ML19339A807

Licensing Input on High Density Spent Fuel Racks for Fermi II Project.

ML19339A807
Person / Time
Site:	Fermi
Issue date:	10/24/1980
From:	Singh K, Sarah Turner JOSEPH OAT CORP., SOUTHERN SCIENCE APPLICATIONS, INC.
To:
Shared Package
ML19339A799	List: ML19339A807
References
OAT-J-2437, TM-586, NUDOCS 8011050294
Download: ML19339A807 (100)

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ENCLOSURE 2 LICENSING INPUT ON HIGH DENSITY SPENT FUEL RACKS FOR FERMI II PROJECT OAT J-2437 DECO P.O. 1A57079 BY K.P. SINGH, Ph.D VICE PRESIDENT, ENGINEERING JOSEPH OAT CORPORATION AND i A N@'4

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TABLE OF CONTENTS Page

1. PREFACE 1

2. GENERAL ARRANGEMENT 2

3. RACK CONSTP'iCTION 4

4. NUCLEAR CRITICALITY ANALYSIS 13 4.1 Design Baser 13 4.2 Geometric and Calculational Models 15 4.2.1 Reference fuel Assembly 15 4.2.2 Calculational Models 15 4 2.3 Reference. Fuel Storage Cell Configuration 16 4.3 Reference Subcriticality and Mechanical Tolerance Variations 17 4.3.1 Nominal Case 17 4.3.2 Boron Loading Variation 17 4.3.3 Storage Cell Lattice Pitch Variation 17 4.3.4 Stainless Steel Thickness Variations 18 4.3.5 Fuel Enrichment and Density Variation 18 4.3.6 Boraflex Width Tolerance Variation 18 4.3.7 Effect of Zirconium Flow Channel 19 4.3.8 Summary of Statistical Variations 19 4.4 Abnormal and Accident Conditions 19 4.4.1 Fuel Assembly Positioning in Storage Rack 20 4.4.2 Temperature and Water Density Effects 21 v

4.4.3' Abnormal Positioning of Fuel Assembly Outside Storage Rack 21 4.4.4 Dropped Fuel Assembly Accident 21 4.5 Summary 22 References 23 1

5. HYDRO-THERMAL CONSIDERATIONS 27 5.1 Design Basis 27.

5.2 Method of. Analysis 27 3.3 Results 29 References 30

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6. SEISMIC ANALYSIS 33 6.1 Analysis Outline 33 6.2 Fuel Rack-Fuel Assembly Model 35 6.2.1 Assumptions 35' 6.2.2 Model Description 36 6.2.3 Fluid Coupling 37 I

6.2.4 Damping 38 6.2.5 Impact 38 6.3 Stress Analysis 39 6.3.1 Stiffness Characteristics 39 6.3.2 Combined Stresses and Corner Displacements 40 6.4 Time Integration of the Equations of Motion 42 6.5 Structural Acceptance Criteria 44 6.6 Results 45 1

References to Secrion 6 46

7. MISCELLANEOUS ANALYSES 55

8. NEUTRON ABSORBER MATERIAL 56 l 8.1 Chemical Composition I

8.2 56 Physical Properties 57 9.

IN-SERVICE SURVEILLANCE PROGRAM FOR BORAFLEX NEUTRON ABSORBING MATERIAL 59 9.1 Program Design Intent 59 9.2 Description of Specimens 59 9.3 Long Term Surveillance 60 9.4 Accelerated Surveillance 60 9.5 Specimen Evaluation 61 9.6 References 62

10. DESIGN CONTROL AND FABRICATION INTERFACE 63

11. QUALITY ASSURANCE PROGRAM 68

12. PRODUCTION CONTROL 71 12.1 Synopsis 71 12.2 Procurement 71 12.3 Shop. Floor Planning 12.4 71-Operations Control & Coordination 72 12.5 Reporting 72 6

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1. PREFACE Enrico Fermi II Nuclear Power Station, owned by the Detroit

, Edison Company is scheduled for fuelloading in .1982. Some years ago the company initiated efforts to develop the design criteria for procuring high density fuel racks for storage of the spent l fuel. Stone & Webster Engi'neering Corporation was engaged to prepare the appropriate'" Design Specification", and develop all input data for the rack design; such as ground motion history l for seismic analysis, pool bulk temperature data for hydrothermal l- analysis, etc.

Joseph Oat Corporation, a Camden based supplier of nuclear plant equipment, was selected to design and fabricate the high density racks. This present document was prepared by oat in callaboration with their nuclear physics consultant, Southern Science Corporation (based in Dunedin, Florida) to give an abstract of all aspects of i rack design, analysis and fabrication. -Each item described here i

has been gleaned out of detailed reports and drawings in such a 1

manner as to pre.sent the physical facts without the encumbrance of mathematical details, or thousands of pages of unabridged results. The focus is on explaining the essential aspects of the design relevant to the equipment safety and reliability. The

! important features of the mathematical models to predict the l seismic response, hydrothermal behavior or the effective neutron multiplication factor are clearly stated to bring forth the built-

in conservatism and simplifications introduced into the analyses.

Abridged results are presented which show that in all analys,es,com-j fortable margins of safety in some cases wide margins are available .

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2. GENERAL ARRANGEMENT The high density spent fuel storage for Fermi II station pro-vides for a total of 2305 storage locations arranged in 14 modules. -Thirteen of these modules each contain 169 storage cells. The fourteenth module (labeled B-1 in Figure 2.1) has 108 cells. All the modules (also hereafter referred to as storage racks) are free standing; i.e: they are not anchored to the pool floor or connected to the pool wall through snubbers or lateral restraints. The minimam gap between adjacent racks is 3.625" at all locations. Sufficient gap is also maintained i between the modules'and the pool walls. The minimum gap between the fuel pool wall and rack modules is 24" (ref. Figure 2.1).

Adequate clearance-from other pool resident hardware is also provided. In this manner, the possibility of inter-rack impact, or rack collision with other pool hardward during the postulated ground motion events is precluded.

Of the 2305 storage locations, 2303 locations are intended for spent fuel storage. Two locations are earmarked for tr e neutron absorber material surveillance program; and also for in-service inspection of the rack structural welds. A comprehensive in-service inspection program for the neutron absorber material and structural welds has been developed to enable a programmed vigilance of vital material properties.

In addition to the spent fuel storage locations, a. rack for storing defective cells, control rods, control rod guides is also provided. It contains 31 storage locations for defective fuel storage containers / control rods, and 4 locations for control rod guide tubes. It is labeled as C-1 in Figure 2.1.

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3. RACK CONSTRUCTION I The racks are constructed trom ASTM 240-304, austenitic steel sheet material, ASTM 204-304 austenitic steel plate material, and ASTM 182-F304 austenitic steel forging material. Boraflex, a patented brand name product of Bisco

• serves as the neutron absorber material. The detailed radiological properties of Boraflex may be found in Section 4 and Section 8.

A typical module contains storage cells which have 6" minimum

(+0.125", -0") internal cross-sectional openis.gs. These cells are straight to within 0.125". These dimensions ensure that fuel assemblies with maximum permissible out-of-straightness can be inserted into the storage cells without interference.

Figure 3.1 shows a horizontal cross-section of an array of 3 x 3 cells. As mentioned in the preceding section, a typical module has an array of 13 x 13 cells.

The cells provide a smooth and continuous surface for lateral contact with the fuel assembly. The construction of the rack modules may be best described by exposing the basic building blocks of this design, namely.the " cruciform", " ell" and " tee" elements, shown in Figure 3.2. The cruciform element is made of 4 angular sub-elements, "A" (Figure 3.3) with the neutron absorber material tightly sandwiched between the stainless sheets.

The long edges of the cruciform are welded using a 3/16" thick ste.inless steel backing strip as shown in Figure 3.4. The bottom of the cruciform assembly has 5" high stainless strips, which ensure against slippage of the " poison" material downwards due to gravitation loads or operating conditions. A number of cross rivets across the cruciform walls provide the redundant safe-guard against slippage of the " poison" sheets. The top of the cruciform is also end welded using a spacer strip as shown in Figure 3.4. Skip welding at tne top ensures proper venting of the sandwiched space in the cruciform spokes.

• Bisco, a' Division of Brand, Inc., 1420 Renaissance Dr. Park Ridge, Illinois l

The " ell" and " tee" elements are constructed similarly using angular sub-element "B", and flat sub-element "C" (Figure 3.5).

Having fabricated the required quantities of the " cruciform",

" tees" and " ells", the assembly is performed in a specially designed fixture which serves the vital function of maintaining dimensional accuracy while welding all the contiguous spokes of all elements using fillet welds. Figure 3.6 shows the fillet welds in a 4 x 4 array. In this manner the cells are produced which are bonded to each other along their long edges, thun in effect, forming an " egg-crate".

The bottom ends of the cell walls are welded to the base plate which has 5.25" diameter holes concentric with cell center lines.

Carefully machined sleeve elements are positioned in the base plate and attached to the base plate through circular fillet welds (Figure 3.7) . The conical machined surface on the sleeve provides a contoured seating surface for the " nose" of the fuel assembly. Thus, the contact stresecs at the fuel assembly nose bearing surface are minimized.

The central hole in the sleeve provides the coolant flow path for

.. cat transport from the fuel assembly cladding. Lateral holes in the cell walls (Figure 3.7) provide the red 2ndant flow path in the unlikely event that the main coolant flow path is clogged.

The rack assembly is typically supported on four plate type supports. Where floor obstruction, (such as an existing sump) interferes with symmetrical support positioning, additional supports at the most desirable available location are provided. 'rne supports elevate the module base plate 7.5"above the pool floor level, thus creating the water plenum for coolant inventory.

As shown in Figure 3.7, a crush allowance of 1" above the fuel l assembly bail is provided to protect the stored fuel assemblies from accidentally dropped objects.

A brief description of the criticality analysis now follows.

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4. NUCLEAR CRITICALITY ANALYSIS 4.1 Design Bases ,

The spent fuel storage racks are designed to assure that a k eff not greater than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity and flooded with unborated water at room temperature (68 F). The calculated reac-tivity includes a margin for uncertainty in reactivity ca'lculations t

and in mechanical tolerances, statistically combined, such that the true k eff will be less than 0.93 with a 95% probability at a 95%

confidence level. Applicable codes, standards and regulations or pertinent sections thereof include the following:

e General Design Criterion 62 - Prevention of Criticality in Fuel Storage and Handling.

• NRC letter of April 14, 1978, to all Power Reactor Licensees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, e NRC Standard Review Plan, Section 9.1.2.

e Regulatory Guide 3.41, Validation of Calculational Method for Nuclear Criticality Safety (and related ANSI N16.9-1975).

e ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.

Conservative assumptions used in performing the criticality calcu-lations for the spent fuel storage rack include the following:

e Standard BWR fuel configuration.

e Uniform average enrichment of 3.2 wt% U-235 (maximum),

corresponding to an average U-235 fuel assembly loading of 15.49 grams per axial centimeter, e Spent fuel storage rack will accommodate, with the required subcriticality, either unchanneled (fuel bundle) or channeled (fuel assembly) fuel with maximum expected distortion.

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ZlRCONIUM FLOW CH ANNEL O.10" THlCK,5.48" OU TSIDE DIMENSION BULGED ZlRCONIUM FLOW CHANNEL 5.93" OU TSIDE DIMENSION 6.22" OUTSIDE CELL DIMENSION Fig ..4.1 Geometric model of Fermi spent fuel storage rack cell. ,

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e No Gd burnable poison present in the fuel.

e Lattice of storage. racks is infinite in all directions, i.e., no credit for axial or radial neutron leakage.

e No neutron absorption in minor structural members, i.e.,

spacers and Incenel springs replaced by water.

o Subcriticslity maintained by neutron absorbers of Boraflex material separating each fuel assembly.

e Pure zirconium used for claddin g and flow channel, neglecting neutron absorption 5, impurities in Zircaloy.

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! 4.2.1 _ Reference Fuel Assembly The BWR fuel assembly and dimensions used for the analysis is shown in-Fig. 4.1. A maximum uniform U-235 enrichment of 3.2 wt%

was assumed corresponding to an average axial loading of 15.49 grams U-235 per axial centimeter in each fuel assembly. In the unpoisoned i

standard-reactor-core geometry (6.00 inch assembly pitch), the AMPX-KENO calculated k gg is 1.364 0.004 (la) with unborated water at i 20 C. Each assembly is an 8 x 8 array of fuel rods with two of th a central rods replaced by a zirconium " water-rod." The square Zircaloy flow channel surrounding the fuel is 0.100 inches thick and 5.48 inches overall dimension, with a maximum distortion (bowing) to 5.93 inches.

4.2.2 Calculational Models Nuclear criticality analyses of the high density spent fuel storage

rack were performed with the AMPX -KENO computer package, I using the-123 group XSDRN cross-section set, and the NITAWL sub-routine for U-238 resonance shielding effects (Nordheim integral j . treatment). AMPX-KENO has.been extensively benchmarked against a i number of critical. experiments (e.g., references 3, 4 7nd 5). The water gap.between fuel assemblies is approximately the same as the 1

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critical configurations analyzed in the benchmark calculations and therefore no correction for water gap thickness is necersary.

Results of benchmark calculations (5) on a series of critical experiments with thin slabs of boron-containing absorber material between fuel assemblies indicate that the AMPX-KENO calculations conservatively over-predict k eff f r boron absorber loadings above

.a boron " worth" of about 15% ak. Since the boron worth in the Fermi high density spent fuel rack is about 40% ak, the AMPX-KENO calcu-lational model will over-predict k by a significant margin (as eff much as 3% ak, 95% confidence limit, based upon extrapolation of the trend analysis). For conservatism, no credit is taken for this over-prediction and the calculational bias is taken as zero.

For investigation of small reactivity effects (e.g., mechanical tolerances), a four-group diffusion / blackness theory method of analysis (NULIF-CNROD-PDQ7) was used (Reference 5) to calculate small incremental reactivity changes. This model has been used previously with good results, although the absolute value of k eff is not significant since this model is normally used only to evaluate small incremental reactivity effects (trends) that would otherwise be lost in the KENO statistical variation. In some cases, trends calculated by AMPX-KENO and by diffusion / blackness theory were compared and found to be in good agreement, well within the statistical uncertainty of KENO calculations.

4.2.3 Reference Fuel Storage Cell Configuration The nominal spent fuel storage cell model used in the criticality analyses, is shown in Fig. 4.1. The rack is composed of Boraflex absorber material sandwiched between two 0.075 .002 inch stainless-steel plates. The fuel assemblies are centrally located in each storage cell on a nominal lattice spacing of 6.22[0.125 9 inches.

For two-dimensional X-Y analysis, a zero current (reflecting) boundary condition was applied in the axial direction and at the centerline tP' agh the Boraflex absorber on all four sides of the cell, effectively cruating an infinite array of storage cells.

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The Boraflex absorber is nominally 0.070 .007 inches thick containing 0.01728 0.00173 grams B-10 per em 2 areal density.

4.3 Reference Subcriticality and Mechanical Tolerance Variations 4.3.1 Nominal Case Under normal conditions, with nominal dimensicas, the calculated k gg is 0.9224 0.0034 (lo, average of two independent AMPX-KENO runs with a different set of random numbers).

4.3.2 Boron Loading Variation The Boraflex absorber plate is nominally 0.070 inches thick with a B-10 areal density of 0.01728 g/cm . Manufacturing tolerance limits are 10% in either thickness or boron content. This assures that at any point where the minimum boron loading 2

(0.0155 g B-10/cm ) and minimum Boraflex thickness (0.063 inch) may coincide, the boron areal density will not be less than 0.014 g B-10/cm .

Calculations were made of k gg with variation in Boraflex absorber loading and thickness. Results of these calculations, given in Fig. 4.2, indicate that the k gg can be described by the following regression fit (least squares) to the data over the range of B-10 loading from 0.010 to s0.020 g/cm 2,

--in k gg = 0.06494 in (B-10, g/cm ) + 0.3440 Within the precision of the calculations, this relationship indi-cates that 10% tolerance limit on either boron content or Boraflex thickness results in the same incremental reactivity change of 0.0063 ak. The trend calculated by both AMPX-KENO and by diffusion /

blackness theory is the same within the analytical uncertainty.

4.3.3 Storage Cell Lattice Pitch variation The design storage cell lattice spacing between fuel assemblies is 6.220 0.

.00 finches. Increasing lattice pitch from 6.220 inches to

O e 6.345 inches (maximum tolerance) reduces reactivity by 0.0113 i0.006 ak, as calculated by AMPX-KENO or by 0.0094 ak calculated by diffusion / blackness theory. Thus, the nominal case exhibits the maximum reactivity and the effect on reactivity of a lattice pitch increase is negative. A larger increase in lattice pitch produces an even larger negative effect. Results of calculations at several lattice spacings and boron loadings are shown in Fig. 4.3 in terms of the overall fuel region volume fraction in the spent fuel storage cell (0.6775 for the nominal design) .

4.3.4 Stainless Steel Thickness Variations The nominal stainless-steel thickness is 0.075 .002 inches. The reactivity effect of the stainless steel thickness tolerance varia-tion was calculated to be .0005 ok by difftsion/ blackness theory, since the reactivity increment is too small to be calculated by AMPX-KENO.

4.3.5 Fuel Enrichment and Density variation The reference enrichment, 3.2 wt% U-235 or 15.49 grams U-235 per axial centimeter in each fuel assembly, represents the maximum U-235 loading. Therefore, any deviations in U-235 enrichment would result in reduced reactivity. Calculations of the sensitivity to small enrichment variations by diffusion / blackness theory yielded an average coefficient of 0.0075 Ak per 0.1 wt% U-235 in the range g

from 3.1 to 3.3 wt% U-235.

Calculations made with the UO 2 fuel density reduced from 10.41 to 3

10.25 g/cm indicate that the storage rack k gg is reduced by 0.0002 ak (diffusion / blackness theory) .

4.3.6 Boraflex Width Tolerance variation To preserve symmetry, the nominal Boraflex absorber width in the calculational model is 5.5 inches. The calculational model (Fig. 4.1) in conservative since in the actual storage cell, the Boraflex absorber joins in opposite corners and is 5.86 inches wide.

e e~, ===e- o e. -e>===+e e*-***w*e . - . .= e-*

_19 The nominal design (calculational model) thus results in the maximum reactivity since any increase in absorber width would decrease-reactivity. An AMPX-KENO calculation with the Boraflex absorber plates described in the actual geometry (5.86 inch width) indicates the calculational model under-predicts k gg by 0.005 .006 ak.

4.3.7 Effect of Zirconium Flow Channel Elimination of the zirconium flow channel resulted in a small negative increment in reactivity as calculated by diffusion / black-ness theory ( .005 ak) and a negligible reactivity increment by AMPX-KENO (+0. 0 0 2 .006). More significant is a small positive reactivity effect resulting from distortion (bowing or bulging) of the Zr flow channel which moves the channel wall outward toward the Boraflex absorber. For the maximum bowing (to 5.93 inches outside maximum dimension) uniformly throughout the assembly, an incremental reactivity of +.005 ak results, as calculated by diffusion / blackness theory using the approximate geometric model for the flow channel indicated by the dotted lines in Fig. 4.1.

This approximation should conservatively overestimate the reactivity effect of flow channel distortions since actual bulging would not normally be the maximum uniformly throughout the assembly- Since this is not a statistical variation, but could potentially add 0.005 ak (maximum), it is not combined statistically but is con-servatively accounted for additively in Section 4.5.

4.3.8 Summary of Statistical Variations Calculated reactivity increments from mechanical and fabrication tolerances are summarized in Table 4-1.

4.4 Abnormal and Accident Conditions Although credit is permitted for absorption by other absorbers under abnormal conditions, the following evaluations were made without any additional absorber material in the spent fuel storage rack. To the extent any additional absorbers may be present in the realistic case, the following analyses are even more conservative.

w.

l Table 4-1 CALCULATED STATISTICAL VARIATIONS IN REACTIVITY Case Tolerance Incremental Reactivity, Ak Boron concentration 10% i.0063 Boraflex thickness 10% .0063 Lattice pitch -0.000 Zer to negative inch

+0.125 SS tolerance i.002 inch .0005 Fuel enrichment and Zero to negative density Boraflex width -0.000 inch Zr to negative

+0.0625 KENO statistical .0075 maximum (95%/95%)

variation Statistical average .0117 (Root-mean-square of

, positive increment)

)

Calculated as Ka, where e is the standard deviation of the mean and K is the two-sided tolerance factor (2.21) for 95% probability at a 95% confidence interval with 114 genera-tions in the KENO calculations.

4.4.1 Fuel Assembly Positioning in Storage Rack The fuel assembly is normally located in the center of the storage rack cell with bottom fittings which mechanically prevent lateral movement of the fuel assemblies. Nevertheless, calculations were made with the fuel assembly assumed to be located on one side of the cell with the flow liner touching the SS-Boraflex' plate. For this case, the calculated reactivity (0.9213 .005) was slightly less than the nominal design case. Calculations were also made with the fuel assembly moved into the corner of the storage rack cell, resulting in an even larger negative reactivity effect (cal-culated k gg of 0.9126 .005). Thus, the maximum reactivity results for the nominal case with the fuel' assembly positioned in the center of the storage rack cell.

4

_ .l

4.4.2 Temperature and Water Density Effects Increasing the water temperature or introducing voids 'to simulate boiling) decreases reactivity as indicated in the following table of calculated reactivities.

Table 4-2 EFFECT OF TEMPERATURE AND VOID ON CALCULATED REACTIVITY OF STORAGE RACK Case Calculated kg , or ak gg Reference, 68 UF 0.9224 .0034 (AMPX-KENO) 176 F -0.0041 ok gg (AMPX-KENO) 212 F -0.0209 ok (Diffusion / blackness theory) 0 212 F.with 50% void -0.185 ok (Diffusion / blackness theory)

Decreasing temperature from the nominal 68 to 39 (maximum water density) is calculated (diffusion / blackness theory) to increase reactivity by 0.0006 ak.

4.4.3 Abnormal Positioning of Fuel Assembly Outside Storage Rack Since the storage rack criticality calculations were made assuming an infinite array of storage cells, positioning a fuel assembly outside and adjacent to the actual rack cannot add reactivity, but would result in a lower k than the k gg calculated for the eff infinite array.

4.4.4 Dropped Fuel Assembly Accident A postulated fuel assembly drop through and into a storage rack cell results in a configuration that is just that of the nominal cell configuration and therefore will not result in a reactivity greater than the nominal design case.

To investigate the possible reactivity effects of other postulated drop accidents, calculaeions were made for unpoisoned assemblies separated only by water. Figure 4.4 shows the results of these ,

i l

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calculations. From these data, the reactivity (kgg ) will be less 4

than 0.95 for any spacing greater than s8 inches. For a straight  ;

drop on top of the rack, an inclined drop or a fuel assembly lying horizontally on the top of the rack, the minimum separation distance is s18 inches. Thus, rod drop accidents will not result in an increase in reactivity above that calculated for the infinite nominal design storage rack.

4.5 Summary The criticality analyses of the spent fuel storage rack under .

normal and abnormal conditions are summarized in Table 4-3.

Table 4-3

SUMMARY

OF CRITICALITY CALCULATIONS Case k gg or ak oo Comment k gg , nominal case 0.9224 Worst combination of +0.0117 ak From Table 2-1, includes

stat.istical variations KENO statistical variation Calculational bias O Conservative Zr flow channel dis- +0.005 Ak Maximum, model bias of tortion or removal .005 Ak not included Increased t1mperature Negative or void Decreased temperature +0.0006 Ak Maximum water density Fuel element posi-tioning in storage cell Negative

, Fuel handling accident Negligible Maximum k gg 0. 9 ~,9 7 The k f 0.9397 is thus conservatively estimated to be the maxi-oo mum k gg under the worst combination of calculational and mechanical uncertainties, with a 95% probability'at a 95% confidence level.

t l

\

REFERENCES

l. Green, Lucious, Petrie, Ford, White, Wright, PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating coupled Multigroup Neutron-Gamma Libraries From ENDF/B, ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.

4

2. L. M. Petrie and N. F. Cross, KENO-IV, An Improved Monte Carlo Criticality Program, ORNL-4938, Oak Ridge National Laboratory, November 1975.

3. S. R. Biermanetal.,gg{ticalSeparationBetweenSubcritical Clusters of 4.29 wtt U Enriched UO, Rods in Water with Pixed

Neutron Poisons, NUREG/CR-0073, Batteile Pacific Northwest Laboratories, May 1978, with errata sheet issued by the USNRC August 14, 1979.

4. M. N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock & Wilcox Company, July 1979.

5. S. E. Turner and M. K. Gurley, Benchmark Calculations for Spent Fuel Storage Racks, Report SSA-127, Southern Science Applications, Inc., July 1980.

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REFERENCE ,

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1 Fuel Assembly Separation (Pitch), inches Fig. 4.4 Reactivity effect of separation between fuel assemblies (unpoisoned) .

1 1

l

5. HYDRO-THERMAL CONSIDERATIONS A central objective in the design of the high density fuel rack is to ensure adequate cooling of the fuel assembly cladding. In the following a brief synopsis of the design basis, method of analysis and computed results is given.

5.1 Design Basis The maximum bulk temperature of the pool water is assumed to be 150 F (the pool water bulk temperature ranges from 85 to 150 F per ref. 1, page 25). The decay heat gener-ation in the pool due to stored spent fuel is based on the following two conditions:

a. A discharge of up to 260 fuel bundles with 30,000 mwd /T exposure with 30-day cooling; assume 764 storage loca-tions are empty and the remainder filled with fuel discharge annually at 191 fuel bundles per year with 30,000 mwd /T exposure.

b. Full core discharge (764 fuel bundles) after 288 hr.

decay, 180 days after the last 1/4 core discharge; spent fuel at equilibrium conditions. A fuel pool cooling water flow rate of 550 gpm will be augmented by a 2000 gpm RHR flow. A 150 F maximum restriction on the fuel pool bulk temperature shall govern for the case of a full core discharge.

The fuel pool cooling system has been shown (ref. 2) to be capable of maintaining the pool bulk temperaturc within the aforementioned limits.

The design criterion requires that the full assembly cladding temperature does not exceed the coincident coolant saturation temperature at any location in the pool at any instance of time.

5.2 Method of Analysis In order to determine an upper bound on the maximum fuel cladding temperature, a series of conservative assumptions are made. The most important assumptions are listed below W8De**M*h$ e=wh we a qq. m9 w

a. As stated in the design basis above, the fuel pool will contain spent fuel with varying " time-after-shut-down" ts. Since the heat emission falls off rapidly with in-creasing ts, it is obviously conservative to assume that all fuel assemblies are fresh (ts = 12 days), and they all have had 180 days of operating time in the reactor. Thus, the heat emission rate of each fuel assembly is equal, and it can be computed from ref. (3).

b. There are 2303 available fuel assembly locations as shown in Figure 2.1.

In addition there are 36 defective fuel container locations. These occupy an irregular floor space in the pool. For purposes of the hydrothermal analysis, a circle circumscribing the actual rack floor space is drawn. It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal pitch of 6.22". (Figure 5.1) c.

The downcomer space around the rack module group varies, as shown in Figure 5.1. The minimum downcomer gap (24")

available in the pool is assumed to be the total gap available around the idealized cylindrical rack; thus, the maximum resistance to downward flow is incorporated into the analysis, d.

No downcomer flow is assumed to exist between the rack modules.

In this manner, a conservative idealized model for the rack assemblage is devised. The water flow is axisym-metric about the vertical axis of the circular rack assemblage, and thus, the flow is two dimensional (axi-symmetric three dimensional). The governing equation to characterize the flow field in the pool can now be written. It is shown in ref. (4) that'the resulting integral equation can be solved for the lower plenum velocity field (in the radial direction) and axial velocity (in-cell velocity field), by using the well known method of collocation. It should be added here that the hydrodynamic loss coefficients which enter 1

- - = a -- A + ,- - g , or - ---

into the formulation of the integral equation are also taken from well recognized sources in the literature; and wherever discrepancies in reported values exist, the con-servative values are consistently used.

After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. The knowledge of the overall flow field enables pinpointing the storage location with the minimum axial flow (i.e: maximum water outlet temperature).

This is called the most ' choked" location. It is recognized that these storage locations, where main supports are located, have some additional hydraulic resistance not en-countered in other cells. In order to find an upper bound on the temperature in such a cell it is assumed that it is located at the most " choked" location. Knowing the global plenum velocity field, the revised axial flow through thic choked cell can be calculated by solving the Bernoulli's equation for the flow circuit through this cell. Thus, an absolute upper bound on the wcter exit temperature and maximum fuel cladding temperature is obtained. It is believed that in view of the preceding assumption, the temperatures calcula cd in this manner grossly exaggerate the temperature rise that will actually be obtained in the pool.

5.3 Results Figure 5.2 shows a plot of the maximum water exit temperature vs. plenum h.eight. Corresponding to the plenum hcight of 7.5", the exit temperature is 1590 F; and the corresponding fuel assembly cladding temperature is less than 170 F.

These numbers clearly demonstrate the wide margin of safety in this aspect of the rack performance.

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4 i REFERENCES TO SECTION 5

1. Design Specification of'the Detroit Edison Co. No. 3071-178,

' for: Enrico Fermi Atomic . Power Plant, Rev. A (1979).

i i

2. . Chapter 9, FSAR, Section 9.1.3.

3. U.S. Regulatory Commission, Standard Review Plan, Branch Technical t

Position. ASB 9-2 Rev. 1, Nov. 1975.

1 i 4. "A Method for. Ilydro-Thermal Analysis of liigh Density Fuel Racks",

Oat Standard Document #20, Rev. 0 (1980).

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F I G. 5. L R ACK SPACE' E N V E L O PI'N G C Y LI N DE R-

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6. SEISMIC ANALYSIS 6.1 Analysis Outline

• The spent fuel storage racks are seismic category I equip-ment. Thus, in accordance with ref. (1) , they are required to remain functional during and after an SSE (Safe Shutdown Earthquake). As noted previously, these racks are neither j anchored to the pool floor, nor are they attached to the side walls. The individual rack modules are not intercon-nected. Furthermore, a particular rack may be completely loaded with fuel assemblies (which corresponds to greatest

]

rack inertia), or it may be partially loaded so as to pro-duce maximum geometric eccentricity in the structure. The i coefficient of friction a between the supports and pool floor is another indeterminate factor. According to

! Rabinowicz (2),'the results of 199 tests performed show a mean value of p to be .503 with a standard deviation of 0.125.

The upper and lower bounds (u i 2a) are thus 0.753 and .253, respectively. Two separate analysis are performed for this rack assembly with values of equal to 0.2 (lower limit),

and 0.8, respectively. In summary, the following four separate analyses are performed:

1. Fully loaded rack (all storage locations occupied);

p = 0.8.

2. Fully loaded rack, p = 0.2.

3. Half loaded rack to produce maximum geometric asymmetry, u = 0.8.

4. Half loaded rack to produce maximum geometric asymmetry, u = 0.2.

Cases 1 and 3 are analyzed to determine maximum rack stresses.

Cases 2 and 4 will yield an upper bound on rack displacement.

The method of analysis employed is the well known Time History method. The ground acceleration coincidently in three directions is specified by Stone & Webster Engineering Corp-oration. The object of the seismic analysis is to determine l

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! 4 the structural response (stresses, deformation, rigid body

motion,.etc.) due to simultaneous application of the three l orthogonal excitations. Thus, recourse to approximate j ' statistical summation techniques such as SRSS

method (ref.-3) is avoided and the dependability of computed 4

j results is ensured.

l The seismic analysis is performed in four steps; namely f

(i) Development of non-linear dynamic model consisting of beam, gaps, spring, damper and inertia coupling l elements. ,

i Derivation and computation of element stiffnesses using 4 (ii) a sophisticated elastostatic model.

j (iii) Layout'the equations of motion, decouple these equations

- and solve them using the " component element time in-

} gegration" procedure (ref. 4). Determine nodal forces.

(iv) Compute the detailed stress field in rack structure using the detailed elastostatic model from the nodal forces calculated in Step III above. Determine if the stress and displacement limits (given in Section 6.5) are satisfied.

A brief descriptio" of the dynamic model now follows.

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6.2 Fuel Rack - Fuel Assembly Model 6.2.1 Assumptions

a. The fuel rack metal structure is represented by five lumped masses connected by appropriate elastic springs. (Refer to Figure 6.1)

b. The fuel assemblies are represented by five lumped masses located, relative to the rack, in a manner which simulates either full or partially filled conditions.

c. The fuel rack base is considered as a rigid body supported at four points,

d. The rack base supports may slide or lift off the pool floor.

l e. The pool floor is assumed to have a known ground d

acceleration in three orthogonal directions,

f. Fluid coupling between rack and assemblies, and between rack and adjacent racks is simulated by introducing appropriate inertial coupling into the

, system kinetic energy.

g. Potential impacts between rack and assemblies is accounted for by appropriate spring gap connectors between masses involved.

h. Fluid damping between rack and assemblies, and between rack and adjacent rack is simulated by inclusion of appropriate equivalent linear damping.

~

i. The supports are modeled as rigid beams for dynamic analysis. - The bottom of the support legs is attached to a frictional spring as described in Section 6.2.2. The elastic properties of the Support beams are derived and used in the final computations to determine support leg stresses.

j. The effect of sloshing is shown to be negligible and is hence neglicted. It is to be noted that the top of the rack is over 20' below the free l water surface.

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6.2.2 Model Description The absolute degrees of freedom associated with each of the mass locations i, i* is as follows (Figure 6.1) .

LOCATION DISPLACEMENT ROTATION (NODE) ux uf uz ex g og y

1 py 94 95 96 p2 P3 1 Point is assumed fixed to base at XB,YB,Z=0 E7 P9 911 9 12

'l

• 2 p 8 P10 1 P13 Pg i 917 9 18 3 p4y pl6 4 pyg p12 923 9 24 4 p20 P22 5 p 25 P27 929 930 931 5 p62 P28 Thus, there are 31 degrees of freedom in the system.

Note that elastic motion of the rack in extension relative to the rack base has been neglected in the

model. Because of the stiff rack construction, this motion will not be excited by the seismic loading.

Torsional motion of the rack relative to its base is governed by q31' A schematic description of the rack supports is given in Figure 6.2. The members joining nodes 1 to 2,

)

2 to 3, etc. are beam elements with deflection due j

. to bending and shear capability (ref. 4, pp 156-161).

The elements of the stiffness matrix of these beam elements are readily computed if 'he effective flexure modulus, torsion modulus, etc. for the rack structure

- are known. These coefficients follow from the elasto- I static model as described later.

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• The node points i (i = 1, 2 ... 5) denote the cumulative mass for all the fuel assemblics distri-buted at 5 elevations. Referring to G.E. specifi-cation (ref. 5), the bending and torsional stiffnesses of the fuel assembly (channeled or unchanneled) are

! several orders of magnitude smaller than the rack beam elements. Hence, it is reasonable

+

to neglect the spring elements joining the lumped masses. The nodes i are located at X = XBe Y=Yq in the global coordinate system shown in Figure 6.1. The coordinates (XB, YB) are determined by the center-of-mass of the set of fuel assemblies.

For a completely loaded rack XB=YB = 0.

6.2.3 Fluid Coupling An effect of some significance requiring careful modeling is the so called " fluid coupling effect".

If one body of mass mi vibrates adjacent to another body (mass m), 2 and both bodies are submerged in a frictionless fluid medium, then the Newton's equation of motion for the two bodies have the form (ml + M11) XI - M12 X2 = applied forces on mass mi (o.1)

-M21 X+ 1 (m2 + M22) X2 = applied forces on mass M2 M11, M12, M21 and M22 are fluid coupliny coefficients which depend on the shapes of the two bodies, their relative disposition; etc. Fritz (5) gives data for Mij for various body shaps and arrangements. It is to be noted that the form of Eq. (6.1) indicstes that effect of the fluid is to add a certain amount of mass to the body (Mll to body 1), and an external force which is proportional to the acceleration of the adjacent body (mass m2). Thus, the motion of one body affects the force field on another, in the manner of

.the celestial gravitational pull. This inertial coupling is called fluid coupling. It has an important effect in rack dynamics. The lateral motion of the fuel i assembly inside the storage location will encounter e

- -s. -

---

,g n - ,v- ---ev

this effect. So will the motion of a rack adjacent to another rack. These effects are included in the equations of motion as described in detail in reference (6). The fluid coupling is between nodes i and i (i = 2, 3 ... 5) in Figure 6.1. Furthermore, nodal masses i are coupled to the reference frame through inertial coupling coefficients.

Finally, virtual mass is included in vertical direction vibration equations of the rack; and virtual inertia is added to the governing equations corresponding to rotational degrees of freedom, such as q4, q5' 96 ' 911, etc.

6.2.4 Damping In reality, damping to the rack motion arises from material hysteresis (material damping), relative inter-component motion in Structurcs (structural damping),

and fluid drag effects (fluid damping).

Only fluid damping is included in the analysis. The fluid damping acts on the i nodal masses, as well as on i nodal masses. The equivalent values of linear dampers for various types of motions are derived in ref. (7).

6.2.5 Impact The fuel assembly nodes i will impact the corresponding structural mass node i. To simulate this impact, 4 impact springs around each fuel assembly node are pro-vided (Figure 6.3). The fluid dampers are also pro-vided in parallel to the springs. The spring constant of the springs is equal to the local stiffness of the vertical panel computed by evaluating the deflection of a 6" diameter circular plate (.075") uniformly loaded,.and built-in around the edge. The spring con-stant calculated in this manner should provide an upper bound on the local stiffness of the structure.

A brief description of the elastostatic model now follows.

6.3 Stress Analysis f

6.3.1 Stiffness Characteristics The fuel rack is a multi-cell folded-plate structure which has what is colloquially called an " egg-crate" configuration. This type of construction is very similar to the so-called " stressed-skin" construction of ribs spars and cover plates which are widely used

in aircraft construction. Techniques developed in the field of air raft structural analysis are utilized 1

herein to find the stresses and deformations in such structures. The<;e methods have been thoroughly tested

and their reliability has been documented in a number

, of well-known publications (e.g. ref. 8 thru 12 ) .

Figure 6.4 shows two cross-sections of the fuel rack

which is modeled as a rectangular network of plates interconnected along nodal lines shown as points in i

Fig. 1-A. An arbitrary load with components Fx i Fy , ,

j Fz i acts at an arbitrary elevation on one of the nodal lines. We find the displacements and stresses due to

]

i such a typical load according to the stressed skin i

j model as follows:

)

1 The torsional deformations are solved for by using the classical theory of torsion for multi-celled thin

] walled cross-sections (ref. 13).

l The bending deformation is found by using the theory of shear-flow (ref. 12) wherein all axial stresses are carried by the erfective flanges (or stringers) g formed by the intersections of the plates and all trans-verse shears are carried by the plates modeled as shear panels.

From a knowledge of the shear flows, the bending and torsional deformations, it is possible to provide a se.t of influence functions or the following section pro qrties for the fuel rack as a whole:

6 6

- - - , . , . . , . . ,,,n,, - ,, , - , - , ~ . ,

-, - - - - - - -. , - - . - , ~ ,, 4- . . _ . .

1 l

(EI)eq

= Bending rigidity (in two places)

(GJ)eq Torsional rigidity (AE)eq

= Extensional rigidity ks = Shear deformation coefficient Such properties are used for the dynamic analysis of seismic loads. The detailed equations are documented ,

in ref. (7).

6.3.2 Combined Stresses and Corner Displacements

! The cross-sectional properties und the Timoshenko shear correction factor calculated in the previous section are fed into a dynamic analysis of the system shown in Figure 6.5 with a specified ground motion simulating earthquake loading. From the dynamic analysis, the stress resultants (Fx, Fy, Fz , Mx, My, M) z acting as shown in Figure 6.6 are computed for a l large number of times t = at, 2 at ... etc, at a selected number of cross sections. The displacements l (Ux , Uy, Uz) Lt selected nodal points on the z axis are also provided by the dynamic analysis as well as rotations (ex , ey , Oz) of the cross-sections at the nodes.

Figure 6.7 shows a typical sub-division of the structure into elements, nodes and sections. The stresses are i

calculated at all sections and the displacements at all four corners of the rack are calculated at these elevations.

4 i

O 9

9 w~ -- -

- - - , , , , , , - - - - > - a,.y - + - -

I I

I i

Since o varies linearly over the cross-section and achieves its extreme values at one of the four corners of the rack. The shear stresses due tc corsional loads (, M z) achieve their extreme values near the middle of each side. The shear stresses due to lateral forces (Fx, Fy) will achieve their extreme j values at the venter of the cross section or at the middle of each side (see Section 4). Thus, candidates 3

for the mout critical point on any section will be the points labelled 1, 2 .. 9 in Figure 6.8. The expression for the combined stress and kinematic dis-placement for each of these points is written out.

Similarly the stresses in the support legs are evaluated.

An Oat proprietary computer program "EGELAST" computes the stresses at the candidate points in each level.

It sorts out the most stressed location in space as well as time. The highest stress, and maximum kinematic

displacement are thus readily found.

Exhaustive details of the mathematical formulations, and documentation of the computer program are avail-able from Joseph Oat Corporation.

4 I Also known as "octahedrol shear stress", or " dis-I tortion energy", or " Von Mises", theory.

4 4

4 ..

. --, . _ , .., , _ . . . . . . - . . ~ . . , , . , _m. , - ,

i i

1

^

6.4 Time Integration of the Equations of Motion Having assembled the structural model, the dynamic equations of motion corresponding to each degree of freedom can be t

written by using Newton's second law of motion; or using Lagrange's equation. For example, the motion of node 2 I in x-direction (governed by the generalized coordinate p7) is written as follows:

~

The inertial mass is "21 + ^211 + U211 where m 21 is the mass of node 2 for x-directional motion.

A 211 is the fluid coupling mass due to interaction with node 2 .

211 is the fluid coupling mass due to interaction of node B

4 2 with the reference frame (interaction between adjacent

! racks).

Hence Newton's law gives I"21 + A211 + B211 IE7+A212 98+B212 U "02 where 0 2 represents all the beam spring and damper forces on node 2. and A 212 is the cross term fluid coupling effect of node 2 ; and B 212 is the cross *'

term fluid coupling effect of the adjacent racks. u represents the ground acceleration.

Let 97

• p7 -u 1.e. q7 is the relative displacement of node 2 in x-direction

] with respect to the ground. Substituting in the above equation, and rearranging, we have (521 + A211 + B211 I97 + ^212 98=02 -

I"21 + ^211 + B211

• ^212 + B212' Similar equation for each one of the 31 degrees of freedom can be written out. The system of equations can be repre-sented in matrix notation as:

1 4

M MMW* -

. . . , , , , , . . . ~ . . -

[M1 (q) =

[0] + (G) where the vector [0] is a function of nodal displacement and velocities, and (G). depends on the coupling inertias and ,

the ground acceleration.-

~

Pre-multiplying above equation by [M] renders the resulting equations uncoupled in mass.

! We have.'

1

~ ~

f b[) = [Mi [Q] + -[Ml (G)

This equation set is mass uncoupled, displacement coupled; and is idealy-suited for numerical solution using the central difference scheme. The computer program developed by G.E.

and described'in ref. (4) performs this task in an efficient manner. This computer program, named "FVAIS3" in Oat's i computer program library is documented in ref. (4), and also internally at Oat. The documentation is available upon request.

Ilaving determined the internal forces as a function of time, the computer program "EGELAST" computes the detailed stress

' and displacement field for the rack structure as described in the preceeding section.

1 I

1 I

J e

, +<g--,-.-w-- ,--w-

• -t-4 *+-- ww p-e- + * - - e- -*-----~y + - *

• v * - -

, -- . - - - -. ... .~ . - . .-. - . __.

6.5 Structural Acceptance Criteria There are two sets of criteria to be satisfied by the rack modules:

a. Kinematic criteria: This criterion seeks to ensure that adjacent racks will not impact during SSE (condition E' in ref. 14), assuming the lower bound value of the pool surface friction coef ficient. It is further re-

, quired that the factors of safety against tilting

, specified in ref. (15) are met (1.5 for OBE, 1.1 for SSE).

b. Stress limits: The stress limits of the ASME code,

< Section III, sub-section NF (latest issue) were chosen to be met, since this code provides the most consistent set of limits for various stress types, and various loading conditions. The following loading cases (out of the set specifled in ref. (14) are meaningful.

SRP Designation ASME Designation i (i) D&E Level B (upset condition)

I j (ii) D& E' Level D (faulted condition) l where i D: Stresses due to dead weight

, E: 0.B.E. (time history loading)

E': S.S.E.-

The supplements to the NF stress limits given in ref. (16) are also incorporated for the specific stress categories.

The stresses in the panels of the cruciform are required l

to be within the corresponding limit for plate-shelt type ,

1

structures. The support plates (vertical & horizontal)  ;

i must satisfy the same stress limits. The welds between l the base plate of the cruciform need not be continuous, but they must meet NF stress limits on welds.

Similar  !

requirements. hold for the base support welds.

i 2 I t

e 2

m *

,_,_.r_ . . . , _ , , .,.

6.6 Results Computed results show wide margins of safety. For example, the maximum stress intensity (membrane & bending) at the root of the cruciform is only 5400 psi (corresponding to fully loaded, u = 0.8 condition) vs. NF limit of 3 Sm.

This occurs at point 3 in Figure 6.8.

(SM = 20,000 psi per table I, 1.2 of section III appendix IX)

The maximum stress in the worst stressed leg is 24000 psi.

The p = 0.2 fully loaded condition gives a maximum rack stress of 3300 psi. The x and y support displacements are 0.15" and 0.11" respectively.

Detailed results and margins of safety are documented in ref. (7).

l l

l summ-

References to Section 6

1. Regulatory Guide 1.29, Seismic Design Classification, Rev. 2, Feb. 1976.

2. " Friction Coefficients of Water Lubricated Stainless Steels for

^

a Spent Fuel Rack Facility", by Prof. Ernest Rabinowicz, M.I.T., a report for Boston Edison Company.

3. Regulatory Guide 1.92, combining Modal nasponses and Spatial Components in Seismic Response Analysis, Rev. 1, Feb. 1976.

Regulatory Guide 1.61, camping Values for Seismic Design of Nuclear Power Plants, Oct. 1973.

4. "The Component Element Method in Dynamics with Application to

. Earthquake and Vehicle Engineering" by S. Levy and J.P.D. Wilkinson, McGraw Hill (1976).

5. General Electric specification 22A5866, Rev. 1, Appendix II, Fuel Assembly Structural-Characteristics.

6. R.J. Fritz, "The Effect of Liquids on the Dynamic Motions of 4

Immersed Solids", Journal of Engineering for Industry, Trans.

J of the ASME, Feb. 1972, pp. 167-173.

7. " Seismic Analysis of Fermi II High Density Racks", Oat report No. TM-588, 1980.

8. J.T. Oden, Mechanics of Elastic Structures' McGraw-Hill, N.Y., 1967.

! R.M. Rivello, Theory and Analysis of Flight Structures, McGraw-Hill 9.

N.Y., 1969.

a

10. M.F. Rubinstein, Matrix Computer Analysis of Structures, Prentice-

! Hall, Eaglewood Cliffs, N.J., 1966.

11. J.S. Przemienicki,. Theory of Matrix Structural Analysis, McGraw-Hill, N.Y., 1966.

i j 12. .P. Kuhn, Stresses in Aircraft and Shell Structures, McGraw-Hill, N.Y., 1956.

13. S.P. Timoshenko and J.N. Goodier, Theory of Elasticity, McGraw-Hill, N.Y., 1970, Chap. 10.

1 4

6

- -,.c- < -y-4. . - - -~w-r er--r-

47_

14. U.S. Nuclear Regulatory Commission, Standard Review Plan, NUREG-75/087, Section 3.8.4.

15. SRP NUREG-75/087, Section 3.8.5.

16. NRC Regulatory Guide 1.124.

l l

l l

l l

Z Ja 3 5 ---

, 5%

/

/ CO U PLIN G ELEMENTS '

• l l

TYPICAL FU EL A S S EM B LY 3

GROUP M AS S H TYPIC AL FUEL R ACK MAS S 2 FU EL R AC K B A S E 2

= AY p  :

~

/ \ /

n Ax _L

"' l I p- ye 7- wY i

' A l--f!

• B N3 2

, t i

j a I '

s h I

sii -

L os r

FUE L R AC K SUPPORT

~

l X -

XB, YB - LOCATION OF C E N TR OI D O F FU EL ROD G ROUR M ASSES - REL ATIVE TO ,

CEN TER O F FU EL R A C K )

~

, n' = UNIT VECTORS - I FI G. 6.1 DYN A M I C MODEL ,

CELLS MODULE B ASE PLATE ,..~~~~I ~ ~q' i lj F  :: 9 s

$\\\\ \ t.-

- bI --

3 INAT5 a '? l .

7' ,,

(..__

l n

I +

da 3/4 2

t f

2

[A [

,/g 1

u il}a xggxgggggy 09 I

15 /4 S Q,

a INN NN NNI , INNNNN%

// N

/

/. /

/ .

/

/ \ '

/

I

i. d I

/_

l3 /4 r 7 /

S Q. /

f/ e

/,

/ -3/4 TYP.

n //l ,r 9

_w NxNxw imxxwr.-

t FIGc6 2 SUPPORT e

9

- , , , ~ . - - . - . - .,,

L IMPACT h SPRIN G S LT /

[.H MASS Mb

. W {* -C n ?

L F L c'I n DAM PERS RIGID FRAME

= X F I G. 6.3 I M PAC T SPRI N GS AND F LUI D DAMPERS

i j, Fy Yd B IL  ;  ; 8

= Fx

'X (o) TOP VIEW Z, ,

i F

J g/z v =- p -.

(b) AXIAL CROSS y u m.,,,,, S ECTlON ( B-B )

FIG.6.4 (a) HORIZONTAL CROSS

. SECTION OF RACK (b) VERTICAL CROSS SECTION OF RACK t

4

~

CELL z(W) '

WALLS ffC t W/#4#/

/ ,:.l/l b 0 (k "g

,n,if n (p( W o b= N y C

\

C a:NxCy

~ -

C A' , y(V)

A RIGlD PLATE B / .

~

= X (u) l re-a),r=7 \

P=

A = dj,?e

//

, / i _ f SUPPORTS FIG.6.5

~

Ji Mz a jMy gzF A ,

A ./ '/ B

,=-M x 1

C A

B

~

e e F G. 6.6 ..

. . . . . . . . . = .

-s3-(

aZ NODEI

/g E L.I l S E C.I --_ --._

NOD E 2 ~ / ,, - E L. 2

. S E C. 2 --- -

N O D E 3 -- E L.3 S E C. 3 - --

N O D E 4 -[ - a

- - -- E L . 4 [- y S EC.4 --.- -

'y , EL.5 rz /

X

• S E C.5 ," /

, /---, n

+-4 . NODE 5 /

,c. . g' ~ ~

/k S E C,6 E L,7 - ~ /ROOT OF R ACK , -EL.8 S E C, 8 ~

.', SEC,9 N O. O F E L E M E N T S = 8 -

N O. OF S EC TI ON S' =9 .

N O. 'O F N O D E S =5

~

F , G. 6. 7 .

\

l I

M

. S .

'b 1 e

e e l iv ,t

- ~- - - ..

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~.

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a

. 6 5

b 7 X 9

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4 {

3 8 {

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1 unnen l G. 6 8 4

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s

7. MISCELLANEOUS ANALYSES In addition to.the ground motion analyses for the following short duration loads are to be perform'ed:

a. Dropped Fuel Accident I A fuel assembly (weight = 600 lbs.) dropping 16'9" through a storage location and impacting the base. Local failure of i the base plate is acceptable; however, a substantial impact I with the pool liner is not allowed. The sub-criticality of the adjacent fuel assemblies is not to be violated.

i

b. Dropped-Fuel Accident II Fuel assembly dropped on top of the rack with an impact energy of 2000 ft-lb. The impact energy is assumed to correspond to a mass of 600 lb. dropped through a height 40". The impact is i

assumed to occur on the top ridge of the rack. The permanent deformation is required to be limited to a region 4 times the cross-sectional area of the fuel assembly.

I

c. Jammed Fuel Handling Equipment A 1200 lb. uplift force applied at the top of the rack in the

" weakest" storage location. The force is assumed to be applied on one wall of the storage cell boundary as an upward shear

! force. The damage, if any, is required to be limited to the affected storage locations.

i 1 d. Horizontal Force n horizontal force of 1000 lb. applied at the most vulnerable j location on the top of the rack. The load is assumed to act

! over the width of one storage cell. The permanent deformation j is required to be limited to the affected cells.

} Because of the immense structural rigidity of the cruciform ,

j construction, the aforementioned conditions are satisfied with i large margins c# safety.

4 O

f l

3

8. NEUTRON ABSORBER MATERIAL The material utilized for neutron attenuation in the racks is Boraflex; a proprietary product of Bisco, a Division of Brand Industrial Services. This material is available in sheet form which facilitates easy handling and a close control of lateral dimensions during fabrication. This material has found wide-spread acceptance due to its durability, and a remarkable reten-tion of physical and mechanical properties when subject to high or low flux irradiation under typical fuel pool environments.

A brief resume' of the established information on this material is given in the following:

8.1 Chemical Composition The elemental composition of the Boraflex proposed can be divided into two catagories, the polymeric matrix system and the boron carbide power. The elemental composition of each to the nearest 0.5 wt. % is listed below:

TABLE I Elemental Composition of Boraflex Components by Weight %

ELEMENT POLYMER B4C Silicon 41%

Oxygen 37%

Hydrogen 4.5% -

Carbon 17.5% 23.5%

Boron - 76%

Iron; soluble borons - 0.5%

40 2 The minimum B loading is 0.014 grams /cm at a nominal thick-ness of .070". The criteria suggests a formulation based on 42 wt. % boron carbide to assure that the specified B con-tent is exceeded at the minimum acceptable manufacturing tolerance thickness (I 10% typical, ! .010" maximum). The elemental content of Boraflex based on this formulation would be as follows:

. o ,

t 57-

' TABLE II .

Elemental Composition of Boraflex Containing - .

?

42 wt. % B3C (by wt. %)

Silicone 24.0% ,

't Oxygen 21.5%

Hydrogen' 2.5%

. Carbon 20.0%

Boron 32.0%

Iron, soluble boron trace Note that the isotopic B 0 content expressed as wt. % of total boron is typically 18.0 1 .4.

2 8.2 Physical' Properties Boraflex has been extensively tested for physical and mechanical characteristics when subjected to high and low i

' rate irradiation while contained in air, deionized water Careful laboratory data on or borated water environments.

neutron attenuation, elemental boron leaching, residual activity, gas generation, etc. were also taken and docu-mented. Bisco report 748-10-1 contains detailed description of the procedures and recorded results. It is shown that 8

the exposure of boraflex in air to 2.81 x 10 rads gamma from a spent fuel source results in no significant physical changes nor in the generation of any gas. Irradiation to the level'l.03 x 10 ll rads gamma with a substantial con-current neutron flux in air, deionized water, and borated water environments causes some increase in hardness and tensile strength of boraflex. During that irradiation a certain amount of gas is generated.but beyond the level of rads gamma drops off considerably. The rate of gas 1 x 10 generation is found to be greater when 4B C is irradiated in deionized or borated water in absence of boraflex, thus confirming the function of boraflex polymer as our escapsulant which mitigates the interaction between boron carbide and

. the environment. . Vent holes are provided on top of each j storage cell compartment to eliminate gas entrapment.

rg - - , - - -

y7, 4 .-- ,--nv v.-e,e.. .=e v ..- ---n -- ,--r- -,ge - , - -yw--

Measurements of the specimen width, _ thickness, weight, specific gravity at pre-and post. irradiation stages indicated minuscule variation in these quantities.

Experiments also show that neither irradiation, environment or boraflex composition has any discernible effect on the i

neutron transmission of boraflex. Tests also prove that boraflex does not possess leachable halogens that may-be extracted into the pool environment in the presence cf radiation. Similar conclusions are reached regarding leaching

• of elemental boron out of boraflex. The results attested to'the efficient encapsulement function of the boraflex matrix in-preventing dissolution of normally soluble boron species.

A critical examination of the voluminous body of evidence on the functional characteristics of boraflex has led Joseph Oat Corporation to recommend its use in the Fermi II racks.

(

j 4

1 i

J t

I i 1

(

i -

s i

_59_

9. IN-SERVICE SURVEILLANCE PROGRAM FOR BORAFLEX NEUTRON ABSORBING MATERIAL 9.1 Program Design Intent All materials used within a storage system for spent nuclear fuel are qualified to a level of performance predicated upon calculated worst case environmental conditions and are based on accelerated testing of the materials to levels of service life corresponding to that calculated environment. Because such environmental compatibility testing is accelerated, it is prudent that each of the system components be monitored to some extent throughout the service life to assure that the actual in-service performance remains within acceptable parameters as defined by the accelerated testing. For many of the materials, monitoring throughout the service life is relatively easy, however, the poison material is encased in a stainless steel jacket precluding both visual and conven-ient physical examination during the in-service condition.

A poison surveillance program will be conducted which allows access to representative poison samples without disrupting the integrity of the storage system. Such a program must include not only the capability to evaluate the material in a normal use mode, but to forecast changes that might occur within the storage system at a time significantly prior to the normal use mode occurence of such changes.

9.2 Description of Specimens The poison used in the surveillance program must be repre-sentative of the material used within the storage system.

It must be of the same composition, produced by the same method, and certified to the same criteria as the produc-tion lot poison. The sample coupon must be of similar thickness as the poison used within the storage system and not less than 2" x 4" on a side. Each poison specimen must be encased in a stainless steel jacket of an identical alloy to that used in the storage system, formed so as to encase the poison material and fix it in a position and with tolerances similar to that designed into the storage system.

The jucket would be closed by tack welding in such a manner as to retain its form throughout the use period yet allow rapid and easy opening without contributing mechanical damage to the poison specimen contained within.

A series of not less than 16 of the jacketed poison speci-mens shall be suspended from a rigid strap so designed as to be hung into the pool from a surface support. The spec-imen location must be adjacent to a designated storage cell with design ability to allow for removal of the strap, pro-viding access to a particular specimen. The detail of arrangement is shown in ref. (2).

9.3 Long Term Surveillance 4

At the time of the first off-load of spent fuel, one specimen strap is located directly adjacent to a non-poisoned wall of a cell designated as a long term testing location and con-taining a freshly spent fuel element. Both the spent fuel element and the specimen strap must be controlled to remain in fixed location throughout the entire storage life of the pool. At each incremental five year period subsequent to the initial installation, the strap should be raised from the pool and one specimen taken Icr evaluation. Based on a minimum of 8 specimens, a 40 year oparational period may be evaluated by the generation of specific evidence of poison performance under normal operational conditions. The long term testing location will be shown in ref. (1).

9.4 Accelerated Surveillance At the time of the first off-load of spent fuel, one speci-men strap is suspended in the pool adjacent to a non-poisoned wall of a cell designated as an accelerated testing location and containing a freshly spent fuel element. At the time of the second off-loading the fuel element in the designated accelerated testing cell is removed from that cell location I

and relocated to a new cell or permanent storage. The specimen strap is withdrawn from the pool and a jacketed specimen removed from the strap for evaluation. The speci-men strap is replaced in the pool adjacent to the same 1 l

1 o e l

I I

designated accelerated testing cell and freshly spent fuel element from the second off-loading is placed into the designated cell. This cycle is repeated 8 times with each cycle being comprised of the removal of a jacketed speci-men for evaluation and the placement within the designated accelerated testing cell of a freshly spent fuel element from the most recent off-loading. By evaluation of the specimens an accelerated monitor of environmental effects on the poison will be obtained, simulating within an 8 year period the effects of cycling freshly spent fuel into the same fuel cell once every 5 years for a period of 40 years.

The 5 year use cycle typically represents the most severe fuel relocation schedule under administrative or procedural controls. The accelerated testing location is shown in ref. 9.1.

4 9.5 Specimen Evaluation After emoval of the jacketed poison specimen from the fuel pool at the designated time, a careful evaluation of that specimen will be made to determine it's actual condition as well as it's apparent durability for continued function.

Separation of the poison from the stainless steel specimen jacket must be performed carefully to avoid mechanically damaging the poison specimen. Immediately upon removal, the specimen and jacket section should be visually examined for any effects of environmental exposure. Specific attention should be directed to the examination of the stainless steel jacket for evidence of physical degradation. Functional evaluation of the poison material is accomplished by the following measurements:

1. A neutron radiograph of the poison specimen will allow for a determination of the maintenance of uniformity of the boron distribution.

2. Neutron attenuation measurements of the specimen made in a fashion consistent with that described in the Poison Material Qualifying Test Data will, by comparing with the attenuation of preirradiated poison as listed in that document, allow evaluation of the continuing nuclear

%g .

'~

effectiveness'of the poison. Consideration must be given in one analysis of the attenuation measurements for the level of accuracy of such measureuents as in-dicated by the degree of repeatability normally ob-served by the testing agency.

3. A measurement of the hardness of the poison material will establish the continuance of physical and structural durability. The criteria of hardness acceptability in-cludes all values not less than the hardness listed in the Qualifying Test Documents for the poison material under the category of pre' irradiated hardness. The actual hardness measurement should be-made after the specimen has been withdrawn from the pool and allowed to air dry for not less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to allow for a meaningful direct correlation with the preirradiated sample.

, 9.6 References

1. Oat DWG D-6939 Rev. 0

2. Oat DWG D-6943 Rev. 0 4

a o

l

10. DESIGN CONTROL AND FABRICATION INTERFACE A. Introduction

.- t In the following, an abstract of the design control from Oat's Q.A. System is presented in flow chart form. This program has been accepted by the ASME for engineered fab-rication of ASME Section III, Class 1, 2, 3 and MC com-ponents. The program has been found to be acceptable to NRC audit teams, as well as to special projects such as CRBRP and U.S. Department of Defenso.

B. Personnel The personnel categories involved in the operations are:

a. General Manager (G.M.)

b. Chief Engineer (C.E.)

c. Project Engineer (P .E. )

d. Professional Engineer

e. Designated Analysts (D.A.)

f. Designated Draftsman

g. Contract Administrator (C . A. )

The flow of work is shown in the following flow charts.

Flow Chart #1 shows the job progress sequence from initi-f ation. Flow Chart #2 gives the operation sequence following customer feedback to the initial document; and customer generated documents.

All documents to be treated in course of a job are divided into five' types as noted in the footnote of Flow Chart #2.

This operational flow chart gives the minimum required number of steps in the processing of a contract. Additional personnel may be called upon for expert help by the Project Engineer wherever deemed necessary. For example, the practical advice of the Shop Superintendent in determining feasibility or economy of a design, advice of the Quality Control Manager regarding NDT and material testing require-ments are frequent types of help sought by the Project Engineer. These are necessary steps for high quality I

,- , ~.

design, although not essential for meeting quality assurance requirements.

The procedure itself is seif-explanatory as laid out in the two Flow Charts.

C. Flow Charts ,

us-a DESIGN CONTROL .

FLOW CHART CUSTOMER FORWARDS CONTRACT DCCUMENTS TO OAT v

CONTRACT ADMINISTRATOR (C.A.)

RECEIVES PURCHASE ORDER AND CUSTOMER SPECIFICATIONS I

IC.A.

EXAMINES THE CONTRACTUAL TERMS WITH THE HELP OF k _,_ _

SALES PERSONNEL AND 4 GENERAL MANAGER l CUSTOMER l

'r REVISED CONTRACT E

DOCypj,NT e l y

I C.A. lC.A.

. ACCEPT WITH EXCEPTIONS l CUSTOMER ACCEPT AND ACKNOWLEDGE A PURCHASE '? DER g Y

C.A.

ASSIGN JOB NUMBER v

lC.A.,  ;

PREPARE DATA FOR PREPARE JOB FILE. FORWARD DOCUMENT SUBMITTAL FORM SPECIFICATIONS AND ESTIMATE AND FORWARD TO WORK MATERIAL TO PROJECT ENGINEER TRANSMITTAL CLERK (T.C.)

CALL PROJECT-REVIEW MEETING WITH CHIEF ENGINEER (C.E.) AND PROJECT ENGINEER (P .E . )

<r l

kP.E. CUSTOMER 2 7

ANSWER / INCORPORATE PAGE REVIEW TECHNICAL DOCUMENT TECHNICAL DOCUMENT 3 FORWARD COMMENTS TO CUSTOMER COMMENTS 1

~~

}

lP.E.

PREPARE SOFTWARE SHEET FORWARD COPIES TO CHIEF ENGINEER (C.E.) AND C.A.

e l P.E.

PERFORM PRELIMINARY DESIGN l lP.E. ,

PREPARE DESIGN REPORTS DRAFTSMAN DESIGNATED DRAFTSMAN PREPARES CONCEPTUAL DRAWINGS l ENGINEER & REQUIRED DETi.ILED DRAWINGS, BM g I

I REVIEWED BY - _ _ . _ _ . . _ .

(DRAFTSMAN FORWARD CONCEPTUAL DESTD DESIGNATED ENGINEER DESIGNATED CHECKER DRAFTSMAN TO DESIGNATED ANALYST (DA)

CHECKS DRAWINGS AND BM '

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l CERTIFIES THE REPORT REVIEWS DRAWINGS AND BM ,

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REPORTS TRANSMITTAL CLERK (T.C.)

ISSUE DRAWINGS & BM TO Q.C.,

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SHQP, O.C., PURCHASING, P.E. REVIEWS THE REPORT HELD PRE-FABRICATION MEETING TC DISCUSS & COMMENT ON DWGS & BM q lP.E. PROFESSIONAL ENGINEEN A A D CERTIFY GIVES DRAWING BACK TO DESIGNATED DRAFTSMAN FOR CORRECTION THE REPORT l DESIGNATED ANALYST l DRAWINGS ANALYSTAPPROVED (NOT BM) BY DESIGNATED iP.E.

APPROVAL AND RELEASE BY PROJECT ENGINEER TO T.C.

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ISSUE TO CUSTOMER WITH DOCUMENT SUBMITTAL FORM COPIES TO C.A. AND P.E.

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' APPROVAL .E. ANSWERS l

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TRANSMIT TO CUSTOMER WITH DOCUMENT SUBMITTAL FORM IF REQUIRED DOCUMENT TYPES: i

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OAT PREPARED DRAWINGS 5l CONTRACTUAL (NON-TECIINICAL) DOCU-

[)) CUSTOMER PREPARED REPORTS l4{ OAT PREPARED REPORTS MENTS

11. QUALITY ASSURANCE PROGRAM Scope ,

This section provides a general description of the Quality Assurance Program that is implemented to assure that the quality objectives of the contract specification are met.

General The Quality Assurance Program to be used on this project is based upon the system described in Joseph Oat's Nuclear Quality Assurance Manual. This system is designed to provide a flexible, but- highly controlled, system for the design, manu-facture, and testing of customized components in accordance with various Codes, specifications, and regulatory requirements.

The philosophy behind Oat's Quality Assurance System is that it shall provide for all controls necessary to fulfill the contract requirements with sufficient simplicity to make it functional on a day to day basis. As this system is applied to most con-tracts which Joseph Oat obtains, implementation of it is almost second-nature to Oat's personnel. The system readily adapts to different designs and component configurations making possible the construction of many varied forms of equipment. The high-lights of this system, as addressed in the following paragraphs, provide an overview of the system and how it has been applied to the customer specifications and regulations.

System flighlights The design control section is organized to provide for careful review of all contract requirements to extract each individual design and quality criteria. These criteria are translated in-to design and quality control documents customized to the con-tract requirements and completely reviewed and approved by responsible personnel.

The system for control of purchased material entails generating detailed descriptions of each individual item of material along with specifications for any special requirements such as impact testing, corrosion testing, monitoring or witnessing of chemical

analysis, provision of overcheck specimens, special treatments or conditioning of material, source inspection, and provision of documentation of performance of any of the above.

Material receipt inspection includes a complete check of all material and its documentation. Upon acceptance, each item of material is individually listed on a control sheet issued once a week to assure that only accepted material goes into fabri-cation.

The fabrication control system provides that a shop traveler is prepared for each subassembly and assembly in each contract.

The traveler is generated specifically to provide step by step instructions for fabrication, inspection, testing, cleaning, packaging, etc. which will address all standard and special requirements of the contract specifications.

Due to the tendency of contract specifications to require special examination techniques or test procedures, all nondestructive examination procedures and test procedures are' custom written to apply to each given component within a contract.

The system provides for qualification and written certification of personnel performing quality related activities including nondestructive examination and fabrication inspection, welding, engineering, production supervision, and auditing.

Other requirements of a solid quality control system are fully covered as specified in the Quality Assurance Manual including document control, control of measuring and test equipment, control of nonconforming material and parts, corrective action auditing, and other areas as specified.

Summary Joseph Oat Corporation's Quality Assurance System provides the full measure of quality assurance required by the contract.

All special requirements of tne specifications will be cavered including source inspection of material and witnessing of-material testing by the Engineer, furnishing of material certi-i fications and test ceports within five days of shipment, and obtaining verification of qualification testing of poison l

o .

materials. We have a long history of providing excellent quality control over a wide range of equipment types such as the high density fuel racks.

J

- 12 . PRODUCTION CONTROL 12.1 Synopsis Production Control at Joseph Oat Corporation is based on the use of critical path diagrams (CPD). A critical path diagram is developed for each component manufactured at Joseph Oat Corporation. The critical path diagram con-sists of a detailed breakdown of the operation's required to fabricate each part, subassembly, and total assembly required to complete the finished product. The critical path diagram is arranged to show inter-relationship of all parts and sub-assemblics, including milestone dates'for the completion of each operation to assure that all parts and sub-assemblies are completed in time to support the overall fabrication schedule.

12.2 Procurement A bill of materials is generated for every component to be manufactured. The bill of material is reviewed against the CPD to determine the required delivery date for each item of material. This information is given to the Purchasing Department to be used-as the basis for purchase delivery requirements. The Purchasing Department has a full-time Expeditor to continuously review the scheduled delivery of all materials from suppliers. Problem items are reported to the Purchasing Agent who is responsible for assuring on-time delivery of all materials. Expediting visits to the supplier in question are performed by the Purchasing Agent or Expeditor whenever necessary. In addition, Production Control reviews the received materials on each component on a weekly basis. Any unreceived item of material which is within 2 weeks of its critical required date is reported to Purchasing and to the General Manager. The General Manager institutes the corrective action which is necessary to maintain the required delivery.

12.3 Shop Floor Planning Daily work assignments on the production floor are generated l

by the Plant Manager. All work assignments are planned out in writing a week in advance. Work assignments are based on completing the operations necessary to maintain the schedule required by the critical path diagrams. The work assignment sheet is checked each week by Production

-Control to assure that all required work is scheduled.

12.4 Operations Control & Coordination The critical path diagram for each component is monitored continuously by the Production Control Department. Once a week each component's status is determined and recorded on the CPD. The diagram is then reviewed to identify any operations which are not on or ahead or schedule. All such operations are reported to the General Manager,

Plant Manager, and top management. Production Control meets with the Plant Manager to determine the action necessary to bring the operation back on schedule. The work schedule for the following week is revised as necessary to assure performance of the work required to support the delivery schedule.

12.5 Reporting The complete status of each component in the fabrication schedule is reported to management by Production Control every week in the form of updated critical path diagrams.

This information is used by management for future work load planning, scheduling, and reporting status to the customer.

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RESULTS bx hx CASE RACK STRESS SUPP STRESS U Uy X

1 (Full) 4319 13350

.8 .393 .429 2 (FULL) 1578 2875 .177

.2 .123 3 (HALF) 3252 11520 .'63 I

4

.8 '.'79 4 (llALF) 1070 t

2603 .31

.2 .25 5 (HALF 3478 17600 .45

.8, .31 CASE 2)

. 6 (SAME AS 1, 6491 25850 .47 10% DAMPING). .27 i

}

q G

-- =- -

^

I STRUCTURAL DESIGN OF POOL AND BUILDING A. GRAVITY LOADS I e ORIGINAL DESIGN LOAD e ORIGINAL IMPOSED LOAD l e NEw GRAVITY LOAD e CONCLUSIONS

B. SEISMIC LOADS e BUILDING MODEL i

e ADDITIONAL MASS e EFFECTS OF ADDITIONAL MASS e VERTICAL ANALYSIS e CONCLUSIONS N

' e o

A 9

m- . . - . . _memene -

---

_ = to uw e a emwee

ORIGINAL DESIGN = 2,000 PSF

.i LOW DENSITY RACKS -

= K SUEMERGED WT 860

=

LOADED AREA 1015 FT2

=

ACTUAL UNIF0FN LOAD 850 PSF HIGH DENSITY RACKS

= K SUBMERGED WT 1663 LOADED AREA = 1005 FT2 ACTUAL UNIFORM LOAD = 1,655 PSF

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/ - 2nd floor - e.e va t ion 613' -6" 2

./ Dryvell i eleva tion 597 ' -0" Center Line _

-1st floor - eleva tion 583'-6" ressure ha b r eleva tion 55 7 '-9" Center Line N ase B slab - elevation 540'-0"

~.

.k ENRICO FERMI ATOMIC POWER PLANT UNIT 2 FINAL SAFETY ANALYSIS REPORT FIGURE 3.7-12 REACTOR / AUXILIARY BUILDING (NORTH-SOUTH SECTION THROUGH REACTOR CE.;TER LINE 400 KING WEST) j SARGENT & LUNDY REMRT NO. SL-2582 I

l

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CONTA I N'A EN T SHIELD

. . . . 18.

ORYWELL EELLOWS

, 27 9 l

X - DIRECTION 4 93bV 6 yf,, REFucLiNo eettowS ONLY CON T A I N '.* E N T

.17. Q 25 4 1

/ vESSEt 3 16 24 ' 32s h . . . l. TN/ STABILIZER i

23 , IJ 15 h 22 h. 31,, y S ACRIFIC SHIELO AL g

2 ,,,,,$,i 21 30 h J BASE OF RE ACTCP i i SK!RT 8

i 20 O 29,g/ REACTOR I 8 i ' PEDESTAL I i 7 1 13 - . 28 9 1gj , j .

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//O//4 / / /// usi//// ///////// ///

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ENRICO FEMill ATOMIC POV.'ER PLANT UNIT 2

,' FINAL SAFETY ANALYSIS REPORT FIGURE 3.7-15 REACTOR / AUXILIARY BUILDING HORIZONTAL OYNAMIC MCDEL i .

SARGENT & LUNDY REPORT NO. SL-2682 E

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c;= - - - .

~ ~ ;l ii LINER DESIGt mm 0 3-ii,-

mou o A- ' I

.- CODES I, o ASME BOILER & PRESSURE VESSEL CODE SECTION VIII j DIVISION I 1

?

o ACI 307-68 rec 0MMENDED PRACTICE FOR FORMWORK CESIGN 1

LOADS l

o IHERMAL OPERATING LOADS

~

0 125 -150 F o IHERMAL ABNORMAL LOADS -

i 212 F INSIDE, 150 F0 OuTSIDE }

1 o CONSTRUCTION LOADS (CONCRETE FORM) {

o SEISMIC LOAD OF PACK ON FLOOR i

.I ACCEPTAfiCE CRITERIA o NORMAL IENSION AND COMPRESSION F

T

=F C

=. Fy

~

' q o ABNORMAL IENSION AND COMPRESSION 4 F

T

=F C .9 F y A  ;

o NORMAL MEMBRANE STRAIN bl,k.3 -

,_m ._ ,

= .003 IN/IN

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m. y...

{.,

o ABNORMAL MEMBRANE STRAIN C" ~ ' ~ ' 1

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= .005 IN/IN -'

a, ,

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=

III. LOSS OF FUEL POOL COOLING ACCIDENT A. MAXIMUM FUEL POOL HEAT LOAD AND BOILING RATE KEY ASSUMPTIONS 0 COMPLETE LOSS OF POOL COOLING 0 POOL FILLED ABOVE CAPACITY WITH 3.5 CORES AT THE RATE OF 1/4 CORE PER YEAR 0 DECAY HEAT RATE BASED ON BTP ASB 9-2 ASSUMING AN IRRADIATION PERIOD OF 4 YEARS AND A COOLING PERIOD OF 30 DAYS BEFORE DISCHARGE TO THE SPENT FUEL POOL 0 BULK Prol TEMPERATURE REACHES 212 F COINCIDENTLY WITH IHE LAST 1/4 CORE DIS-CHARGE (I.E., COINCIDENTLY WITH PEAK POOL HEAT LOAD)

O POOL HEAT LOAD AND BOILING RATE REMAIN CONSTANT AT THE CALCULATED MAXIMUM VALUES FOR 30 DAYS 0 MAKEUP WATER IS PROVIDED TO MAINTAIN IHE

['00L INVENTORY

~

0 NO CREDIT IS IAKEN FOR IHE HEAT ABSORBED BY IHE MAKEUP WATER RESULTS 0 POOL MAXIMUM HE.7 LOAD -- 2.11 MwT i O POOL MAXIMUM BOILING RATE -- 2,06 LBM/SEC 0 POOL HEAT LOAD FOR NcaxAL 2.0 CORE LIMIT - 1.778 MwT

-c~~~-- . .. __ _ _

e .t III. LOSS OF FUEL POOL COOLING ACCIDENT (CONTINUED)

B. RADIOLOGICAL CONSEQUEtlCES DATA / ASSUMPTIONS:

0 FUEL POOL BOILING RATE 2,06 Lan/SEC 0 FUEL POOL WATER VOLUME 48,000 FT 0 INITIAL FUEL POOL WATER 60 uCI/s I-131 CONCENTRATION O IODINE WATER-IO-STEAM PARTITION 10 FACTOR 0 IODINE REMOVAL EFFICIE*JCY (SGTS) 99% ,

O DISPERSION DATA (X/0 (SEC/M )):

0-2 HOURS aSITE BOUNDARY 6.10 x 10-5 0-8 HOURS aL.P.Z. 8.58 x 10-6 8-24 H0uas aL.P Z. 1.81 x 10-6 24-96 HOURS aL.P.Z. 1.08 x 10-6 96-720 HOURS al.P.Z. 3,30 x 10-7 O DOSE EVALUATION PER REGULATORY GUIDE 1.3 RESULTS:

0 2-H0ua THYROID DOSE a SITE BOUNDARY 17 MREM i

0 30-DAY IHYROID DOSE a L.P.Z. 20 MREM )

WELL WITHIN 10 CFR 100 DOSE CRITERIA

,---- -- -. _ _