ML19261B853
| ML19261B853 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 03/01/1979 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-08-01, TASK-8-1, TASK-RR LAC-6141, NUDOCS 7903070325 | |
| Download: ML19261B853 (10) | |
Text
_
\\
DitIR1 LAND PO WER COOPEltatTIVE Sa Crone, 0Yhconsin 54601 March 1, 1979 In reply, please refer to LAC-6141 DOCKET NO. 50-409 Director of Nuclear Reactor Regulation j ~
U.
S. Nuclear Regulatory Commission
~
Washington, D. C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 PROPOSED MODIFICATION - SPENT FUEL STORAGE
REFERENCE:
(1)
Telecopy from J. Wetmore (NRC) to R.
y dated February 14, 1979 Gentlemen:
Enclosed with this letter is additional information required for your review of the proposed modification of the LACBWR spent fuel storage pool.
This information is provided in response to a tele-copy from J. Wetmore (NRC) to R.
Shimshak (DPC ), received February 14, 1979.
Please contact us if additional information is required.
Very truly yours,
DAIRYLAND POWER COOPERATIVE rdUw x
Frank Linder, General Manager FL:NLH: abs Enclosures cc:
(See attached sheet. )
\\
1 7903070395
---mi-u w
Director of Nuclear Reactor Regulation LAC-6141 Washington, D. C.
20555 March 1, 1979 cc:
J.
G.
Kepp le r, Regional Director U.
S. Nuclear Regulatory Commission Directorate of Regulatory Operations Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Charles Bechhoefer, Esq., Chairman Atcmic Safety and Licensing Board Panel U.
S. Nuclear Regulatory Commission Washington, D. C.
20555 Mr. Ralph Decker Route 4 Box 190D Cambridge, MD 21613 Dr. George C.
Anderson Department of Oceanography University of Washington Seattle, WA 98195 O.
S.
Hiestand, Jr.
Attorney at Law Morgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D.
C.
20036 Kevin P.
Gallen Attorney at Iaw Morgan, Lewis & Bockius 1800 M Street, N.
W.
Washington, D. C.
20036 Coulee Region Energy Coalition P.
O. Box 1583 La Crosse, WI 54601
RESPONSE TO NRC REQUESTS FOR ADDITIONAL INFORMATION SUBMITTED BY FEBRUARY 14, 1979, TELECOPY FROM J. WETMORE (NRC)
TO R.
PROPOSED MODIFICATION - SPENT FUEL STORAGE NRC ITDI L DFC response to I5] Item 2 (a) - Indicate whether the lotxt grid referred to in the response is in the upper or lover tier and justify applying the uplift load at this point vs. in the tier not examined.
DPC-RESPONSE The lower grid referred in the response to Item 2 (a) represen ts the lower grid of the lower tier rack.
Due to the uniform support provided by the upper grid of the lower tier rack, the uplift load will produce lower stresses in the lower and upper grid structure of the upper tier rack.
It should be noted that the stress analysis for the uplif t load has been performed very conservatively by assumiag that the lower grid of the lower tier rack alone, will resist the uplif t load.
In reality, the upper and lower grid structure will act together in resisting the effects of the uplif t load.
Therefore, the actual stresses in the storage rack structural members will be less than those calculated in NES Report 81A0546, Revision 2.
(t) Item nunbers refer to n:chered responess in enclosure submitted by DPC Letter (latC-6067), Linder to Director of Nuclear Reactor Regulation, dated January 4, l979.
NRC ITDI 2 DM response to Item 2b (iv) - If a fuel as.sembly is pcstulated to drcp straight through' a storaga cell at a locaticn where both the upper and icuer ners contain a fuel assembly, quantify the resultant rack stresses, reacer lcac,a, effect on both fuel acaembly suppcrt plates, and the a-ber of equtvalent fuel assemblies damaged.
DPC RESPONSE If a fuel assembly drops straight into a storage cell at a location where both the upper and lower tiers contain a fuel assembly, the external kinetic energy (30. 3 k. in. ) of the dropped fuel assembly will be absorbed in the deformations of various resisting elements essentially acting in series through the process of axial, flexural and shear deformations.
The resisting elements consist of the handle and fuel pins of the impacted upper tier fuel assembly, upper tier fuel assembly support plate, support plate bracket weld to the egg-crate grid, lower tier storage cell, and the rack base.
The load deformation characteristic of series combination of resisting elements can be evaluated by determining the load deformation characteristics of the individual elements.
Elements with low load carrying capacities will be deformed first and will absorb part of the external energy, the remainder of the external energy will be absorbed in the deformation of other stiffer elements.
Analysis indicates that the handle and segmented fuel pins of the Allis-Chalmers fuel assembly are the weakest elements (load carrying capacity of 10.0 Kips) and will absorb approximately 16. 7 k.
in of the external energy.
Similarly, the handle and longer tie rod fuel pins of the Exxon fuel assembly will absorb approximately 20.1 k.
in.
of the external energy.
The remaining 10.2 k. in. to 13.6 k.
in, of the external energy will be absorbed in the axial and flexural deformation of the regular fuel pins ofthe fuel assembly (load carrying capacity of 20.2 Kips).
The load and energy carrying capacity of the fuel assembly support plate and the weld between the support plate bracket and the weld between the support plate bracket and the egg-crate grid are in the order of 53 Kips and 25.7 k. in. respectively.
Thus, even if it is conservatively assumed that the regular fuel pins of ths fuel assembly do not absorb any of the remaining external energy, the remaining external energy is not sufficient to shear off the weld between the support plate bracket and the egg-crate grid.
There-fore, the following conclusions have been drawn for the fuel assembly drop into a storage cell at a location where both the upper and lower tiers contain a fuel assembly:
i i
(1)
The dropped and the impacted fuel assemblies will be damaged by the drop, (2)
The fuel assembly stored in the lower tier will not suf fer any damage, and (3)
The maximum reaction load that will be generated during the impact does not exceed 53 Kips, which is significantly less i
than the reaction load developed during the fuel assembly l
drop on top of the storage cell (9 4. 6 Kips ).
The over-all structural integrity of the fuel storage rack, rack base i
structure, rack feet and pool floor have been evaluated for the reaction load of 94.6 Kips.
Since the reaction load is less for the fuel assembly drop into a storage cell at a l
location where both the upper and lower tiers contain a fuel l
assembly, the effects of the drop on the rack, rack base structure, rack feet and pool floor will be less severe than that for the fuel assembly drop on top of the storage rack.
NRC I m 3 DPC response to Item 2b w.
Quantify the reaction load genera:ed during i.~rac:.
DPC RESPONSE The maximum reaction load generated during the fue1 assembly drop through the storage cell and impact on top of the fuel assembly support plate in the upper egg-crate grid of the lower tier rack is in the order of 53 Kips.
The maximum reaction load of 53 Kips represents the load carrying capacity of 3/16-inch welds between i
the support plate bracket and the upper egg-crate grid of the l
lower tier rack.
i
~
l NRC tim 4 DEC response to Item 2b (vi) - Indicate wh,ather the 38% increase *in yield strength used in the rack analyses (taken frcm CE data and athstantiated by tensile testa perfomed by NES) is an average value or an upper cr Ecuer bcund.
If 39% is an upper bound,,lustify using the increase for calculating alicuable s tresses.
I 3-
DPC RESPONSE In the fuel assa #,2 drop analysis, the dynamic yield strength value for stainless etw1 is taken as 38% greater than the static yield strength value.
The 38% increase in the yield strength represents an upper bound value, however, its use in calculating the allowable stresses is justified considering the conservative assumptions that have been made in the fuel assembly drop analyses (Response to Item 2b (ii) indicates various conservative assumptions made in the fuel assembly drop analyses).
Even if an average increase of 25%
is considered in the dynamic yield strength value, the allowable stress value of 41.4 ksi given in Table 8.5 of the NES Report 81A0546, Revision 2, will change to 37.5 ksi.
This reduction in the allowable stress value for stainless steel will not change the results and conclusions of the fuel assembly drop analysis presented in NES Report 81A0546, Revision 2.
NRC ITD! 5 DPC respcnse to Item 4 - Specify the magnitude of the clearances berJeen racks.
and betueen the racks and the pool 1.ntis.
DPC RESPONSE The clearances between racks will be in the order of 0 to 1/32 inches.
The clearances between the racks and the pool walls will be in the order of 3/16 to 5/16 inches.
This essentially accounts for the thermal expansion requirements.
NRCITDI6 DPC respcnse tc Item to - Clarify whether Table L.53 was used instecd of Table 2. 5. 2. 2 as stated.
DPC RESPONSE Table 1.5.2.1 mentioned in DPC response to Item 10 should be l
corrected to Table 1.5.3.
l i
i i
NRC ITEM 7 DPC respcnse to Item 22 - Yne load carrying capacity of similarly acnfigured fuct assemblies is disc:rged and used to justify the capacity of MCEhR fuel.
Ynis response is qualitative at best. Provide a quantitative, technical justification showing that the UCBh*d fuel assemblies .22 not suff3r damage frcm the marimum eapected impact loading.
DPC RESPONSE The limiting impact load that the fuel assembly can withstand without suffering any damage is in the order of 1500 pounds.
The minimum fuel assembly impact load carrying capacity of 1500 pounds is larger than the maximum impact load of 936 pounds sustained by the fuel assembly during a seismic event.
The LACBWR fuel assembly fabricated by Allis-Chalmers can with-stand a shock load of 6G (equivalent total impact load of 2300 pounds) without sufiering any damage (Reference 1).
Similarly, the LACBWR fuel assembly fabricated by Exxon Nuclear Company can withstand a shock load of 4 G (equivalent total impact load of 1500 pounds) without suffering any damage (Reference 2).
The impact load carrying capacities of Allis-Chalmers fuel assembly (2300 pounds) and the Exxon Nuclear fuel assembly (1500 pounds) are larger than the maximum impact load of 936 pounds sustained by the fuel assembly during a seismic event.
NRC ::DI 3 DPC response to Item 24 - Indicate whether both cases of L having its full value of being acr:pletely absent, vere checked.
DPC RESPONSE The stresses in the LACBWR fuel storage pool floor and walls are maximum when they are subjected to the full value of live loads associated with the hydrostatic pressure, and the weights of the fully loaded racks, crash pad and the spent fuel shipping cask.
Due to the small size of the pool (11 ft. x 11 ft.) and the presence of the supporting wall under the middle of the fuel storare pool, the load case with the live load being completely absent wil.'.
result in lower stresses in the fuel storage pool structural components.
Therefore, fuel storage pool structural analysis has been performed for the case of live load having its full value only.
NRC ITDI 3 DPC response to Item 27 - Discuss hou the refueling canal gate uaa modated in the seismic analysis and discuss the gate seal integrity when the hydrcatatic, hydrodynania, and uater sioshing effecta are neglected.
DPC RESPONSE The refueling canal gate was not included in the structural analysis model of the reinforced concrete walls and floor of the fuel storage pool.
The reaction loads resulting from various loads on the refueling canal gate were applied at the appropriate nodes of the pool wall model.
During a SSE Seismic event, the canal gate will be subjected to an additional pressure load of 1.85 psi resulting frcm the seismic inertia load of the constrained water mass and the water sloshing effects.
The additional pressure load of 1.85 psi resulting from a SSE Seismic event is fairly small as compared to the maximum hydro-static pressure load of 8.67 psi.
Therefore, the canal gate and the gate seal will maintain their integrity during a seismic event.
The compression load produced by tightening the canal gate bolts will maintain the gate seal integrity if the hydrostatic, hydro-dynamic, and water sloshing effects are neglected.
NEC ITDI 20 DPC respcnse to Item 20 - Clarify whether gross stresses in the racks remain beleu :he L.GS li:nt.
DPC RESPONSE For the case of the cask drop on top of the fuel storage rack, the maximum stresses in the impacted storage cells and the rack base structure under these cells are greater than the 1.6S limit.
Therefore, the impacted region of the storage rack will not maintain its structural integrity.
However, analysis shows that by providing a 3/8 inch thick stainless steel barrier plate under tfie storage racks, the drop of a shipping cask on top of the storage racks will not damage the pool liner plate and the pool floor sufficiently to adwrsely affect the leak-tight integrity of the pool.
NRC ITC'4 ii DEC response to Item 22 - If it is asswned that the cack i. pacts a rack or racks alcng an edge or corner of the cask rather chan in a vertical pcsition, verify that this event vill not result in the total collapse of the storage rack a:d/or adversely affect the value of keff. If the cask impacts a atcrage rack at the corner of the cask at a positicn directly above a leg assembly, disc:es the ocnsequences including the reaction Icad transmitted to :he liner and ficor.
DPC.iSPONSE If the cask impacts the fuel storare racks along an edge or corner of the cask (inclined cask drop), the total number of impacted storage cells will be less than that for the vertical eask drop (cask dropping with its axis vertical), considered in the cask drop analysis (NES Report 81A0550, Revision 2).
For the inclined cask drop, the local and over-all damage to the impacted storage cells will be greater than that for the vertical cask drop.
However, due to the lower resistance offered by the fewer number of impacted cells, the maximum reaction loads generated during the inclined cask drop will be less than that for the vertical cask drop.
Even if the inclined cask impacts the storage cells direcity above a leg assembly, the maximum reaction load transmitted to the support leg will be smaller than that considered in the cask drop analysis.
Therefore, the damage to the storage base structure,
rack support feet, pool floor liner, and pool floor will be less severe for inclined cask drop than for a vertical cask drop.
In the unlikely event that a fuel shipping cask is dropped on the storage racks, it has been concluded that keff will decrease.
The basis for this conclusion is that the principal effect of dropping a cask will be to displace water from the fuel rack.
Depletion of water leads to a decrease in keff.
It would not be possible for a dropped cask to eject the poison material from the rack; the crushing effect of the dropped cask could only act to compress the fuel and poison together.
l l
~
REFERENCES (1)
LACBWR Fuel Assembly Design Report, ACNP-65564, Revision 1, April 1966, Allis-ChaLmers, Atomic Energy Division.
(2)
Design Report La Crosse Reload Fuel Assemblies, Type III Fuel, XN-75-28, November 1975, Exxon Nuclear Company, Inc.
i t
I i
!