ML20207E379
| ML20207E379 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/27/1988 |
| From: | Boyle M Office of Nuclear Reactor Regulation |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8808180026 | |
| Download: ML20207E379 (55) | |
Text
.-
s MUL 2 71968 Cocket No: 50-245 PEMORANDUM FOR:
John F. Stolz, Director Project Directorate I-4 Division of Reactor Projects I/II FROM:
Michael L. Foyle, Project Manager Project Directorate I-4 Divison of Reactor Projects I/II
SUBJECT:
SUMMARY
OF JUNE 16, 1988 PEE 1ING ON MILLSTONE UNIT NO. 1 SPENT FUEL POOL RERACK l
By letter dated May 5,1988, Northeast Nuclear Energy Ccapany, the licensee, sutaitted inforn.ation to the NRC concerning a proposed increase in capacity of the spent fuel pool. This submittal addressed all tt.chnical areas (except for certain structural analyses) that the licensee anticipated would be necessary to support its proposed rerack.
On June 16, 1988 the staff and the licensee rret to discuss this sutaittal. A list of attendees is attached.
During this neeting, the licensee presented a general overview of the nodifications necessary to increase the spent fuel pool capacity frcrn 2184 to 3299 spent fuel assenblies, plus 20 defective fuel containers. Currently, the spent fuel pool has the capacity for the offload of approximately 2/3 of a core.
It has been the practice of the licensee to offload the entire core during refuelings. The increase in spent fuel pool capacity would provide full core offload capability through 1999.
In case of an c-n.ergency requiring the offload of the core before the spent fuel pool capacity is increased, the licensee has stated that it has a spent fuel stcrage rock in inventory which can be put in the cask lay-down area of the pool that will provide the necessary ttrporary storage capacity.
The licensee stated that in order to be finisted by the start of the next refueling octage, April 1989, spent fuel pool nodifications would have to connence by Septenber 1,19E8.
It was agreed upon by the staff and licensee to break the review process into three parts:
- 1) revlew of the seismic adequacy of the present spent fuel pool racks as free-standing racks (September 1,1988);
- 2) review of seismic, structural and civil engineering aspects of the installation of the new spent fuel pool racks (Novenber 1,1988); and
- 3) review of the remaining aspects (thermal hydraulics, criticality, etc.) and issuance of the license anandnent (April 1989).
The staff indicated that they wculd try to veet this anibitious schedule, hcwever, the staff directed that the licensee should suba:it its formal license arrendnent proposal by July 1,1988, together with the final structural analyses (a draft copy, attached, was provided to the staff at this reeting) and the licensee should develop alternative plans in case the staff can't neet the review schedule.
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8808180026 880727 PDR ADOCK 05000245 P
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Tte staff and the licensee agreed to ncet in the first week in July to discuss the anendnent proposal and to discuss phase 1 of the review, MW Michael L. Boyle, Project Panager Project Directorate I-4 Division of Reactor Projects I/II Attachnents:
as stated cc w/attachnents See next page DISTRIBUTION
' Docket File.
NRC & Local PCRs PDI-4 Ecading File JStolz FBoyle OGC-Pockville EJordan BGrires NRC Participar,ts:
PPoyle DJeng FRinaldi 0Fiero JKudrick RGoel LKopp HRichings ACRS(10)
HBClayton DA, LA PJI-4 y FM:PDI-4 PD:FC$1-J SNoItTI MEoyle JStoTz 07/j/88 07/)f/88 07/t]/88 i
. i
<6 Pr. Edward J. Mroczka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. 1 CC' Gerald Garfield Esquire R. M. Kacich, Manager Day, Berry and Howard Generation Facilities Licensing Counselors at Law Northeast Utilities Service Company City Place Post Office Box 270 Hartford, Connecticut 06103-3499 Hartford, Connecticut 06141-0270 W. D. Rorberg, Vice President D. O. Nordquist Nuclear Operations Fanager of Quality Assurance Northeast Utilities Service Corpany Northeast Nuclear Energy Corpany Post Office Bcx 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director Regional Administrator Radiation Control Unit Region I Department of Environmental Protection U. S. Nuclear Regulatory Commission State Office Building 475 Allendale Road Hartford, Connecticut 06106 King of Prussia Pennsylvania 19406.
Bradford S. Chase, Under Secretary First Selectren Energy Division Town of Waterford Office of Policy and Panagerent Hall of Records 80 Vashington Street 200 Boston Post Road Hartford, Connecticut 06106 Waterford, Connecticut 06385 S. E. Scace, Station Superintendent W. J. Raymond, Resident Inspector Millstone Nuclear Power Statien Millstone Nuclear Power Station Ncrtheast Nuclear Energy Corpany c/o U. S. Nuclear Regulatory Commission Post Office Box 128 Post Office Box 811 i
Waterford, Connecticut 06385 Niantic, Connecticut 06357 J. P. Stetz, Unit Superintendent Millstone Unit No. 1 Northeast Nuclear Energy Company Post Office Box 128 Waterford, Connecticut 06385 i
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Participants at June 16, 1988 Millstone Unit No. 1 Spent Fuel Pool Rerack Nare Affiliation RIUiael L. Boyle NRC David Jeng NRC Frank Rinaldi NRC Dan Fiero NRC j
dack Kudrick NRC Raj Goel NRC Laurence Kopp NRC Howard Richings NRC John Burnett NU i
Paul Blasioli NU Bill Vogel NU Thomas Mawson NU i
R. Bruce Roy NU 3
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6-34 6.12 FLOOR SLAB AND BUILDING ANALYSIS 6.12.1 Introduction This report describes the work performed by URS/ John A.
Blume & Associates, Engineers (URS/Blume),
for Northeast Nuclear Energy Company (NNECO),
under Purchase Order Number 865642.
The work described comprises the study to determine the structural capacity of the Millstone Unit 1 spent fuel storage pool to accommodate a new 5uel rack configuration.
The spent fuel storage pool at Millstone Unit 1 i s.
located in the reactor building 6etween elevations 65.75 ft and 108.5 ft, as shown in Figure 6.8.
The reactor building (and fuel pool) is a reinforced concrete structure in which lateral loads are primarily resisted by the perimeter walls and heavy concrete walls around the reactor.
The fuel pool sits on a 5 ft 4 in, thick concrete slab which is supported by the peri-meter wall and reactor wal)..
Figure 6.9 presents a plan of the fuel pool.
The pool was originally designed as described in the FSAR (Ref. 12).
The pool was later evaluated for a two-region consolidated fuel. loading configuration (Ref. 13).
In this configuration, some of the existing racks in the pool perimeter were to receive consolidated spent-fuel, thus increasing the total fuel storage capacity of the pool.
NNECO has now decided to have a new fuel rack.. configuration to accommodate more fuel than the original design.
The new fuel rack configuration is
~
in accordance with rigure 6.10 (Ref. 14), in, which new fuel j
r,acks are.added;in the perimeter of the fuel storage area.
The new rack configuttation results in increased fuel mass in the c*{
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Qb 6-35 pool.
This increased load is to be adequately carried by the pool and its supporting structure under various load condi-tions.
A parametric study was performed, based on the results of consolidated fuel load analyses, to determine the adequacy of the fuel pool to accommodate the new rack configuration.
This study also identified the critical load case to be con-sidered for the new fuel rack configuration evaluation.
Details of this study are included herein in section 6.12.6.
The current effort comprises performing thermal and seismic analysis of the fuel pool for the critical load case identified in the above-mentioned parametric study, and evalu-ating the ability of the spent fuel storage facility to accom-modate the new fuel rack configuration.
Sections 6.12.2 and 6.12.3 describe thermal and seismic analyses of the fuel pool.
Pool evaluation results are presented in Section 6.12.4.
Sum-mary and conclusions are given in Section 6.12.5.
Section 6.12.6 reports the results of the parametric study.
6.12.2 Dead and Thermal Lead Analysis This section describes the structural analysis er sults of spent fuel pool subjected to a dead load, which con-sists of the weight of the concrete, hydrostatic pressure and buoyant fuel load from the new rack cenfiguration, and a
thermal load from accident conditions.
A nonlinear static analysis is performed to compute the resulte.
The three-dimensional nonlinear finite element code ADINA (Ref. 15) is used, which considers the nonlinear behavior of concrete due to l
cracking, plastic flow, and the elastic-plastic behavice of the steel liner and reinforcement.
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6.12.2.1 Model Description A three-dimensional model of the fuel pool is devel-oped in order to perform a rigorous structural evaluation.
The pool is assumed symmetric about Section E-E, shown in rigure 6.9, so that only half of the structure is modeled.
rigure 6.8 illustrates the elevations in section E-E.
The finite element model includes the portion of the structure from the elevation of 52.85 ft, up to the top of the pool (elevation 108.5 ft),
and from the north wall to the reactor wall, rigure 6.11 identifies the pool model geometry, and Figure 6.12 presents the finite element mesh.
rigure 6.13 presents the mid-span section at the line of symmetry, rigure 6.14 presents the east j
wall elevation, and Figure 6.15 presents the finite element mesh of the fuel pool slab.
In the mathematical
- model, the concrete slab and walls of the fuel pool are modeled with 8-node isoparametric solid elements.
The fuel pool slab has five layers of solid elements across the depth (rigure 6.13) to rigorously simulate the stress distribution.
The top and bottom layers of the slab are modeled with 12-node isoparametric elements to facilitate placement of top and bottom reinforcement layers in the slab.
The steel reinforcement layers in the slab and the walls are codeled by two-dimensional isoparametric elements and truss elements.
Figure 6.16 presents a representative reinforcemer.t detail.
The liner plate inside the fuel pool is modeled with two-dimensional isoparametric elements.
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The model has a total of 2,951 nodes, 7,709 degrees of freedom, 38 element groups, 1,545 three-dimensional solid
,, c4 1
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thi 6-37 elements, 1,803 two-dimensional elements, 642 truss elements, and 9 beam elements.
For the three-dimensional solid elements, a sophis-ticated nonlinear concrete material model is prescribed.
The model employs three basic features to describe the material behavior, namely, (1) a nonlinear stress-strain relation in-cluding strain-softening to allow for the weakening of the concrete material under increasing compressive stresses, (2) failure envelopes that define cracking in tension and crushing in compression, and (3) a strategy to model the post-cracking and crushing behavior of the material.
The numerical solution allows for unloading and reloading of the material.
Figure 17 presents the stress-strain law and the failure envelopes for the concrete model prescribed in the ADINA program (Ref. 35 and 16) and used in the current analy-sis.
The stress-strain law is presented for the uniaxial stress condition for clarity.
It shows that when the stress exceeds a prescribed tensile strength, e,
the material will s
crack in a direction perpendicular to the stress direction and the stress will be reduced to zero for tensile strains beyond this point.
If the stress is, compressive, it will follow the nonlinear compressive stress-strain law until it reaches the compressive strength, e,.
The material becomes much softer at strains beyond this point and crushes at the ultimate strain, e,.
The stress is reduced to sero at strains beyond this point.
For multiaxial stress-state, the relationship of rigure 6.17b is used to determine the values of the parameters
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ars 6-38 of the stress-strain law of Figure 6.17a, such as, tensile strength ag, compressive strength o,, and crushing strength, c,, etc.
The failure envelope is based on the experimental work of Kupfor, et al. (Ref. 17).
The ma*.orial models available in ADINA to model steel reinforcement and steel liner behavior includes
- elastic, thermo-elastic, elasto-plastic, thermo-elasto-plastic, and bi-j linear models.
Previous analyses done for the consolidated fuel case (Ref. 14) indicated that steel reinforcements and the liner did not yield under the dead and thermal load conditions.
As such, the thermo-elastic model has been prescribed for steel liner and steel reinforcement models.
However, the computed i
stresses are checked to make sure that the stresses did not exceed the steel yield strength.
6.12.2.2 Loading The loads considered in this analysis are the weight of the concrete structure, the hydrostatic pressure on the pool floor and walls, the fuel buoyant weight, and the thermal loads due to temperature distribution.
The ambient temperatures of the surrounding areas and l
the nodal temperature distribution (for mid-span section),
determined from the heat transfer analysis for the accident condition, is illustrated in rigure 6.18.
The heat transfer analysis was carried out to determine the temperature distribu-tion corresponding to the accident condition in Reference 13.
This calculation assumed steady state conditions with radiation W
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boundary conditions on the pool walls, using reference tempera-tures and heat transfer coefficients corresponding to stagnant air or water, as appropriate for the particular wall.
Because of the very slow heating rate, the temperature field and the
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change in temperature from the reference (unstressed) tempera-I ture a't any load step was assumed propertional to this final l
steady state temperature distribution.
l The hydrostatic and varying fuel loads were applied as equivalent pressures on the steel liner which covers the pool floor and walls.
The hydrostatic pressure of the water for a 37 ft 9 in. depth is applied over the floor and varies linearly to zero up the pool walls.
6.12.2.3 Analysis Procedure 3
Nonlinear analysis of the fuel pool is conducted for D+T, load condition, where D is the sum of all dead loads and T,
is the postulated accident temperature distribution.
This load combination is applied to the model in several loading increments.
At each stage of the loading, iterations are taken with no additional lead to allow the load redistribution from 1
cracking to equalise.
A full Newton iterative solution scheme with line search option (Ref. 15) is adopted to obtain rigor-ously convergent solutions.
First, dead load is applied over several increments i
to allow any concrete cracking and load redistribution to 4
develop.
The analysis procedure utilizes a
sophisticated I
three-dimensional meterial model for concrete behavior (Ref. 15 1
e h(,Y dt I
t
,,, r
- 6-40 and 16).
This model allows cracking in any of the three ortho-gonal directions at an integration point in the element when the tensile stress in the particular direction exceeds the specified tensile strength.
When a crack develops, the load j
capacity of the element at that integration point is reduced to
]
zero in the direction perpendicular to the crack while retaining some shear capacity in the plane of the crack.
The loads are then redistributed to other elements or rebars during the next load increment.
If the crack closes (compressive strain across the crack face),
the element strength in compression is reinstated, but the direction is remembered so that subsequent crack openings occur in the same direction
]
under any tensile strain.
After the application of the total dead load, the thermal loads from the temperature distribution calculated in the heat transfer analysis are applied in increments with the dead load maintained on the structure.
These thermal loads are applied in increments to allow concrete cracking to develop.
The final step in the analysis is to calculate sec-j tion forces from the finite element integration point stresses i
for use in the structural evaluation of the pool slab and 1
walls.
I 6.12.2.4 Results Under total dead load, the pool floor deforms in a concave shape and produces tension on the slab.
As the thermal load grows in the pool, the pool floor develops compression due to the restraint of the east wall on the thermal expansion of 1
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bE4 6-41 l
l the pool floor.
In effect, because of the end walls, the ther-mal load acts as a prestressing mechanism for the floor.
An 1
examination of the deformed shape of the pool under dead and I
thermal loadings (rigures 6.19 and 6.20) illustrates the additional bending induced in the floor and walls due to the thermal loads.
This analysis clearly shows the importance of three-dimensional effects for the load carrying capacity of the spent fuel pool.
The back and interior walls are significant in eterying loads.
More importantly, these walls constrain the i
pool floor so that thermal loads prestress the floor.
This effect could r.o t have been captured from a two-dimensional
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analysis.
6.12.3 Seismic Analysis The purpose of the seismic analysis was to rtvise the existing dynamic models of the reactor building to reflect the new fuel rack configuration and determine the forces in the fuel pool structure due to seismic loading.
l 6.12.3.1 seismic Input i
l The three orthogonal input ground acceleration time-3 histories for the SSE case are provided by NNECO (Ref. 13).
The maximum horizontal peak ground acceleration for the SSE i
case is 0.2.
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6.12.3.2 Seismic Analysis Models Three models have been utilized to obtain the seismic forces in the fuel pool structures.
These are described in the next sub-sections.
6.12.3.3 Horizontal Dynamic Model The original lumped-mass stick model (rigure 6.21) of the reactor building was recreated by URS/Blume, in a previous
- project, for the consolidated fuel load analysis (Ref.
13) based on data provided by NNECO.
In that project, the model 4
had been updated as follows:
(1) mass of fuel at elevation 65.75 ft was revised to reflect increased mass due to con-solidated fuel; (2) fuel pool masses are revised to incorporate sloshing of 20 ft of water above fuel racks; and, (3) soil-structure interaction effects were considered by including frequency-dependent soil springs (stiffness) and dashpots (radiation damping) at the base of the model.
The model has 66 nodes, 90 doorees of freedom, and 34 beam elements.
The above-mentioned horizontal model has been updated by modifying the mass of fuel at elevation 65.75 ft to reflect increased mass due to the new fuel rack configuration.
This
{
model is then used to generate maximum horizontal floor acceleration responses at the fuel pool due to SSE horizontal motions.
6.12.3.4 vertical Dynamie Model The original vertical lumped-mass stick model of the reactor building was recreated by URS/Blume as part of the p
e O
o 4
6-43 consolidated fuel analysis project (Ref. 13) frJa data provided by NNECO.
This model had been extensively updated by URS/Blume in that project as follows:
(1) a detailed model of the fuel pool area, between elevation 65.75 f t to 108.5 ft, was devel-i oped and coupled to the lumped-mass vertical stick model.
This is shown schematically in rigure 6.22; (2) the pool floor slab and walls were modeled by plate elements as shown in Figure j
6.23.
The fuel racks inside the pool were each represented by a vertical single-degree-of-freedom (SDor) system corresponding to the first significant vertical mode of each rack; and, (3) ssI considerations were included by providing frequency-dependent soil springs (stiffness) and dashpots (radiation I
damping) at the base of the model.
The model has 475 nodes,-
1,162 degrees of freedom, 277 plate elements, 175 beam elements and 32 truss elements.
This model is updated in the current project for the new fuel rack configuration by adjusting the masses and stiffness of the SDor fuel rack models.
Then the model has been used to obtain maximum vertical seismic response accelera-tions at different locations of the fuel pool slab.
6.12.3.5 Three-Dimensional static ruel Pool Model A three-dimensional model of the fuel pool is used to i
combine the forces and moments in the fuel pool slab due to the vertical accelerations in the fuel pool obtained from the analysis of the vertical model described above and the seismic reactions at the fuel rack legs obtained from separate seismic analyses of the fuel racks.
This model includes the fuel pool structure from elevation 42.5 ft, up to the top of the pool w p*
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6-44
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(elevation 108.5 t), and from the north wall to the reactor wall.
Figure 6.24 provides a schematic representation of the model.
The model has 470 nodes, 1,292 degrees of freedom, and 284 shell elements modeling the fuel pool floor slab and walls.
The seismic accelerations obtained from subsection 6.12.3.4 are input as static loads in this model.
The vert'. cal seismic forces on the fuel pool, obtained from the seismic analysis of the fuel racks, are specified by NNECO to be 0.2 times the fuel rack weight in the vertical direction (Ref. 18).
These loads are also input to the model and static analyses are performed to obtain the seismic forces and moments in the pool, structure.
seismic forces and moments, due to the two orthogonal horizontal seismic accelerations, are computed separately and the co-directional resulte are added on a square-root-of-the-sum-of-the-squares basis.
These resultant seismic forces and moments are used for the pool evaluation.
6.12.3.6 Results from Seismic Analysis The fixed-base frequencies and mass participation factors for the horizontal and vertical models are presented in Tables 6.6 and 6.7.
The fundamental ver'.. cal frequency of the fuel pool slab is determined to be 19.5 Hz Fou Table 6.7.
The maximum horizontal accelerations in east-west and north-south directions at elevation 65.75 ft and 108.5 ft are presented in Table 6.6.
The vertical accelerations at different nodes of the fuel pool slab are presented in rigure 6.25.
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6.12.4 Load Combinations and Pool Evaluation 6.12.4.1 Load Combinations The load combinations considered are in accordance with those specified in NUREG-0800 (Ref.
19).
From these l
combinations, the vertical load cases are extracted in UWS/
Blume criteria document (Ref. 20).
The acceptance criteria is in accordance with NUREG-0800 and ACI 349-85.
Specifically, the following load combinations and acceptance criteria are to be used i
Load combination Allowables 1
I (a) 1.4D + 1.7L U
(b )
1.4D + 1.7L + 1.9E U
3 (b,)
0.75 (1.4D + 1.7L + 1.9c + 1.7T,)
U (c )
D+L+T,
+ E' U
3 (c )
D+L+T,
+ E' U
a i
where l
D
=
Dead load of structure which included weight of i
i the
- concrete, the hydrestatic pressure of the l
water, buoyant weight of fuel ricks, and cask L
=
Live load on structure
~ T, Thermal loads during normal operating or shutdown
=
conditions T,
Thermal loads under postulated break conditions
=
E
=
Loads generated by OBE l
E' =
Loads generated by SSE U
Ultimate strength
=
)
)
l l
- r.
l The load combinations, mentioned above, produce dif-ferent load demands on the fuel pool structural components.
However, examination of previous work (Ref. 13) on the Mill-stone 1 fuel pool for consolidated loadings indicates that two load cases produce the critical loads for the fuel pool struc-tures.
These are load case (ca) which produced the maximum moment d e '.-
the mid-span section of the fuel pool slab, while load (b ) produced maxin.um shear demands at the edge 1
of the fuel puol slab.
The fuel load configuration of the cur-rent study in different than that used in the previous study.
However, total fuel loads in both the cases are very close and combinations (b ) and (c,) are anticipated to remain the criti-3 cal load cases.
A parametric study has already been conducted by URS/
Blume to determine the effects of new fuel rack configuration on the pool slab for load case (b ).
This study was based on 3
new analyses as well as results extrapolated from the previous analyses for load case (b ) (Ref. 13).
The report describing 3
this study is presented as section 6.12.6 of this report to consider the effects of load case (b ).
The parametric study 3
also examines the load cases (a),
(b,),
and (c1), which are concluded to be not as critical as load cases (b ) and (c,).
1 Based on the above discussions, analyses and evalua-tion are conducted for load case ( e, ) for the new rack load configurations.
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l 6.12.4.2 Pool Evaluation Tabla 6.9 presents ratios of demand versus allowable at critical locations within the fuel pool structure for load case D + T,
+
E'.
These results indicate that shear and moment demands at various cri'tical sections of the fuel pool slab and walls art within the code allowables, except for bending moments at the fuel pool floor slab mid-span which rhows an overstress of 5%.
This does not represent a failure to rieet the acceptance criteria for the Millstone 1 fuel pool, because of concervatism in the analysis process as detailed in Section 6.12.5.2.
Additionally, the liner plate and embedments are evaluated and meet the code allowables.
The fuel pool slab is also evaluated locally for the impact loads from the fuel rack legs and m ets the code allowables.
Table 6.10 present's the ratios of demand versus allowables for critical locations within the fuel pool struc-ture fot load combination 1.4D + 1.9E.
For shear, the critical location is at the pool slab face close to the reactor wall, which shows an ovorstress of 1%.
For moment, the critical section is at the fuel pool mid-span, which has a demand /
capacity ratio of 0.78.
The demand / capacity ratios at other sections are lower and are not reported.
The shear overstress of 1% does not represent a f ailure to meet the neceptance cri-teria for Millstone 1 fuel pool because of conservatism in th:
analysis process as detailed in Section 6.12.5.2.
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6-48 6.12.5 Summary and conclusions 6.12.5.1 Summary This project has evaluated the capacity of the spent fuel pool at Millstone Unit 1 to support loads from the new fuel rack configuration.
The fuel pool evaluation is performed in accordance with applicable USNRC regulatory requirements and NNECO licens-ing commitments for Millstone Unit 1.
The pool is evaluated for rack and fuel loads, as established for the project by,
NNECO, hydrostatic and hydrodynamic loads, thermal loads for both operating pool temperatures and accident conditions due to a faulted pool and seismic loads from both OBE and SSE inputs established by NNECO for this project.
Rigorous nonlinear analyses, described in more detail above, are performed for dead and thermal loads.
Dynamic time-history analyses are performed for seismic loads.
The pool is then evaluated for these loads per NNECO regulatory requirements.
The pool is also evaluated for seismic rsaction forces on the pool floor from the sefsmic motion of the fuel racks.
These seismic reac-tion forces are specified by NNECO from a separate rack analy-sis effort.
Stressrs are calculated for the various load combina-tions for the steel ^iner plate and its attachments in the e nerete, for the concrete and for reinforcing steel.
These then compared to allowable stresses as defined in esses are
,salicable codes.
Stresses; for all load combinations are allowables in the liner place and its supports and in W
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6-49 the reinforcing steel.
Stressas in the con t.re te are also within allowables except for two locations.
For load case 1.4D + 1.9E, the demand shear is found to be 1% above allowa-bles for shear stresses at the pool perimeter wall.
For load case D + T,
+
E', the moment at pool mid-span is 5% over code 1
allowable moment.
These do not represent a concern for the Millstone 1 pool because of conservatisms in the calculation of allowable
- loads, as described in more der. ail in Section 6.12.5.2.
l 6.12.5.2 C_enclusions This study concludes that the Millstone fuel pool slab and walls fulfill applicable USNRC regulatory requirements and NMECO licensing commitments 'o support the new fuel rack configuration loads.
Stresses for the combined dead,
- live, temperature, and seismic loads were within code allowable stresses in all but two cases.
Shear stresses at the south boundary of the pool slab are 1% higher than code allowables for the factored load case combining 1.4 dead load plus 1.9 OBE seismic load.
Moment at pool mid-span is 5% over code allowa-ble moment for dead, accidental thermal, and SSE seismic load (D +
T,
+ E') case.
These calculated overstresses are con-sidered to be within acceptable limits because of several mit!-
gating factors as detailed below:
i a.
Shear 1.
The analysis increases dead loads by a factor of 1.4, even when the dead load used in the analysis includes the full weight of slab, concrete fill, I
9 9 (%
7
+
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,r-
=
6-50 steel liner, water, fuel racks, maximum consoli-dated fuel, and a full 150 kips cask loac.
In-creasing this total dead load by 40%, introduces substantial conservatisms to this evaluation.
2.
The allowable shear strens used in this study is that identified by the code for slender section (oommonly considered to be members with 10:1 span to thickness ratio).
The code increases the allowable shear stress by about 300% for deep sections which have a span to thickness ratio of 5:1.
Since the fuel pool slab has a thickness to span ratio of about 6:1, it is reasonable to accept a 1% increase in allowable shear stress.
3.
The 1% calculated shear overstress is within the numerical accuracy of the analysis procedures used in this study.
4.
The OBE peak ground acceleration (PGA) of.08g has been conservatively used in this analysis as compared to the.079 PGA used in NNECO FSAR.
b.
Moment 1.
The moment overstress of 5% indicates yielding of tensile steel reinforcement.
The yield strength of steel reinforcement is taken to be 40 ksi, while in reality, such reinforcement has higher yield values around 45 to 50 ksi.
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6-51 2.
The temperature loads considered a r e-conserva-tive.
Data provided in this repo.t, in section 5.1.3 and 5.5, indicate that the actual tempera-ture profile has lower values than those used in this structural analysis.
3.
The 5% calculated moment overstress is within the numerical accuracy of the analysis procedure used in this study.
6.12.6 Parametric Study for 1.4D + 1.9E Load Case 6.12.6.1 htroduetion The Millstone Unit 1 fuel pool had been evaluated by URS/Blume for a two-region consolidated fuel loading configura-tion (Ref. 6.12).
In this configuration, some of the existing racks in the pool perimeter were to receive consolidated spent-fuel, thus increasing the total fuel storage capacity of the pool.
NNECO has then decided to have a new fuel rack configur-ation (Ref. 6.14) to acconmodate more fuel than the original design.
The current study evaluates the capacity of the fuel pool to accommodate the stew fuel rack configuration for the load case 1.4D + 1.9E.
This evaluation is based on the URS/
Blume analysis results for the consolidated fuel loading.
Figure 6.26 prest nts the rack arrangements for the consolidated loading.
The hatched area contains existing racks which will receive consolidated fuel.
Figure 2.1 presents the new rack configuration.
Here, the existing racks are moved closer to the north and west pool walls by removing the lateral e
e' o, t.
a.
,v.,
..es 6-52 seismic restraints.
Then additioncl fuel racks are added near the east and south pool walls.
Though the load distributicn on the fuel pool from the two configurations are different, the total fuel loads are very close.
The fuel load for the con-solidated case is 2,625 kips, while that for the new rack con-figuration is 2,536 kips.
The total pool dead load, including weight of pool water, is 6,750 kips for consolidated fuel load-ing and 6,660 kips for new rack configuration.
6.12.6.2 Analysis Approach The objective of the analysis is to use the rigorous nonlinear analytical results for the consolidated load case td obtain those anticipated for the new rack configurhtion at different critical sections.
To achieve this, the following procedures are adopted.
First, two linear dead load analyses are made for the pool structure for the consolidated and new rack configuration.
The moments and shear at different sections obtained by the two analyses are compared and ratios are obtained.
Second, the moments and shears at different critical sections, obtained from the nonlinear analysis for 1.4D for the consolidated fuel laad case are adjusted with the dead load ratios of step one to obtain the estimated forces for tne new fuel rack configuration load case.
This is expected to give good estimates of forces snd moments for the new rack configur-ation beesuse (i) the total fuel loads are very close for the two cases, (ii) the total dead loads on the pool slab are very close, and (iii) the differences in moments and shears at dif-ferent pool locations., due to differences in fuel load distri-bution in the two configurations, are considgred by the ratios obtained from step 1.
i
@'M
~ - -,
e-
,.,e--
.v,,,..,
-.s.
w-
HAFT 6-53 Third, the seismic responses for the new fuel rack configuration are obtained from those calculated for the con-solidated fuel load case by using the mass ratio;.
Fourth, the fuel pool shears and momer.ts at critical locations due to dead and seismic loads are combined, and the pool is evaluated for the new rack configuration.
6.12.6.3 Three-Dimensional Static Fuel Pool Model A three-dimensional model of the fuel pool is devel-oped to obtain the ratios of moments and shears at different critical pool locations due to total dead loads from consoli-dated and new fuel rack configurations.
This model includes the fuel pool structure from elevation 42.5 ft, up to the top of the pool (elevation 108.C ft), from the north wall to the reactor wall, and from the east to the west wall.
rigure 6.27 provides a schematic representation of the model.
The model has 470 nodes, 1,292 degrees-of-freedom, and 284 shel1 elements
~
modeling the fuel pool floor slab and walls.
Dead load from each of the two fuel configurations at e input to this model and linear static analyses are con-ducted to obtain the shear and bending moments at different critical locations.
l 6.12.6.4 Pool Evaluation The fuel pool critical locations are evaluated for 1.4D + 1.9E load combination.
Table 6.11 presents the demand versus allowables at critical locations within the fuel pool structure for consolidated fuel loading (Ref.
6.12).
The
.._,m._.
..-,.,,,.~,.,~__.,,,,...c,
._,,.,,_.....,_,.-,.-,,,_.,___m,,,,
__.,_,,_.-._.4.
'e
?n' j
- IWI 6-54 demand versus allowables estimated for new fuel rack configura-tion are given in Table 6.12.
This indicates that all the demand moments and shears are within the allowables 1xcept for 1% overstress for shear at fuel pool south edge which is within the numerical accuracy of the solution.
The results also indi-cate that ratios for new fuel rack configuration are similar to those for consolidated fuel load configuration.
This trend is consistent with the fact that total fuel load fo. the two con-figurations are very similar.
6.12.6.5 conclusion Based on the results of the parametric study, it may' be concluded that the fuel pool structures satisfy the accept-an:e criteria for 1.4D + 1.9E load case.
The 1% overstress for shnar at pool south edge does not represent a failure to meet the acceptance criteria for the Millstone 1 fuel pool because of conservatisms in the analysis process as detailed in Section 6.12.
t 4
. ~..,.,,, - -, - -.. _ ~.
-,. ~ - -
t Di1:lket ri 6-55 TABLE 6.11 D2 SIGN EVALUATION
SUMMARY
LO/iD CASE:
1.4D + 1.9E CONFIGURATION:
CONSOLIDATED FUEL LOAD (REF. 1)
Shear Moment Item V
V V/V M
M M/M all all all all (psi)
(psi)
(k-ft/ft)
(k-ft/ft)
Fuel Pool Floor Slab 97 92 1.05 376 464 0.81 North Fuel Pool Wall Section G 68 102 0.67 258 321 0.80 Stress Item r
r r/r all all (psi)
(psi)
Stainless Steel Liner Plate 17,400 25,200 0.67 es
(,,
_, ~..,,
- n. p P '?
6-56 TABLE 6.12 DESIGN EVALUATION
SUMMARY
LOAD CASE:
1.4D + 1.9E CONFIGURATION:
NEW RACK 1
l shear Moment 1
Item V
y y/y M
g 373 all all all all (psi)
(psi)
(k-ft/ft)
(k-ft/ft) ruel Pool j
Floor Slab 93 92 1.01 360 464 0.78 North ruel Pool Wall i
Section G 69 102 0.68 263 321 0.82 m
k
._.,,_.___.,,,,..,m,,,...,,,,,___.,,
6-57 at.. -
=
22* 43.2S*_ _ _49.75*
90" h....
1
/
b 5
A
'A
- s g
~
~
- K1 N
fj
[
/
W'W{fWWW/T4W
~
u b
~
/
1
~~~
N A
4-+
Figure 6.26 Two-Region Consolidated Fuel Configuration j
.s 6-58 Y-
/ p ***(\\
%,%s
<p/
=%
/
/
s
~'
y
\\
)
),N,),,) <
$d
~
s
),,,
% /,i f
s
'%,:(*s,)-
(
e
'N
/
/
);.
~
/s
/
/
- ' p
'~s*% /,/,/,,,f,s
.,P,j
\\
- p ***
p
/
,n, Y, s '
a.
Finite Element Mesh of Pool Walls s
.'N*%
,,, /,,,p
/
/
f
,s s*%
/
'p%
N y
f,,,,
i -
y(
c j
N g s,<
/
+
/
s i
,4 1
g
$j:
s g
'p <, '(
M "
.5:
i
,; map,,<
y,.
/
/
b.
Finite Element Mesh of Pool Walls and Slabs rigure 6.27 Three-Dimensional Spatic Fuel Pool Model Cki ea
_, _ _ _ _ _ _. _ ~. - - -
y yW"*
,_,m"'"
,, _ _ _ _ - - " ' ~ ~ "
6-59 l
6.13 DEFINITION OF TERMS USED IN SECTION 6.0
{
S1, S2, 51, S4 Support designations pi Absolute degree-of-freedom number i I
g, Relative degree-of-freedom number i Coefficient of friction j
U Pool floor slab displacement time g
history in the i-th direction x,y coordinates horizontal direction z coordinate vertical direction K
Impact spring between fuel r
assemblies and cell K,
Linear component of friction spring K,
Axial spring of support leg locations N
Compression load in a support foot K,;
Rotational spring provided by the pool slab Subscript i When used with U or x indicates direction (i = 1 x-direction, i-2 y-direction, i = 3 z-direction)
-4
-.._r
.-r.
_ _ - -,.,, + _.--..,.,
w.,
6-60 6.14 REFERENCES l
6.1 USNRC Standard Review Plan, NUREG-0800 (1981).
6.2 ASME Boiler
& Pressure Vessel
- Code, Section
- III, Subsection NF (1983).
6.3 USNRC Regulatory Guide 1.29, "Seismic Design' Classification," Rev. 3, 1978.
j 6.4 "Friction coefficients of Water Lubricated Stainless Steels for a Spent Fuel Rack Facility," Prof. Ernest Rabinowicz, MIT, a report for Boston Edison Company, 1976.
6.5 USNRC Regulatory Guide 1.92, "Combining Modal Responses and Spatial Components in Seismic Response.
)
Analysis," Rev.
1, Februa y 1976.
1 6.6 "The Component Element Method in Dynamics with Application to Earthquake and Vehicle Engineering,"
S.
Levy and J.P.D.
Wilkinson, McGraw Hill, 1976.
6.7 "Dynamics of Structures," R.W.
Clough and J.
- Penzien, McGraw Hill (1975).
6.8 "Mechanical Design of Heat Excnangers and Pressure Vessel Components," Chapter 16, K.P.
Singh and A.I.
Soler, Arcturus Publishers, Inc., 1984.
6.9 R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Imaersed Solids," Journal of Engineering for Industry, Trans. of the ASME, February 1972, pp 167-172.
6.10 "Dynamic coupling in a closely Sp*;.J !>:-sody system Vibrating in Liquid Medium:
The Case of Fuel Racks,"
K.P.
Singh and A.I.
- Soler, 3rd International Conference on Nuclear Power Safety, Keswick, England, May 1982.
6.11 USNRC Regulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear Power Plants," 1973.
_....v--.
i UMM*l m.
6-61 1
6.12 NNECO, Millstone Nuclear Power Station Unit 1,
- FSAR, Section 12.0, Structural Design.
6.13 URS/Blume, Structural Evaluation of Millstone-1 Spent Fuel Pool for Consolidated Fuel Loading, URS/Blume Job No. 8586-02.
l 6.14 NUSCO transmittal no. RJS-1-19-88-1, R.
J.
Skwitz to S.
Bolourchi, dated January 19, 1986 6.15 ADINA Engineering, Automatic Dynamic Increment 01
{
Nonlinear Analysis, Users Manual, Report AE 84-1.
December 1984.
l 6.16 K.J.
Bathe and S.
Ramaswamy On Three-Dimensional i
Nonlinear Analysis of Concrete Structures, Journal j
Nuclear Engineering and Design, vol. 52, 1979.
j 6.17 H.
Kupfer, et al, Behavior of Concrete Under Biaxial'
- Stresses, Journal of American Concrete Institute, Vol. 66, 1969.
6.18 NUSCO Letter No.
PSE-SA-88-135, R.B.
Roy to R.J.
f Skwitz, dated June 8, 1988.
6.19
- USNRC, NUREG-0800 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Plants, LWR Edition, July 1981.
6.20 ACI 349-85, code Reqdirements for Nuclear Safety Related Concrete Structures, American Concr3te Institute.
0
._,_....._..,.~ _._.._ _... _ _ __. _. _ __._. _._ _. _.
em T AJ3LE 6. 6 EgEIZONTAL FREOUENCIES OF THE REACTOR BUILDING
_ Modal Particinatina Factora Mode Frequency Eni_
(CPS)
I-Direction Y-Direction 1
5.41 30.73 7.85 2
5.71'
-8.35 50.85 3
9.66 6.69 5.66 4
10.34 7.43
/
-5.84 5
10.36 5.17 6.66-6 16.58
-16.38 0.38 7
17.89
-0.43 22.32 S
20.60
-12.92
-2.90 9
21.48
-3.79 3.76 10 21.50 1.87 3.41 11 22.38 2.78
-4.21 12 23.36
-2.32
-3.09 13 27.84
-23.46 0.03 14 29.68 0.06
-19.48 15 29.88
-3.85
-0.50 16 30.49
-0.12 9.75 17 31.24
-6.57
-0.14 18 32.59 0.81
-0.13 19 33.13 0.11 1.76 20 33,21
-0.13 0.41 Mala: 2 - worth-south direction Y = East-Vest direction r*
p
M 2 6.7 YZRTICAL FREQUINCIES OF THE REACTOR BUILDING Mode Frequency Modal Participating Factors A
Efd__,,,
I-Direction Y-Direction L-DJrection 1
5.83 0.00
-4.98
-0.45 2
14.23 0.00
-0.01 51.88 3
19.57 0.00
-0.01 5.95 4
23.24
-0.00 0.10
-4.89 5
21.27 0.00 0.16 5.23 6
21.77 0.00
-0.19 19.18 7
22.68
-0.00 0.11 0.71 8
25.57
-0.00
-0.98
-0.21 9
25.77
-0.00 1.29
-0.05 10 26.06 0.00
-2.09 0.53 11 26.26 0.00 0.17 0.03 12 26.29 0.00
-0.16
-0.42 l
13 26.30 0.00 0.27 0.72 14 26.35 0.00 0.04
-0.08 15 26.36 0.00 0.08
-0.04 16 26.38 0.00
-0.02
-0.11 17 26.38 0.00 0.02
-0.08 18 26.39 0.00
-0.02
-0.02 19 26.39 0.00 0.00
-0.01 20 27.73 0.00
-0.72
-2.'01 4
21 22,90
-0.00 0.03 0.23 22 29.18 0.00
-0.01
-0.17 23 29.72 0.00
-0.16
-1.40 24 34.80
-0.00
-0.32
-0.72 Malt I = North-South directfon Y = East-West direction 2 = Vertleti directica
.45
$3 9
4 9
e em
.m 4*
e
- O 4 eene
_e_
_m eeG64 h *ee e e e f ese. e ce e emapam e
- m'
., 9 e
e O
8
.e 9
..,w..-----,,-,.
O p en TABLE 6.8 MAXIMUM EORIZONTAL SEISMIC ACCELERATIONS (SSE)
Maximus Acceleration Node 3 1 (N/S) 0.463 Y (E/V) 0.453 Node 10 X (N/S) 0.40s Y (E/V) 0.393 4
i e
G
)
j
+--~.-,_,.,.-.-n-,_,,__-.
_,.,,e-m.
n,_g,,,,,.,,,_,.,.,..e,,..,.e_.,..-,,.,---,.n.-,-gnp-w-,---,----,-o,----
---,gm--~w--.-v..
O 4
o
,5 e
TABLE 6.@
DESIGW EVALUATION
SUMMARY
LOAD CASE: D+Ta + E' Shear Moment I,tga V
Y,gg V M,gg M
M,gg M/M,gg
.[3g11
- J2111, tk.f t /f t,}.
(k.ft/ft1 Fuel Pool Floor Slab 107 115 0.93 936 898 1.03 North Fuel Pool Vall Section G 17 107 0.16 333 461 0.72 North Exterior Vall Section P 32 107 0.30 41 77 0.33 Stress D
I
- all F/F,gg
.f.RAll 12111 Stainless Steel Liner Plate 20,460 23.200 0.81 e
~ '. '.. ~
..7'
. ~. ~. - ~.. ~.....
.. ~...
i*
i i
T e&
TABLE 6.10*
DESIGN EVALUATION SUMM41X LOAD CASE:
1.4 D + 1.9 E Shear Moment I
I III all all all a11 123.11 12111 (k-ft/ft) ik-ft/ft) l Fuel Pool Flesr Slab 92.7 91.5 1.01 360 464 0.78
- From parametric study for dead and CBE loads f..s
'v T.~.' ~.*.*:
l
~^
" * :~
, ' ~ '
~ec e
aw-..
e,
,,,w
,-wgn,e--,e-,
---,-,.------,---,,.----,,,----,----------,-.--,n,-,
.------,--r-
Dhhgo\\
d D. 108.5 '
\\
INTERIOR WALL
. NORTH WALL REACTOR WALL i
Y i
SPINT FUEL POOL c. ss.08
\\
=
i POOL FLOOR IL. 52.85' REACTOR SYMMETRY U. 42.5' I
I I
i I
FIGURE 6.8 MILLSTONE 1 FUEL POOL and SEACTOR BUILDING (Section E-E) f'1I
- } {
~
l 1
i rV
)
NORTH
)
EAST WALL n
h
+
FIGURE 6.9 PLAN VIEW AT TOP OF POOL FLOOR bdb i
Wa,e 9
O
~ "
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--v-
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.w.,,.-.w. -.. -
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oW
' r,/-I
)
,a k
l
.a 1
Z i
r i
o i
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0 1
d a
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FIGURE 6.11, EAST HALF OF THE FUEL POOL AND SUPPORT STRUCTURES R* h 4
-~,.,-,,_,___._,-.y.
me I
I i
4: f>
fRj s;, )
dI i
b :::
- g ei g/
f.-
gj,9 s
4:
j g
p g
d s
,e 6
>l
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s i,
ldOd
>(
4 l
N
- ~-'
l
$( ;>.
lk h
( b f
dg#
c
- MfMJrrQpet m g g _
FIGURE 6.12 FINITE ELEMENT MESH 3-D THERMAL / DEAD LOAD MODEL 1
4 4
~
i
,-m-,w.,,-,,--,
n,n,,,,,rra,
,-n,,--,w--.,,-,,re,
__wn_,-__-
wen, n w,1r
.-.,--,,n,->--.._-.-n-w
. e.---en mem,,-,-w,,---.n-_-e-
,.8 e
i j
i I
l l
l g
O RERJ M e
0 l
M M
M M
s e
9 e
a eese.,8 m-g 9'
8' e*p e
M M e
e M
4 e'
8e e 4
4 FIGURE 6.13 FINITE ELEMENT MESH AT MIDSPAN SECTION E-E 3-D THERMAL / DEAD LOAD MODEL 9
e 9
l e
l 4
0 1
g
- M mm MM e
FIGURE 6.14 FINITE ELEMENT MESH OF EAST WALL 3-D THERMAL / DEAD LOAD MODEL 4
.., b gj\\D*
h i
+ - -. -,,,,, - - - - -,
,w.,_.w,...w.-3.-,,,...._..,_,-w_,__,-w,,_,,,_.,m mw-w_-,.,..,
-_.,-mm.,-
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DMb..
L.
l 4'
4 q
., ~
j r.
1 i
1
\\
- m w a run, pas.==. f h l
i FIGURE 6.15 FINITE ELEMENT MESH OF POOL FLOOR 3-D THERMAL / DEAD LOAD MODEL
- . a
-,, T 1
('\\(hd, -
\\
___,.-_,..m._
-_r...
._,,--._7 y.--.~_
I EL. ks t '
ei Olt' 1
1
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gif
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fl F./
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en 76, eiN eID p ta'
,a 1
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A
/
/
Y' gg ff IV i
.x 6 125 g
n PIGURE 6.'16 REINFORCE'4ENT DETAILS,- MIDSPAN SECTION E-E t
9 4
e u e
M
,-,,,y,
-}
'I/Is d
i coespasssivt stAANs. la#ts k%
-e m
.aeos
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l 3 e e t, i
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et '
te h*fb.o
- a. Nonlinear Uniaxial Stress-Strain law for Concrete l
( 4.o.01 J
t
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'~~~n
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/
go gp
/
i s
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a,i f
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7,'e t (kars
) l
- b. T u axial Failure
/
I Envelopes Reduced to
[/
2-D Conditions v-(0,0,()
~ ~ ~
'e g(t >e.?s
)(t.o.ts
)
- c. Triax al Tensne Failure 8'
Envelope FIGURS 6.17 NON1,1NEAR 3-D MATERIAL MODEL OF CONCRETE USED IN THERMAL /DE D LOAD ANALYSIS g
g,
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e FIGURE 6.18 TEMPERATURE DISTRIBUTION AT IfIDSPAN SECTION E-E
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FIGURE 6.21 SCHEMATIC REPRESENTATION OF HORIZONTAL SEISMIC DYNAMIC MODEL
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FIGURE 6.23 SCHEMATIC REPRESENTATION OF EAST HALF OF FUEL FOOL VERTICAL MODEL (COUPLED TO VER-TICAL REACTOR BUILDING MODEL) p.y;
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3., g gesM OF 3 D FUEL POOL 3
STATIC MODEL 3 f \\g 4
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asu 3.00' O'3200' 45 3N 46.33' l
(vertical accelerations SSE in g's)
FIGURE 6.25 PLAN OF FUEL POOL SLAB SHOWING VERTICAL SSE SEISMIC ACCELERATIONS
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