ML19320B092
| ML19320B092 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 07/07/1980 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8007090291 | |
| Download: ML19320B092 (21) | |
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BALTIMORE GAS AND ELECTRIC COMPANY P. O. BOX 1475 BALTIMORE MARYLAN D 21203
, July 7, 1980 ARTHUR E. LUNOVALL.Jn v.cc PREssotNT j )
s... ,r ,7 Office of Nuclear Reactor Regulation A > '
U. S. Nuclear Regulatory Commission 9 Washington, D. C. 20555 l,]i , (g /
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Attn: Mr. R. A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Subject:
Calvert Cliffs Nuclear Power Plant Units Nos. 1 & 2, Dockets Nos. 50-317 & 50-318 Spent Fuel Pool Modification
References:
a) BG&E letter dated 7/3/79 from J. W. Gore, Jr.
to H. R. Denton, Request for Amendment.
b) NRC letter dated 9/21/79 from R. W. Reid to A. E. Lundvall, Jr., same subject.
Gentlemen:
Regarding the referenced request for license amendment the
~
following additional information has been requested by your Staff.
Spent Fuel Pool Structural Analysis The pool structure was reanalyzed considering the new loading conditions due to the rack modification. The new loading conditions
' include the additional rack and fuel veight and the effects of temperature.
The temperr.ture effects include the overall te=cerature effect, thermal gradient for normal operation, and thermal gradient for the accident case.
A three-dimensional linear finite element model was used to determine the controlling elements in the structure. These controlling elements were
. then checked for allovable stresses and allowable strength.
Because the pool was originally designed and constructed under the ACI 318-63 code, the initial design check was ~ar compliance to this code and the project FSAR. As a result of questit ts from your Staff, additional checks were made for compliance to some of the later codes and standards. The design was checked for new loading combinations and strength requirements as defined in ACI 3h9 and Standard Review Plan 3.8.h. This check addressed loading combinations and strength requirements only, and did not address other aspects of the code.
All loading combinations in both standards were considered and the following cases were found to be controlling:
THIS DOCUMENT CONTAINS
-8007090 M ,,,,
Mr. R. A. Clark July 7, 1980 4
Standard Review Plan, Section 3.8.k U = 1.kD + 1 7L + 1 9E -
U = 0.75 (l.kD + 1.7L + 1 9E + 1 7To + 1.7Ro)
U = D + L + Ta + E '
ACI 3h9 U = 1.hD + 1.hF + 1.7L + 1 7Fo + 1.7Ro U = 0.75 (l.hD + 1.hF + 1 7L + 1.7Eo + 1.kTo)
U = D + F + L + Ta + % s The loading of each controlling element has been checked against its ultimate capacity. A summary of the ultimate moment capacities and the moments due to the controlling load combinations is in Table 1.
In no case is the ultimate capacity exceeded by more than 1%.
Liner Plate Analysis Attachment A to this letter contains a rt: vised ans :er to -
Question No. 10 in reference (b). Attachment B contains additional information on Question No. 10.
Very truly yours,
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cc: Mescrs. E. L. Conner, Jr.
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. ! l Attachment Al NRC Ouestion:
- 10. For the accident fuel assmebly drop condition, describe in detail the assumptions, type of analysis, ductility ratios and allowable stresses used in the analysis to insure that the acceptance criteria for this case are satisfied. Provide, also, your basis for concluding that the leak tightness of the fuel pool is maintained.
NES Response:
For the accidental fuel assembly drop condition,1300 pound weight (fuel assembly) was postulated to drop on the rack from a height of 24 inches above the top of the rack. Three cases were considered: 1) a direct drop on the top of a 2x2 module, 2) a subsequent tipping of the assembly onto the surrounding storage cans, 3) a straight drop through the storage cell with impact to the rack base structure.
Linear and non-linear analysis techniques using energy balance methods as indicated in Appendix D of Reference 1 are used to evaluate the structural damage resulting from a fuel assembly drop onto the rack.
The acceptance criteria for the accidental fuel assembly drop on the rack are 1) the resulting impact will not adversely affect the overall structural integrity of the rack, the leak-tightness integrity of the fuel pool floor and liner plate and 2) the deformation of the impacted storage cells will not adversely affect the value of keff or the ability to cool adjacent fuel elements.
The results of the fu,el assembly drop analysis using energy balance methods are summarized in Table 1. From Table 1 it can be seen that for a straight drop of fuel assembly on top of the storage cell, the maximum stress in the storage cell is slightly greater than the dynamic yield stress for stainless steel, thus indicating that the ctorage cell and its flare will undergo local permanent deformation but the overall storage rack structure will not yield. It can also be seen that the maximum stresses in the rack base structure, rack support legs and the bearing stress on the concrete floor under leg are within the allowable values. The maximum punching shear stress in the liner plate is greater than the allowable shear stress value but less than the ultimate shear stress value, thus indicating minor imprinting of the liner plato around the periphery of the impacted leg base plate. The punching shear stress calculations are performed by using a very conservative assumption that the concrete under the rack support feet base plate does not provide any vertical support and the load is transmitted by the punching shear mechanism. The external kinetic energy of the dropped fuel will be l absorbed in the local deformation of the storage cell flare at the top i of the storage cell, and in the axial deformation of the storage cell.
However, the liner plate will not be perforated insuring the leak-tight integrity of the fuel pool liner place will be maintained.
A2 The free fall of a fuel assembly through the storage cell from a height of 24 inches above the top of a storage rack and its Lapact on top of the cell base plate and rack base structure was analyzed using empirical missile equations (the Ballistic Research Laboratory) .
The results indicate that the maximum thickness of steel plate that could be perforated by such a missile is slightly less than the thickness of the cell base plate. Therefore, during a fuel assembly drop accident of this type, the fuel assembly lower end fitting feet will perforate the cell base place, however the lower and fitting '
support place will prevent further penetration of the fuel assembly and subsequent impact to the pool floor liner place. The kinetic energy developed during the free fall will be absorbed by both the bending and shearing of the cell base plate.
- Since for this fuel assembly drop case, the external energy is absorbed in the flexural deformation of the flexible cell base place and rack i base structure, the reaction load transmitted to the rack base structure, rack feet and pool floor is less than that for fuel assembly drop on top j of the storage cell. Therefore, the damage to the pool floor will be 1 less severe for the fuel assembly drop through the storage cell than that i for the fuel assembly drop on top of the storage rack.
4 The fuel assembly drop analyses have been performed by conservatively l assuming that no energy will be absorbed by the fue. assembly itself.
l The energy absorbed in the deformation of the flexible fuel assembly I will result in reduced damage to the storage rack and the pool liner i
plate than that predicted by the conservative analysis.
It.has, therefore, been concluded that neither the straight drop of the fuel assembly on top of the storage cell or the straight drop of the fuel assembly through the storage cell with impact on top of the rack base structure will not adversely aff ect the value of keff or the leak- i tight integrity of the pool.
i TABLE 1 1
RESULTS OF AN ACCIDENTAL FUEL ASSEMBLY DROP (Reference 1)
Straiaht Drap on Top of Storage Cell Calculated Value Allowable Value Weight of Q2el Assembly (kip) 1.380 --
Maximum Drop Height (in) 24.0 --
Kinetic Energy of Drop to be Absorbed (in-k) 33.12 --
Maximum Strain-in Storage Cell (in/in) 0.001723 0.4851 Maximum' Cell Axial Deformation (in) 0.290 --
A3 TABLE 1 Cont'd Maximum Stress in Cell (ksi) 32.74 30.02 Maximum Transmitted Reaction Load (kips) 137.0 --
Maximum Stress in Rack Base Structure (ksi) 22.9 30.0 Maximum Stress in the Weld Between the Beams and Support Legs (ksi) 10.73 23.33 Maximum Stress in Rack Support Leg (ksi) 17.73 74.74 Maximum Local Bearing Stress on Concrete Floor (ksi) 1.74 3.57 Maximum Punching Shear Stress in the Liner Plate (ksi) 24.99 66.05\
Drop Through Storage Cell (50.5/
Maximum Drop Height (in.) 192.5 Maximum Free Fall Impact Velocity (f t/sec.) 32.13 --
Maximum Unsupported Plate Thickness that may be perforated by missile free fall velocity, (in.)
BRL Formula 0.454 0.50 Maximum External Kinetic Energy (in.k) 265.7 --
Maximum Transmitted Reaction Load (kips) 48.1 --
- 1. Ultimate strain for stainless steel.
- 2. The allowable stress value represents dynamic yield stress for stainless steel.
'3.
Allowablestressintheweld-1.6x21xff=23.3ksi.
- 4. Buckling Stress for 17-4 PH stainless steel at design temperature.
- 5. Ultimate Shear Stress Value = 2/3 of ultimate tensile stress value, United States Steel Corporation "U.S. Steel Design Manual", ADUSS, 27-3400-03, May 1974, = 2/3 x 75.0 = 50.0 ksi.
1 Attachment B i i
Additional Response to NRC Question 10
- 1. Ductility Ratio the detail structural calculations for the accidental fuel assembly drop analysis are given in Appendix D of Reference 1 (copy attached).
For a straight drop of a fuel assembly on top of a storage cell, the ductility ratio has been calculated to be 1.61 (page D-3 of attachment) . Since the kinetic energy for the inclined drop of a fuel assembly on top of a storage cell is less than the kinetic energy for a straight drop ota top of a storage cell, the ductility ratio tor the inclinei fuel assembly drop will be less than that for a straight drop. For the cese of a fuel assembly drop through the storage cell with impact on the cell base plate, the energy absorp-tion mechanism is essentially by the perforation of the cell base plate rather than the ductile flexural deformation of the cell base plate. Therefore, the evaluation of a ductility ratio for this fuel assembly drop case is not applicable.
Fuel Pool Liner Integrity The structural and leak tight , integrity of the fuel pool liner is essentially determined by the structural integrity of the concrete underneath the rack support feet.
From page D-6A of the attachment, it can be seen that the maximum bearing stress (1.74 ksi) in the concrete is smaller than the allowable concrete bearing stress value of 3.57 ksi. Therefore, there will not be any plastic deformation of the concrete under the rack support feet / liner plate and the load will be transmitted by direct compression of the liner plate. The resulting vertical compressive stress in the liner plate (Figure 1) is equal to 1.74 kai which is significantly smaller than the allowable axial stress value of 30.0 ksi. (dynamic yield stress for stainless steel)
Therefore, there will not be any plastic deformation of the liner plate and the leak tight integrity of the fuel pool liner plate will be maintained.
In order to evaluate the degree of local imprinting in the liner plate around the periphery of the rack support feet base plate, the punching shear stress was calculated. This calculation requires the conservative assumption that the concrete under the rack support feet base plate does not provide any vertical support and the load is transmitted by the punching shear mechanism. The maximum punching shear stress (24.99 ksi) in the liner plate is found to be less than the shear stress value required for punching which typically is greater than the ultimate shear stress value (2/3 of ultimate tensile stress value per Reference 2 = 50.0 ksi). Consequently, the local imprinting of the liner plate around the periphery of the base plate of the impacted leg will be minor.
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References:
- 1. Nuclear Energy Services, Inc. Document NES 81A0566, " Structural Design Analysis Report for the Calvert Cliffs Unit 1 Nuclear Plant High Density Fuel Storage Racks"
- 2. U.S. Steel Design Manua'.; United States Steel Corporation; ADUSS 27-3400-03: May 1974
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5.to NE5 bw@. 6oEE 7.'75 g Issur.tJ o-3 i
NES 105 (2/74) l l
l
Sy N.E DATE 3"f-M PROJ. 5/ M YAsy D*
NUCLEAR ENERGY SERVICES INC. dd p"/i' CHKD.I M DATE3 U~E PAGE OF NES DIVISION c.A uv.e re.T cuch =s wore s vc.a.
A4 4-sDt.P.TAL 707\ DMP AN AW5ts REF.
Em THE 'BASF_ 2es TMs* Loa. css Wiu. s c Tet.4 N Sacr o n C T Mt vree t- T5y weOs. Assums 7e7. or 7s6 r o,4o wiv ser TAesN ISy n4e. weios ON Tam Mosr- HeaAiy t.o e o e n t sc,.
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- z by / Moa.s aomm b/eto D L.
x n NES 105 (2/74) l l
SY 2 DATE 3' '- M PRO!. U 3 Y TAsy h NDCt. EAR ENERGY SERVICES INC. . -
NES DIVISICN CHKD. 1. M OATE - & 73 PAGE NA OF c m an:r c-s m .- s =us. ucx Ac.c.tER.NT4 FUE.t. D1tcPANAt.W.S REF.
cwec.w woe. s.use_ wwree ow n e.ca.ese.t> t_ v 6,64WCW STW,G. Ow TheagAc>i . og T-oo t" gr og $ 6.g a rt. .
h.a t8a m e ,== on T.4e r e ne.o s wi~ e hcr_ is 3k - Sow = s.os7/.l.su seae. Aaea. . 7ra. 4 meon - n(s.es@(/.6) 7 f
a 14. 6'5' ?.4 1. S
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i<.ss -
,-ma s,.
a um= (i.1)t'zsjac k' /i; l% \
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ceAa mc. Ae a.= Trf/OI k gq.,5 %ss'
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l %ms Men : 77"O f = T!"lf o)( lt)
- 6. 6'1 y, _/4 W = 24,% .l. < S01< .1t.,,
C.39 l
l Yt. rim 4 r? SNae4 T rar.e n : N //t.nra f73 72us/t.e Sneess IA4 pq r
(0,2%6D-lk 'f,.r'7.6;~O=SD0 Wl. D(k 3
j a?r.orrt //pa ess/y.si.c 3 it can be seen that the :naximum bearing stress ,(1.74 ksi) in the cencrete is, smaller than the, allovable l concrete bearing stress value of 3.57 ksi. Therefore, there will not be any plastic deformation of the concrete under the rack support feet /
l liner place and the load will be transmitted by direct compression of the liner plate. The resulting venc>l . compressive stress in the liner plate is equal to 1.74 ksi which is significantly smaller than the allowable axial stress value of 30.0 ksi. (dynamic yield stress for stainless steel) Ther efore, there will not be any plastic deformation of the liner plate and the leak tight integrity of the fuel pool liner place will be maintained.
NES 105 (2/74)
. SY /*'b I#'M DATE / '-//'#O PROJ8/3 O TASK NUCLEAR ENERGY SERVICES INC. kE NES CIVISIGN CHKD. DATE H -80 PA A AS OF S-g gjg ,j g
[ cc.A E w m L % E L 'Dito 9 ANALYS& REF.
In order to evaluate the degree of local imprinting in the liner plate around the periphery of the rack support feet base plate, the punching shear stress was calculated. This calculation requires the conserva-tive assumption that the concrete under the rack support feet base plate does not provide any vertical support and the load is transmitted by the punching shear mechanism. The maximum punching shear stress (24.99 ksi) in the liner plate is found to be less than the shear stress value tequired for punching which typically is greater than the ultimate shear stress v'alue (2/3 of ultimate tensile stress value per Reference W = 50.0 k.si). Consequently, the local imprinting of the i liner plate around the periphery of the base plate of the impacted leg will be minor.
l 3 U.S. Steel Design Manual; United States iteel Corporatio'n; ADUSS
~
27-3400-03; May 1974
( _.
EvAtuATian oF .5TRAN 'C\l Tv L.# ER Pt.rkTE MO Cst 4CEETE.'
Niew Sbraiu m b liwa fIcEf b'.h A:dd 5 hress ( ) 74- Ksi = MidSMS _JM'8 coo
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M Alusy EbhON E ~ 7 54va'm ed W eir), == 3--- ==
tsooo c . ool) iw),u ho. o e6 9'w 1. i q g,-4-u
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, $ c),oo00$ $Hhg, 11f
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' gy \ ha Gw DATEI e-II-TO PROJ.5M TASK 33' NUCLEAR ENERGY SERVICES INC. *//~N PAGE E~S C OF CHKD. -
DATE h Il Nas oiwsi:N r ,t Cgg g i rug gcKs ACCT BEN.TAL FusL. 'bR.oP % 4Y515 REF.
h CX b
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L VM 'N Typical concrete streer .ia curve u M/ / r -
1000 t -
C crete s a in./in.
1.00 ^'
[ N- 'sN, 100 days N Iday 0.7S g g s Ihr
. H9. LE Stress-etrain curves at various attain rates, concentric
, 0.50 -
g t min cornpression. ,
L' sk 0.25 Ed
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0 I 0 0.0 01 0.002 0.003 0.004 NES og Concrete strain
, gy 2.Y DATE 3 - /2-79 PROJ. 6'3+ TASK 3##
NUCLEAR ENERGY SERVICES INC. \. M NES DIVISION eg gg, OATE3*EG PAGE D-7 OF ~
d'ALVE.RY dL\ M w I 7"O E L. h c. W.
AcctosNTAL rug.L o tt:o9- AH o. Lysis REF.
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NES 105 (2/74)
sy RR DATE 3-/279 PROJ. 8/34 TASK
- NUCLEAR ENERGY SERVICES INC. ~
NES DIVISION CHKD. IM OATE 3 #bAGE S"'3-OF
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OF cau :5, LEN GTH. apr FOG'L A 5 5 E M '5 L Y = /S S uV Humass ow csu.s em piAc.ow %
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. SY I*E DATE 3 -12 M PRoj. m 7A3g 1oo NUCLEAR ENERGY SERVICES INC.
CHKD. I' M DATE t -?.A-7} AGE D OF NES DIVICION C.ALVEP T" GL)SF4ss -cua s E Ac./.e:.
e A/C.lDENT241, FVE.L C1P.oP ANAL:<5\S REF.
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, gy 92 DATE T 83'D PROJ.8/ M TASK ~2 8 0 NUCLEAR ENERGY SERVICES INC. b OF b I\
CHKD. IM DATE '3*I'#-HPAGE NES DIVISION c 4Lv acT" 4.t W5% -s\ var.t e 7Ac.v hg DsNTAb. EP.t DEog. Aed*6YSt% REF.
Um f.owen a.Mu vrtTtwGis w'ics. Pas e.raArc 7"?+E d " ?ASE Pl. ATE AELsoASING ACONStoGRASt.E AMOUNT 0 5~ THE ORol* BNEMY, WG s*'ve L C.EE LL w it L " Toe PmNET8tATIN6- Af" THE REhMtow GR\ D. C;'aus;e % st Ensa PY,, ATE IS Loc Aftt O Asout 14k
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8Y '
DATE '8-'3-W PROJ.h TA3r
- NUCLEAR ENERGY SERVICES INC.
CHKD. TV OF b
~
DATES -M-W PAGE NE3 DIVISICN CAi vezr c_um: s ., / ::v e L. a c k M r4.nowwTAL, Ett DRoD MNALys,as REF.
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RE FEREN dE5 .*
b.l . Nucles.n En Ie W \'r a.s I h c .
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a.a . wa - %A w %h .d
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