s).
Identi:y any co::puter codes ~ployed, and provide a description of their capabilities.
If test data was eoployec, provide it and describe its applicability.
- 3.
CONCLUSION Provide an evaluation co:c:;paring the results cf this analysis -~th Criteria III and IV of K1JREG 0612, Section 5.1.
~'here safe-shutdo~-n equipment has a ceiling or 'IJall se?arating it fro~*an overhead handling system, provide a~ evaluatio~ to de~onstrate that postulated load dro?S do not penetrate t:-ie *ceiling or ca~s.e secondary missiles that could pr.:vent a safe-shutCOlo.'n s::;;te::i fro=: perfor.:.ing its safety function.
(3)
A description of any Engineered Safety Feature filter system which includes infor-mation sufficient to demonstrate compliance with the guidelines of USNRC Regulatory Guide 1.52, "Design, Testing, and Maintenance Criteria for Engineered Safety Feature Atmos-phere Cleanup System Air Filtration and Absorption Units of light-~ater-Cooled Nuclear Power Plants."
(4)
A discussion of any initial conditions (e.g., manual valves locked shut, containment airlocks or equipment hatches shut) necessary to ensure that releases will be terminated or mitigated upon Engineered Safety Feature actuation and the measures en:ployed (i.e., Tech-nical Specification and administrative controls) to ensure that these initial conditions are satisfied and that Engineered Safety Feature systems are operable prior to the load lift.
- 2.
~.ETHOD o: ANALYSIS Discuss the method of analysis usec to demonstrate that post-accident dose
~ill be ~ell ~ithin 10CFRJ.00 limits.
In prP.senting methodology used in determining the radiological consequences, the f ollo~ing information should be provided.
- a.
A description of the mathematical or physical model e~ployed.
- b.
An identificr.tion and sulI!lLar.y of any cou:puter program used in this analysis.
- c.
The consideration of uncertainties in calculational
~ethods, equipment perfor~ance, in&trumentation response characteristics, or other indeterminate effects taken into account in the evaluation of the results.
- 3.
CONCLVSIO~
Provide an evaluation comparing the results of the analysis to Criterion I of !-H.TREG 0612, Section 5.1.
If the postulated heavy-load-drop accident analyzed bounds other postulated heavy-load drops, a list of these bounded heavy loads should be provided.
2-2
bounds other postulated heavy-load drops, a list of these bounded heavy loads should be provided.
3-2
Attachment (5) 3 of Q f
SHIELDED SHIPPING Ct..S l~S CERTIFICATED
-~
FOP. NUCLEAR P OtJER PLANTS II - t.:aste CROSS LOT IN CE:'.T.
MOJEL PRIY..;...RY LICENSEE us. (APPROX.)
SECONDARY L!CE?\\SE 67L.4 Poly Tiger 1uclear Engineering Co.
35,000
>.PL, BEC, _CPC, DLP, MEC, ?-."PP' s~~. VIP 67il SJ\\-1
- Nuclear Engineering Co.
60,000 APL, CPC,.DLP, KPP, SMU, VEP 9074 AP-100 28,000 DLC 9079 8'-100 Ser. 2 Hittman ~"uclear and 96,000 APL, BGE, CEC, a.=:,
De\\'elop::Dent Corp.
DLP, ll!!:,
- "'P
... I.. ' M'iA, MEC, J-."?-::.
--. FEC 90SO E;-600 Hit ti:.an Nuclear and 42,000 P.GE, C'.. "!:, c::c, !)T '::>
~.
Developt!len: Corp.
- u-.-:--*
IE!., JCP, \\.!'\\' !
}'!Ef"
~wt h7!'. F'!:C, YAC
- ~t*
- -100 Ser. 1 Eitt~n Nuclear and 45,000 APL, BGE, C'io."E.
D~ 'P -.
Develop:;:.c..-;t Corp..
D>i-' JCP,
?-~:'.A, }'"l:'i
-~*
K??-,
?-..TN~' ?:t.C' ii.GE,
\\""YC 9CSS FS-1005 Hi ttoan l~uclear an cl 36,500 BGE, C'... "!:. CEC, 1'V"":"
Develop:;e:::it Corp.
JCP, MYA, l:l'P
- P:t.C 9092 ES-300 Eitt::.an Nuclear anci 1.3,000 MYA Develop::;ent Corp.
9093 EN-400 Bittman Nuclear and 43,000 MYA Develop:;e:lt Corp.
9094 C\\SI-14-19.5-H Che~Nuclear Syste::s, 56,500 APC9 APL,
~~~
--I... C?L, Inc.
(:\\.. "'E. CY.A* c::c, CPC, DPC, FPL, F?C, GPC, JCP, ~c. }\\"'Vl>
1'~E, "NSP, OPP, ?GE, ?EC, PGC, PN'"J, FEG, T\\'A,
\\"LP S::l96 c~;s J-21-.300 Chei::-N...,clear Syste::is, 57,450 AFC, A?L, C?L, c::::'
lnc*
DPC, F?l, r?C, GPC, JCP, ~EC, '"""':>
- "* - ' t-.:::::,
}>!;)'. PEG, \\"EP
- se£: at ta.: ~.f d l is:
of c:~::i:-c-*.-i=: ::~:-:s.
. CERT.
~*:;~;:;,.
5971".
GE-200
.5980 GE~6::JO 6275 u-2e-4 9081
~S-1600 SHl HOED SHlPPif\\G CASKS CERTIFICATED FOR NUC~EA?. POlt.'ER PLANTS I I I - Byproducts PRn'A.llY LlCENS!E Che?:?-Nuclear Syste:s, lnc.
Che~-Nucle.c.r Syste~,
!DC.
GROSS LOT IN LBS. (APPROX.)
10,000 18,500 30,000
.4,6.000 At1:achment (5)
S of 6 SECO~"D>-~Y llC~SEE
- PEC NNE, NSP **
~c. CPL, DPC. FPL, FPC, NPP, VD>
AFC, BGE, CPL,, DPC, FPL, FPC, GPC,, NSP, TVA, \\TEP
- See attached list of abbreviatio~s.