ML19220A910
| ML19220A910 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/07/1975 |
| From: | Arnold R Metropolitan Edison Co |
| To: | Anthony Giambusso Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904250085 | |
| Download: ML19220A910 (10) | |
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- m.x.m METROPOLITAN EDISON COMPANY PCSr CF*tCi sox 542 RE A OING. PE!.NSYLVANI A 19603 TE LEPHCNE 2?5 - 9 3 0601 JAN 7 1975 Mr. A. Giambusso Deputy Directer for Operating Reactors Directorate of Licensing Office of Regulation U. S. Atomic Energy Consission f-
"].g)
Washington, D.C.
20545
Dear Mr. Giambusso:
SU3 JECT: THREE MILE ISLAND NUCLEAR STATION - UNIT 2 DOCKET NO. 50-320 TEMPORARY POLAR CRANE On Nosenber 18, 1974, we advised you that we would provide infor=ation regarding the temporary polar crane embed =ents at Three Mile Island Nuclear Station Unit 2 (IMI-2) by January 7, 1975. This coesit=ent was made in accordance with the verbal agreements made at a meeting on this topic on October 21, 1974. The attached documents provide that information in
- vo parts.
1.
Analysis of the Concrete Used in the Fuel Handling Pool Supportinr; Walls 2.
Structural Analysis of the "as-built" Fuel Handling Pool Supporting Walls.
Very truly yours j% q, AW(-(
/73.',
R. C. Amold Vice President RCA/1h Attachments 72 tC81 790425 cogf b
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ANALYSIS OF THE CONCRETE USED IN THE FUEL HANDLING POOL SUPPCRTING L' ALLS The Reactor Building Fuel pool supporting structure was placed Lato plac aent from elevation 282 ft. to 294 ft. 6 in, and from 294 ft. 6 in.
to 305 ft.
Place =ent records indicate that ec solidation in these placements was satisfactory. Libration was achieved with five 4 in. diameter and one 3 in. dia=eter air-operated vibrators for the first place =ent and six 3 in.
dia=eter vibrators for the second place =ent.
Lif ts within the placement were limited to 18 in.
Siu=ps ranged from 1-3/4 in. to 3h in. for these place =ents.
Concrete test cylinder strength at 28 day averages ranged froc 5490 psi to 6750 psi.
The fuel pool slab from elevation 305 ft. to elevation 308 ft. was
=ade in two 18 in.lif ts.
Place =ent records indicated consolidation was fair with a concrete slump of 2-3/4 in.
Four 2 in. diameter vibrators were used. Ccucrete test cylinders indicated a 28 day average strengen of 6440 psi.
The fuel poel wal?.s were =ade in six placements with construction joints at 51evations 308 ft., 318 ft. 6 in., 322 f t., 335 ft., 347 ft. 6 in., 360 f t.
6 in.
The top of the aall is at elevation 367 ft. 6 in. Placement records indicate tnat the consolidation ranged from satisfactory to good and that slumps ranged from 2-3/4 in. to 3b in.
At least six 3 in. diameter air operated vibrators were used below the 360 ft. 6 in. elevation and four 3 in. dia=eter vibrators were used above the 360 f t.
6 in. elevation. Lifts within the placement vera limited to 18 in.
28 day strength tests of the concrete yielded averages ranging between 5160 pai and 6790 psi.
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STRUCTURAL ANALYSIS OF THE "AS-EUILT" FUEL HANDLING PCOL SUPPORTING h7.LLS Backcround and Summary The original construction sequence for the reactor build-ing structure of the Three Mile Island Nuclear S tation, Unit No.
2, called for removal of the temporary supports for the polar crane (four braced sccel latticed-tewer structures) prior to placement of concrete for interior structures.
De-lays in sc=e phases of construction work have required changes in this original sequence.
The composi;e liner / concrete ex-terior wall structure, the permanent support for the polar crane, was not available to receive t'ie loads imposed by the polar crane,at the time the construction schedule called for removal of the temporary tower sup*, orts to allev for crection of interior structures.
Because cf the above, and after care-ful engineering evaluation of conditions created, the con-struction plan was =cdified to allcw partial embedment of scme ecmpenents of the rwo north te. ears (temporary supports for polar crane) in the structural cencrete of the fuel handling pool walls, thus permitting continuance of construction effort.
The bottom plates (5 ft. X 5 ft. ) af these tcwers were left embedded in the walls supporting the fuel handling pool structure, thus causing an interruption in structural continuity 0-gg=*u Lu
in some portions of the same walls from the reactor building basement floor elevation and extending 5 feet upward.
During the October 21, 1974 meeting at Bethesda, Mary-land, the AEC requested that the owner sub _t to the AEC 3 structural an apprcximate analysis intended to prove adequacy of the as-built fuel handling pool supporting walls in resisting all loads (including 'ead loads, live loads-earthquake loads, thermal effects, pressure associated with postulated accidents, etc.) in such combinations as expected or postulated to occur during the life of the plant.
The results of such an approximate analysis, based on the assu.1ption that the portions of the walls bounded by the em-tedded plates have no structural strength (i.e.,
are equiva-lent to openings) are presented in what follows and prove caeymatc to r: c_rt :11 crra""ed /
that the as-cullt structure 13 postulated leads during its useful life.
3ETHOD OF.M!ALYSIS In accordance with the decisicas of the meeting mentioned performed.
above, an approximate analysis of these valls is Steps of such an analysis can be summarized as below:
1.
North-east fuel handling pool wall is selected to in represcr.t the most critical case of the two pool walls which the bettem plates are embedded.
These embedded
-?
plates will be assumed to cause permanent openings in the walls.
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2.
The north-east wall is analyzed assuming that the wall is continuous and that there are no openings.
The forces, and moments obtained from such an analysis and acting en one-half of the width of the opening, are then distributed to an adjacent strip of the same wall.
This strip is as-sumed to have a width equal to 1.5 times the thickness of wall.
As a result of the above assumption, the strip adjacent to the opening will have to withstand leads that are directly acting en the strip plus the loads acting on half the opening.
3.
Loads acting en the north-east wall are dead 1 cads, live loads, thermal effects, earthquake loads, and a pressure differential due to peatulated accidents.
These are as lutiows:
a.
Dead Loads including the weight of concrete walls and slabs and the hydrostatic pressures frcm the water contained in the fuel pool.
b.
Live Load due to laydcwn equipment weight en the bettem slab of the fuel pcol.
c.
Thermal Ef fects - Effects due to a temecrature in-crease of 6T = +40 F.
This is associated with summer conditicns which create the worst effects, and represents the difference between the summer normal operating temperature inside containment
(=ao nara.-nco
- n. ) and the to.perature durin: : n-struction.
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d.
Earthquake Loads (for OBE and S5E) derived frca seismic / dynamic analysce o f the containment structure (see Reference 5,
~
Differential Pressure of ZLp = 20 psi, considered e.
acting across the north-east wall.
This pressure differential includes an uncertainty factor of approximately 1.25 (refer to answer to FSAR ques-tion Q. 3.1) and a conservative dynamic load factor of 2.0 to account for dynamic application of load.
4.
To analyze the wall, the plate theory is used (Refer-ence 1 and 2).
Different edge conditicas for the wall are assumed in order to maximize forces acting on the opening and the adj acent strip.
The largest moments induced near the bettc= cf the wall are due to locas from thermal growth.
These mcments are cbtained in the rollowing way:
It is assumed that the slab of the fuel pccl expands freely.
The uniform deflections are then assumed to displace a unit width of the wall acting like a fixed-fixed beam.
As a result vcry conservative =cments are obtained acting at the bottom of the wall.
Similar ccn-servative assumptions are made when analyzing the wall structure for the effects of individual design loau s.
5.
Individual load cases are ccmbined in accordance with the AEC requirements (Reference 6) for service and fac-tored loac conditicas.
The maxi.num mcments and forces m..
I #9 s..
Lto
(i.e.,
the requi.ed section strength) obtained frem these ccmbinations are then compared with the section capacity of the as-built structure.
For such comparisen the ACI Codc
.28-71 (Reference 3) is used to calculate the available strength of the section (moments and forces) including the appropriate capacity reduction factor.
The sumnary of such a comparison is tabula.ed belcw.
SUM *GRY CF RESULTS REQUIRED AS-EUILT SECTICN SECTION RATIO = A.B.S.S.
FORCE STRENGTH ST RENGTH R.S.S.
Moment (k-f t/f t) 312.34 717.45 2.30 Axial Force @/f t)
.03 365.30 2.30 In-Plane Shear (k/f )
40.52 129.86 3.20 OLc-2. t race v:
Wall (k/ft) 37,87 77.73 2.05 As seen frem this tabulation, the available sectional strength of the "as-built" structure exceeds the required section strength by 105 percent to 220 percent.
CCNCLUSION The results of the approximate structural analysis of the "as-built" fuel handling pool supporting walls presented herein prove the adequacy of the walls to resist all expected / postulated loads during the plant's life.
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O REFE RENCES 1.
- Moddy, W.
T., "Mcaents and Reactions for Rectangular P l.a t e s ", Engineeri:.g Monograph 27, Washington: Sureau of Reclamation of the U.
S. Department of the Interior, 1966.
2.
Timoshenko, S., and S. Woinewsky-Krieger, " Theory of Plates and Shells", New York: McGraw Hill Ecok Co.,
1959.
3.
ACI 318-71, " Building Ccde Requirements for Reinforced Concrete", Detroit, 1971.
4
- Dunham, W.
C.,
"The Theory and Practice of Reinforced Concrete", McGraw Hill Ecck Cc=pany, 1966.
5.
Final Safety Analysis Report, T:H Nuclear Station, Unit No.
2, Docket :To. 50-320.
G.
Acu o utuvuural Engineering 3rancn Directorate of Licensing,
" Structural Design Criteria for Category 1 Structures Other than Containment", June 1974.
7.
- Ferguson, M.
P.,
" Reinforced Concrete Fundamentals",
New York: Jchn Wiley and Sens, Inc.,
1969.
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