ML18043A323
| ML18043A323 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/04/1978 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Ziemann D Office of Nuclear Reactor Regulation |
| References | |
| TASK-09-01, TASK-9-1, TASK-RR NUDOCS 7812120100 | |
| Download: ML18043A323 (19) | |
Text
consumers Power company General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201
- Area Code 517 788-0550
©rPW December 4, 1978 REG UL/\\ TORY DOC!~H FILE COPY.
Director, Nuclear Reactor Regulation Att:
Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - HIGH DENSITY SPENT FUEL RACKS
References:
(1)
DLZiemann letter to DABixel *dated June 13, 1978 (2)
DPHoffman letter to DLZiemann dated August 8, 1978
( 3)
WGCounsil (NE Utilities) letter to DLZiemann date.d November 3, 1978 Consumers Power Company was requested by the NRC in Reference (1) to discon-tinue the storage of irradiated fuel in the new spent fuel storage racks until the problem experienced at Haddam Neck was understood and satisfactorily re-solved for the Palisades design.
Reference (l) also requested additional information concerning the B4C material used in the spent fuel racks.
Reference (2) provided Consumers Power Company's response -to the Staff's questions in Reference (l) along with a schedule for submitting test results.
Three of the four tests described in Reference (2) have been completed.
This
.information was presented to the NRC at a meeting in Bethesda on October 26, 1978 and is contained in the Carborundum Company Test Program Report, No CBO-N-78-299 which was submitted to the NRC by Northeast Utilities on November 3, 1978 (Reference (3)).
The Test Program Report in conjunction with additional analyse_~_~~~: Jlr~vided the information necessary to address the NRC Staff, concerns relative to the Palisades Plant spent fuel racks.
Attachment l restates the NRC concerns and provides appropriate responses to close the subject.
Consumers Power Company concludes that the results of the first three tests demonstrate and verify the structural adequacy of the B4C material when exposed to the spent fuel pool environment.
This conclusion will be supple-mented by the results of the fourth test (Long-Term Mechanical Properties Test).
p
2 Consumers Power Company intends to vent the spent fuel racks to preclude gas pressure buildup. provides an update of the safety analysis performed by NUS for the Palisades Plant which takes into account our use of the racks in a vented condition in the spent fuel pool.
This design change has been reviewed and approved and found not to constitute an unreviewed safety question.
Upon completion of venting, Consumers Power Company will install the remainder of the high density racks and begin storing fuel in them.
David P Hoffman (Signed)
David P Hoffman Assistant Nuclear Licensing Administrator CC:
JGKeppler, USNRC
L ATTACHMENT NO 1 Provided below are Consumers Power Company's responses to the NRC's "Request for Additional Information" in Enclosure 1:
ITEM 1 Provide a description and schedule of your program to investigate and resolve the effects'of gas generation within the annular space containing the boron carbide (B4C) neutron absorbing material utilized in your spent fuel storage rack design.
Provide the results of any work completed to date including those efforts which took place prior to the observation of fuel rack swelling at the Haddam Neck Plant.
Response
This information was provided in Reference (2).
ITEM 2 Your program should specifically address the following:
- a.
The mechanism and rate of gas generation including quantity and composition of the gases.
Response
This information is provided in the Carborundum Test Report No CBO-N-78-299 and was discussed with the NRC Staff on October 26, 1978.
- b.
The magnitude of the stresses in the neutron absorber plates considering the limiting loading conditions for the unvented geometry for the rack with the maximum anticipated swelling and demonstrate that such stresses are acceptable.
This analysis should consider mechanical properties of the material including the effects of irradiation, temperature, and strain rate.
Response
The B4C neutron absorber plates have a minimum compressive strength greater than 12,000 psi, and this strength would be virtually unaffected by the effects of the irradiation exposure and temperature experienced to date in the Haddam Neck pool environment.
It is projected that the racks which experienced the highest swelling (ie, those cells which had the inner cell walls contact the stored fuel assembly) could have experienced an internal pressure of approximately 60 psig.
If it is conservatively assumed that this pressure is resisted
solely by compressive loading on the edges of the boron carbide plates, the maximum compressive stresses on the plates are calculated to be less t,han 2, 600 psi which is acceptable when compared to actual strength of 12,000 psi.
- c.
Qualification testing which will demonstrate satisfactory material per-formance of the B4C plates in the spent fuel pool environment for the design life of the racks including the effects of irradiation and temperature.
Response
This information is provided in the Carborundum Test Report No CBO-N-78-299 and was discussed with the NRC Staff on October 26, 1978.
- d.
Should you propose a design change which would expose B4C plates to the pool water, i)
State the maximum pe:r:cerit_~~~ of boron oxide, B203, in the B4C.
Since B203 is soluble in water, it will either be necessary to assume that this amount of boron is leached from the boron plates or to experimentally demonstrate that this will not happen during the life of the racks.
Response
This information is provided in the Carborundum Test Report No CBO-N-78-299 and was discussed with the NRC Staff on October 26, 1978.
ii)
Address any possible corrosion effects in the annulus between the absorber plate and its protective canning and its safety consequences considering the ~iivTrori.iiierit.
Response
The boron carbide plates contain boric oxide which when exposed to the spent fuel pool water might form a boric acid solution in the annulus between the absorber plate and its protective canning.
General corrosion of the Type 304 stainless steel cans is negligible in this environment.
There are also small quantities of leachable chlorides and fluorides (less than 100 ppm total) in the finished plates; but as none of the seal welds on the cans are load bearing welds and the normal service temperatures of less than 120° are less than the threshold for significant halide induced stress corrosion, the potential for and effects of this type of corrosion are not significant.
Because venting of the cans obviates the need for a leak-type seal, the effect of such corrosion, should it be postulated to occur, would be to provide greater water flow through the annulus and thereby reduce the halide concentration in the water.
2
I L
iii) Identify the structural effect of long-term exposure of the B4C plate to boric acid, if applicable.
Response
This information is provided in the Carborundum Test Report No CBO-N-78-299 and was discussed with the NRC Staff on October 26, 1978.
iv)
If there has been swelling of the pool racks, determine the mag-nitude of the stresses in the neutron absorber plates considering the limiting loading condition for the vented geo:metry for the v) rack with the highest swelling and demonstrate that such stresses are acceptable.
This analysis should consider mechanical properties of the material including the effects of irradiation, temperature, and strain rate.
Response
The high density spent fuel racks at the Palisades Plant have not experienced any swelling to date.
Provide the results of Qualification testing which will demonstrat~
satisfactory material performance of the B4C plates in the spent fuel pool environment for the design life of the racks including the effects of irradiation, temperature, water, and boric acid.
Response
This information is provided in the Carborundum Test Report No CBO-N-78-299 and was discussed with the NRC Staff on October 26, 1978.
3
ATTACHMENT 2 Updated Safety Analysis for Vented Racks The original safety evaluation (dated November 1976 and titled "Spent Fuel Pool Modification Description and Safety Analysis") remains valid for the vented fuel rack condition except as noted below:
- a.
Section 3.1 should be changed as follows:
3.1 Spent Fuel Storage Rack Each fuel assembly will be stored in a concentric can roughly i2' long and with an inside square cross sectional length of 8.56".
Each storage cell will consist of two concentric 1/8" cans with neutron absorber plates installed in the annular gap between the cans.
The top and bottom of the two concentric cans will be closed with spacers and welded to provide an annulus within which the neutron absorber will be held.
A 1/4" diameter rod will be run the length of each corner of the annulus and welded in place to maintain the spacing between cans and to provide lateral support for the absorber plate.
A 3/8" thick fuel support plate will be welded at the bottom of the can to provide support for the fuel.
The plate will contain a 5" diameter hole to allow cooling water to flow upward through the fuel assembly to provide for removal of the decay heat from the fuel element.
The plate will also contain four 3/4" holes to accept the two fuel assembly alignment pins.
The top of each can will be flared slightly to facilitate fuel assembly insertion.
The rack will be constructed of stainless steel with the exception of the B4C absorber plates (see Figure 3-1).
The annulus contains a 3/16" vent hole located near the top of the annulus above the top of the neutron absorber plates.
This vent hole will allow any gas generated by irradiation of the neutron absorber plates to be relieved to the pool environment, thereby preventing any buildup of internal pressure in the annular cavity and preventing any distortion of the fuel rack can walls.
- b.
Section 3.2 should be changed as follows:
3.2 Neutron Absorber The neutron absorber plate is B4C powder bonded together in a carbonaceous matrix.
The absorber is 50% B4C by vollUile with the remainder being phenolic binder material and voids.
Specifications for the B4C powder used for the absorber plates will require that the median particle size by 125 microns by vollUile consistent with maintaining the criticality allowance for heterogeneity.
Specifi-cations also require that no more than 3% of the boron in the powder be in the oxide (B2o3) form.
This boron is not considered in the minimum boron loading considered in the criticality analysis.
The absorber is fabricated in 0.21" (minimum) thick plates.
These plates are inserted in the annular spaces formed by the concentric square cans.
The B4C plate is chemically inert in borated water and is thermally stable under all temperatures expected in the pool.
In. the unlikely event of a postulated seismic event, or significant vibrations, the neutron absorber plates are restrained from shifting by the spacers and the concentric tubes.
In the extreme case of the postulated seis-mic event and assuming the worst mechanical tolerances, there will be no settling of the neutron absorber material below the top of the active fuel.
The B4C plates are inserted and the spacers welded to form a protective envelope around the neutron absorbing material.
This annular envelope is vented to the pool environment to allow release of gas generated by the plates during irradiation.
The neutron absorber material used in these racks is of the same type approved for the Connecticut Yankee Rack Modification (Docket 50-213).
- c.
Section 3,7 should be changed as follows:
3,7 Material Compatibility Because the replacement racks, their associated hardware and the seismic restraints are of all stainless steel construction, as is the spent fuel pool liner, there is no potential for galvanic corrosion.
Material compatibility between the spent fuel pool and the new storage racks and between the fuel assemblies and the new storage racks is not a problem as stainless steel has been shown to be compatible with both fuel assemblies and spent fuel pool water.
Material compatibility between the boron carbide neutron absorber plates and the stainless steel spent fuel racks is excellent because the absorber plates, do not exhibit a gal-vanic potential with respect to the stainless steel storage cans.
The gas generated by the neutron absorber binder material, under irradiation, is primarily hydrogen and is reducing rather than oxidizing in nature.
The rate of hydrogen production, as deter-mined from measurements made during rack service in the Connecticut Yankee spent fuel pool, is approximately 1 in3 per cell per megarad of irradiation.
Since the maximum source strength of the spent fuel at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown is less than 2 megarads per hour, the maximum gas bubble generation rate is less than 2 in3/h (at STP).
Assuming that the bubbles generated have the same approximate initial diameter as the hole, this corresponds to a product.ion rate of about 5 bubbles per minute with an_initial velocity of approximately 0.01 inches per second.
This will have a negligible impingement velocity and effect on the stored bundles.
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- d.
The following should be added to Section 4.0:
4.0 Criticality Consideration Regarding the effect of water inside the cans, NUS has performed analyses on the WNP-2 and the Kewaunee projects which show in ~oth instances that Keff decreases when water is present in the poison annulus.
- e.
The following revisions reflect as-built conditions and were made to the safety evaluation report prior to the venting modification update.
3
Docket No.
- 50-255 CONSUMERS POWER COMPANY PALISADES NUCLEAR GENERATING STATION SPENT FUEL POOL MODIFICATION DESCRIPTION AND SAFETY ANALYSIS November 1976 Revised June 1977
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and is thermally stable under all tempera~ures expected in the pool..
Ln the unlikely event of a postulated seismic event, or I
significant vibr~tions, the neutron absorber plates are restrained f ram s,hif ting by t_he spacers and the concentric tubes.
In the extreme case of the postulated seismic event and assuming the worst mechanical toleta*nces, *there will be no settling of the neutron absorber material below the top of the active fuel.
The B4C plates are inserted ~nd the spacers seal welded to form an envelope around the neutron absorber material.
All these seal welds will be dye penetrant inspected.
The neutron absorber ma-terial used in these racks is of the same type approved for the Connecticut Yankee.Rack Modification (Docket 50-213).
3.3 Fuel Rack Assemblies The assembled cans are. formed into rack assemblies by attachment to rectangular itainless steel bars.
These bars are fabricated from Ni tronic-33, a _specialty austeni tic stainle~s steel having
- a yield strength substantially greater than type 304 stainless steel while possessing virtually identical corrosion resistance.
The bars run horizontally near the top and bo~tom of the rack assembly forming a unitized lattice arrangeme~t. Each can is continuously welded on all four sides to the lattice for~ing a single rigid structure, the rack assembly.
There are five dif-f~rent sizes of rack assemblies~ the size of each having been chosen to maximize the storage capacity of the pool (see Figure 2-1).
The racks are:
!YE£ A
B c D
E Spaces 10 x 5 10 x 6 8 x 8 6 x 5 10 x 5 In the main pool all the racks ate similar, having 8.56" square inner cans and a 10 1/4" center-to-center spacing. In the tilt 3-3 L.L.
{::I pit pool the E type rack designed* for: storage of* control rods as well as fuel, has a 9" square inner can~ and i~ arranged on a,
10.25" by 11.25" center-to-center spacing.
The two D racks in the tilt pit pool are similar to racks in the main pool.
Each of the racks is supported by four 'legs.
Each of the legs has the capability of being rdjusted to compens.ate for any tilt in the floor.
The leg supports are located between fuel cells and are reinforced with gussets.
The ~acks in the main pool are resfrained during a seismic event by compression-type restraints on the periphery of the array.
These restraints, in two rows, one near the top and the other near the bottom of the rack assembly, will be set during install-ation to have clearances to accommodate expansion due to temper--
ature changes of the pool water.* The maximum gap between the restraints and the pool wall will be approximately 0.3" and will accommodate a temperature increase from 70° to 2200F.
The racks in the tilt pit are provided with lugs that.will mate with keyways embedded in the pool side walls.
These keyways will have a clearance to allow for thermal expansion:
The keyway re-straints transfer the seismic load to the east ana*west walls.
A jib crane will be installed adjacent to the tilt pit to facili-tate fuel handling.
The north side of the tilt pit can serve as an alternate cask laydown area if "the 30 element rack is re:r.\\oved.
This capability will also extend plant operation for o~e more core cycle *if the need should arise.
Each rack is provided with four lifting lugs at the*top attached to the lattice bars that will allow the attachment of lifting hooks both for initial installation and, in the case of the A and and D type racks so they may be removed to permit installation of the shipping ~a~k, etc.
The-rac~s are designed only to be lifted while empty.
3-4 -------------------
3.4 Codes, Standaras, Rr.d Practices for Fuel* Assembly Rack Design, Construction, and Assembly The following are the codes, standards, and practices to which the "fuel assembly racks will be designed, constructed, and assem-bled.
(Revisions utilized are those in effect as of November 1, 197~)
- 1.
Design Codes
- 2.
- a.
AISC specification for the design, fabrication and erection of structural steel for buildings, 1969, including supplements 1, 2 and 3.
- b.
ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components (Tables I-7.0 and I-8.0 are used for allowable stress values for ma-
- . terials of construction)
- Ma'terial Codes
- a.
ASME Specification SA-240, Speci"fication for Stainless
- and Heat-Resisting Chromium and Chromium-Nickel Steel Plate Sh~et and Strip for Fusion-Welded Unfired Pres-sure Vessels.
- b.
ASME Specification SA-320, Specification for Alloy Steel Bolting Materials for L'.Jt.; Temperature Service.
- c.
AS:ME Specification SFA-5.9, Corrosion Resisting Chrom-ium and Chromium Nickel Steel Welding Rods and Bare Electrodes.
- d.
ASTM-A-312 Seamles~ and Welded Austenitic Stainless Steel Pipe.
3-5
Fuel rack handling will be made with the existing crane facili-ties *.
C~ane movement will be controlled by written administra-tive pr~cedures which will prohibit the movement of spent fuel' racks or dontrol rod racks directly over locations in the pool where* fuel assemblies ar*e being stored.
-:3. 7 Material Comoa tibili Lv i3eeaese th.e r.eplaeement ?'aelts 1 their aeeseiated hardwat:e *and tae
-tac BpEIIL feel peel liner, Lhete is no E1eteAtial fs:r 9al.ua.Ric car;-
.;osioR.
Mate:rial compaLibiliey eetween the spent tael pool aRQ...
t;he new stora~e raele:!!l and fietweea tag fugl. as.eetRhl.iea aad the ne~
.Ste:Ea9e racks is eilse HSt a prgelem as stainleps aeeel has been
-sagwa ~e be ee~atiale wit'b. both fnel *assembJ i es and speJ;lt ftiei s also tested in a a genera c ev-Se: e:
Re,_
1 c
(c* J i¢d 3-8
.I 1
6*
- i. **
I.*
4.0 CRITICALITY CONSIDERATION The racks in the main pool are designed for a 10.25 inch center-to-center spacing with B4C plates around each assembly.
The re~
sults of the criticality analyses are as follows:
4.1
- 1.
The center-to-center spacing of 10.25 inches between fuel assemblies with neutron absorber surrounqing each fuel assembly results in a k~ of 0.890 under nominal conditions.
- 2.
The worst case situations, considering maximum varia-tions in the position of fuel assemblies within the storage rack, neutron absorber positioning, variations in can dimensions, the most reactive temperature, cal-culational uncertainties arid worst case accidents result iri a k= of 0.928 with a confidence level of 95%
- Assumptions and Method of Analvsis The referenced set of calculations were based upon the following assumptions:
, 1.
Fresh fuel of 3.05 weight % U-235 nominal average enrich-ment
- 2.
Water temperature of 680
- 3.
No credit taken for soluble ~oison
- 4.
Fuel racks are infinite in three dimensions
- 5.
Control rods and ot~er fixed poisons ar~ not present in the fuel assembly.
4-1
L
'o f:1 differenc~ between the calculations and.experimental results was 0.009
~k. The KENO code, using the 123-group GAM-THERMOS cross section library, has been extensively benchmarked also.
For a series of ten critical experiments r~porteaC3} the average kef f as calculated using KENO and 123-group cross sections-was 0.9914 +.0020.
Using the same method, NUS has performe9 another ben~hmark on one of the Yankee critical experiments(4} with Ag-In-Cd cruciform control rods banked at 26 *. 37 cm from the bottom of the fuel.
The calculated keff was 1.008.+/- 0.006.
On the basis of the above comparisons with criticals, a calculational uncer-tainty of 0.008 ~k was assigned to the KENO calculations.
Also, statistical analysis of the Monte Carlo results shows
- a. standard deviation of.+/- 0.004, giving a 2a uncertainty of
"'"i 0.008
~k 00 Thus, an additional *o.ooa
~k uncertainty is as-signed to the KENO calculations
- The worst case criticality condition was obtained by using the maximum tolerances for the positioning of the fuel assemblies within the storage can as well as the relative can-to-can posi-tioning~ The rack tolerances are calculated on an overall rack width basis, such that cumulative tolerances between cans are accounted for.
The calculation was performed at a water temper-ature of 68°F.
The nominal boron carbide content of the neutron absorber plates is specified to be 0.0959 gm B-10/crn2 plate based on a 0.21 inch thickness.
(3}"Validation of Monte Carlo Calculations of Shipping Cask Systems" by L. M. Petrie and P. G. McCarty, ORNL, CONF 131101-14, 1973.
(4) 11 Yankee Critical Experiments...,.. Measurements on Lattics of Stainless Steel Clad Slightly Enri9hed Uranium Dioxide Fuel Rods in Light Water" ~y P. w. Davison, et al., YAEC-94 April, 1959, page 82.
4-4
f_
a t!!f
~
e
- e.
The unlikely case of one absorber plate missing from one side of one can in a group of 25 storage cclla was.calctilated.
The results.show that the increase in reactivity is 0.002.1k 00 (PDQ analysis).
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- 4.3 Worst Case Analysis of Tolerances and Calculational Uncertainties
- The following are* the results'of the KENO analysis of the worst case of tolerances and calculational uncertainties:
- Nominal Condi ti6ns, k00 Enrichment, 3.05%
Mechanical Spacing, 10~2~"
Pool Temperature,. 5gop B4C Partici~ Self-Shielding, ~koo Worst Tolerances, ~k 00 Boron Loading Variation Enrichment, 102% of nominal Loss of Poison Mechanical Tolerancea Pool Temperature (400F)
Worst Tolerances*
.0.872 0.004 0.004 o*.004.
0.002 0.023 0.003 I LL\\
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- An algebraic sum overestimates the effect of combiriing such taler-
~o~~
)
ances.
The.root mean square.of the first
~i.,:i;-~ tolerances, plus the ~col temperature tolerance~ may be more appropriate, which yields a total tolerance effect of approximately 0,02~~
4-5
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W16
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Calculational Uncertainties,~k=
KENO Benchmark Statistics (2a )
Total Calculational Uncertainties Max imu:i:n, koo Nominal, koo B4C Particle Self-Shielding, k 00
. Worst Tolerances Calculational Uncertainties Maximum, k 00.
4.4 Parametric Studies
= Q.,004
- 0. 036.
. 0. 016 0.056 0.008 0.008 0.016*
0.872 0.928 The base case, as established in the preceding sections, refers to the rack design with 10.25 inch spacing, 3.05 wt% nominal enrichment,* O. 0959 gm B-10/cm2 plate B4C loading a*na 68°F pool wate'r temperature.
The ko0 of the base case is 0.872 based on the 123-group KENO calculation.
Parametric studies were performed.. to
.determine the effect on k 00 of varying the base case conditions one at a time.
The results are presented below:
- 1.
k00 vs. Center-to-Center Spacing (KENO)
- (Nominal) 10.125" 10.250 11 10.375" 4-6
. +O.1256.koo (Base)
-0.02066.koo
I I.
I
~
2 *
- 3.
- 4.
- k. 00 VS. Enrichment (PDQ)
{Nominal)
(102%)
{110%)
k 00 VS. B-10
{Nominal) k00 VS. water (Nominal) 3.05 w/o U235 3.11* w/o U235 3.36 w/o U235 Loading (PDQ) 0.0850 gm B-10/crn2 0.0959 gm B-10/cm2 0.1050 gm B-10/cm2 Temperature (PDQ}
4-QOF 680F
\\
lQOOF 2120F 4.5 Accident Reactivity Analys~s (Base)
+O. 0036 ~k 00
+0.0173 ~k~
+O. 0039 ~koo (Base)
-0.0029 ~koo
+0.0031 6.koo (Base)
-0.0023 ~koo
-0.0243 6.k 00 Two fuel handling incidents were. analyzed:
(a) a fuel assernbiy drop during spent fuel handling landing horizontally on top of storage racks; (b) a fuel assembly inadvertently brought alongside the outer periphery of the storage racks in the vertical position if racks are removed (see Section 3.3).
For the case (b) accident, which is t:iC:: v;or st, the k of the pool was determined (PDQ) to be co about 0. 001 k 00 higher than the nominal k 00
- Thus, this situation affects reactivity only slightly.
4.6 Tilt Pit Reactivitv The tilt pool contuins a SO-storage c~ll special rack to accommo-date storage of control rods and fuel assemblies.
The geometry 4-7
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5.3 Structural Adequacy Using the previously listed loads and load combinations, stresses have been'calculated at critical sections of the rack.
The results of the structural and seismic analyses demonstrate that the spent feel racks are struc~u~ally adequate and meet the design criteria.
5.4 Pool Wall and Floor;Loadino The ability of the fuel pit and tilt pit floors and walls to with-stand the loads imposed by the new fuel racks containing spent fue*l was determined.
In addition to the dead loads imposed by the full loaded racks and the rack reactions* due to seismic loading*, hydro-static, seismic, thermal, sloshing and hydrodynamic forces were also considered.
F.orces an¢!. moments were obtained at sufficient po~nts on the reinforc~d concrete walls and floors.
These forces and moments were then combined in. accordance with the er i te.r ia *out-lin~d in Appendix A to the Palisades FSAR and the resulting quan-
. ti ties compared with the yield capabilities of the reinforced con-crete wall.
Based upon a conservative analysis, the minimum factor of safety (defined to be the ratio of the yield capacity moment to the load combination moment) was found to be 1.1.
The average factor of safety for the seven walls and ~wo floors analyzed was 4.5.
It is therefore concluded that all walls and floors comprising the main fuel pit and tilt pit.have adequa t_e margins of safety unoer the imposition of the new loads.
5-9 i
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