ML17346A280
| ML17346A280 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/29/1984 |
| From: | Gurley M, Sarah Turner AFFILIATION NOT ASSIGNED |
| To: | |
| Shared Package | |
| ML17346A276 | List: |
| References | |
| SS-153, NUDOCS 8404120145 | |
| Download: ML17346A280 (36) | |
Text
SS-153 CRITICALITY ANALYSIS OF THE TURKEY POINT PLANTS UNITS 3 8I 4 STORAGE RACKS WITH INCREASED ENRICHMENT Prepared for the Florida Power 8 Light Co.
by S. E. Turner, Ph.D., P.E.
M. K. Gur1ey February 1984 r s4o4o4 seoeisoies 05000250 PDR *DDCK I P PDR
Pp L
TABLE OF CONTENTS Page
1.0 INTRODUCTION
AND
SUMMARY
............,....................... 1 2.0 CRITERIA AND METHODOLOGY FOR CRITICALITY ANALYSES.........,. 3 2.1 2.2 Design Bases......................-..
Reference Fuel Assembly................
" ~ ~
~ ~
3 2.3 Reference Fuel Storage Cell............ ~ ~ 4 2.4 Analytical Methods..................... ~ ~ 7 2.4.1 Calculational Models......... ~ ~ ~ ~ ~ ~ ~ 7 2.4.2 Calculational Bias and Uncert ainty ~ ~ 8 3.0 CRITICALITY ANALYSIS OF SPENT FUEL STORAGE RACKS............ 10 3.1 Summary of Criticality Analyses................. 10 -~.
3 ' Uncertainties Due to Manufacturing Tolerances.... 13 3.2.1 Fuel Enrichment and Density Variations. ~ ~ 13 3.2.2 Inside Cell Dimensional Tolerance...... ~ ~ 14 3.2.3 Storage Cell Lattice Spacing Variation. 14 3.2.4 Stainless-Steel Thickness Variations... 14 3.2.5 Eccentric Positioning of Fuel Assembly within Storage Rack.............,...... 14 3.3 Abnormal and Accident Conditions................. o ~ ~ 15 3.3.1 Temperature and Water Density Effects.. 15 3.3.2 Fuel Assembly Abnormally Located Outside Storage Rack................... 15 4.0 CRITICALITY ANALYSIS OF NEW FUEL STORAGE RACKS.............. 18 REFERENCES
LIST OF TABLES No. ~Pa e 1 OPTIMIZED FUEL ASSEMBLY DESIGN SPECIFICATIONS................... 6 2
SUMMARY
OF UNCERTAINTIES IN k DUE TO TOLERANCES................ 10 3 INFINITE MULTIPLICATION FACTORS OVER ANTICIPATED RANGE OF FUEL ENRICHMENTS AND DENSITIES.................................. 13
LIST OF FIGURES No. ~Pa e Reference fuel assembly and configuration of spent fuel storage rack for the Turkey Point Plant.........................
Infinite multiplication factor of spent fuel storage rack for various U-235 loadings in fuel assembly......................... 11 3 Effect of coolant temperature on reactivity of spent fuel storage racks................................................... 12 4 Variation in kwith fuel assembly spacing (infinite lattice without stainless steel)........................................ 17 Fresh fuel storage rack configuration and analytical model...... 19 6 Reactivity effect of low-density moderator in fr'esh fuel storage rack with fuel of 4.5X enrichmenmt...................... 2C
1.0 INTRODUCTION
.AND
SUMMARY
Both the new fuel and spent fuel storage racks in the Turkey Point Plant, Units ,3 5 .4, @re currently licensed to store fuel of 43.9 grams U-235 per axi al centimeter of fuel assembly corresponding to 3. 5 wt. X U-235 ini al ti enrichment. The previous criticality analysis, submitted in support of the current Technical Specification limit on fuel enrichment, documented a neutron multiplication factor substantially below the NRC limiting reactivity value of 0.95 including all uncertainties. The evaluation reported here was prepared to justify the criticality safety of an increase in the Technical Specifica-tion limit on fuel enrichment in the existing storage racks for both new and spent fuel.
Results of the present evaluation confirm that the maximum reactivity gf the spent fuel storage racks will be less than 0.95, including all uncertain-ties, with the racks fully loaded with fuel containing 52.40 grams U-235 per axial centimeter of fuel assembly and flooded with unborated water at a tem-perature corresponding to the highest reactivity, provided the U02 stack density is no less than 10.08 g/cm (93$ of theoretical density). The lim-iting axial U-235 loading includes tolerances on fuel density and enrichment and corresponds to a nominal enrichment of 4.085 wt.% U-235 at a U02 pellet density of 97~ of theoretical.
Criticality safety of the new fuel storage rack (a separate facility located near, but independent of, the spent fuel pool) was evaluated for fuel of 57.72 grams U-235 per axial centimeter, corresponding to a nominal enrich-ment of 4. 5$ at a U02 pellet density of 97$ of theoretical. Transport theory calculations confirm that* +e neutron multiplication factor under optimum moderating conditions (e.g., fog, spray, or foam) is substantially less than the limiting value of 0.98 specified in SRP 9.1.1, "New Fuel Storage."
Other than criticality safety, increasing the enrichment capability of the spent fuel storage pool and new fuel storage racks does not introduce any significant new or unreviewed safety considerations. On the basis of the analyses and evaluations presented herein, it is concluded that the spent fuel
pool and the new fuel storage racks can safely accommodate fuel of 52.40 and 57.72 grams 0-235 per axial centimeter of fuel assembly respectively with no significant new or unreviewed hazard considerations under the guidelines of 10CFR50.92(.c).
2.0 CRITERIA AND METHODOLOGY FOR CRITICALITY ANALYSES 2.1 ~D The objective in the spent fuel storage racks for the Turkey Point Plant is to assure that a neutron multiplication factor (keff) equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest antic-ipated reactivity and flooded with unborated water at a temperature corre-sponding to the highest reactivity. The maximum calculated reactivity includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined, such that the true keff will be equal to or less than 0.95 with a 95$ probability at a 95$ confidence level.
4 Applicable codes, standards and regulations, or pertinent sections thereof, include the following.
~ General Desi gn Cri teri on 62 - Preventi on of Criti cal ity in Fuel Storage and Handling.
~ NRC letter of April 14, 1978, to all Power Reactor Li censees - OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.
~ USNRC Standard Revi ew Pl an, NUREG-0800, Secti on 9.1.1, New Fuel Storage, and Section 9.1.2, Spent Fuel Storage.
~ Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis (proposed), December 1981.
e Regulatory Guide 3.41, Validation of Calculational Method for Nuclear Criticality Safety (and related ANSI N16.9-1975).
~ ANSI N210-1976, Design Objectives for Light Mater Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
~ ANSI N18.2-1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Hater Reactor Plants.
To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made.
~ Moderator is pure, unborated. water at a temperature corresponding to the highest reactivity.
i Lattice of storage racks is infinite in all directions; i.e., no credit-~ taken for axial or radial neutron leakage (except in the consideration of certain abnormal/accident conditions).
~ Neutron absorption. in minor structural members is neglected; i.e.,
spacer grids are replaced by water,
~ Pure zi rconium i s used for cl adding, control rod gui de tubes, and instrument thimbl es; i.e.', hi gher neutron absorpti on of al 1 oying materials in Zircaloy is neglected.
2.2 Reference Fuel Assembl The reference design fuel assembly, illustrated in Fig. 1, is a 15 x 15 array of fuel rods (Westinghouse design), with 21 rods replaced by 20 control rod gui de tubes and one instrument thimbl e. Two al ternati ve fuel assembly designs have been used in the Turkey Point reactors: an optimized fuel assembly design with Zi rcaloy grids and an earlier design using inconel grids. The optimized fuel assembly is more reactive and has been used as the III reference in the fuel rack criticality analyses. Table 1 summarizes the optimized fuel assembly design specifications and expected range of signi-ficant fuel tolerances.
2.3 Reference Fuel Stora e Cell The nominal spent fuel storage cell model used for the criticality anal-yses is shown in Fig. 1. The rack is composed of 0.25-in. stainless-steel boxes of 8.790-in. inside dimension. The fuel assemblies are centrally located in each storage cell on a nominal lattice spacing of 13.659 in. The outer water space constitutes a flux-trap between the two steel plates. For two-dimensional X-Y analysis, a zero current (white albedo) boundary condition was applied in the axial direction and at the centerline through the outer water space (flux-trap) on all four sides of the cell, effectively creating an infinite array of storage cells.
I S. 659
+0. 588 8.790,
+O. l 25 STAlNLESS STEEl BOX O. 2 5 THl CK 000000000000000 000000000000000 0000000000000:.
00000000000000'000e00000eOOOO 00000000000000 000000000000000 0006000000000 000000000000000 00000000000000 WATER GAP 4.369 0000000000000 +0. 58 8 00000000000000 00000000000 000000000000000 000000000000000
( HOT TO SCALE)
Fig. 1 Reference fuel assemb1y and configuration of spent fue1 storage rack for the Turkey Point P)ant:
Table .1 OPTIMIZED FUEL ASSEMBLY DESIGN SPECIFICATIONS Fuel Rod Data
=:
Outside dimension, in. 0.422 Cladding thickness, in. 0.0243 Cladding material Zr-4 Pellet diameter, in. 0.3659 Axial dishing factor 0.988 UOz density, g/cm 10.08 - 10.514
$ T.D. 93 - 97 Fuel Assembly Data Number of fuel rods 204 (15 x 15 array)
Fuel rod pitch, in. 0.563 Active fuel height, in. 144 Control rod guide tube Number 20 O.D., in. 0.533 Thickness, in. 0.017 Material Zr-4 Instrument thimble Number 1 O.D., in. 0.533 Thickness, in. 0.017 Material Zr-4
2.4 Anal tical Methods 2.4.1 Calculational Models bilityy Four different methods of calculation were used to enhance the credi-of the analysis and to provide assurance that the true reacti vity will be less than the limiting value of Oe95 including uncertainties. These methods of calculation include the following.
e CASMO - a two-dimensional multigroup transport theory code (based upon capture probabilities), which provides the capability for a detailed geometric description of the storage cell and each fuel rod.
~ AMPX-KENO - a multi group Monte Carlo code package, 3 s 4 using the 123-group GAM-THERMOS cross-section set developed by ORNL, and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral.
treatment). AMPX-KENO has been benchmarked against a number of criti-'al experiments (Refs. 5, 6, and 7) with generally good agreement for ti most cri cal experiments analyzed, although both Ref s. 6 and 7 indi-cate an underprediction in reactivity in arrangements with large water gaps between fuel assemblies.
~ AMPX-XENO - the Monte Carlo technique described above, but using the more recent 27-group SCALE cross-section set deva a[ad by ORNL for criticality safety analysis. Benchmark calculations 'ndicate that the 27-group SCALE cross-section set consistently underpredicts reac-tivity by '0.012 hk.
~ Diffusion/Blackness Theory - a calculational technique based upon the multigroup cell homogenization code~ NULIF, to calculate diffusion theory constants for input to PDgg7. 2 ' small correction, based upon tionss blackness theory, was applied to the macroscopic absorption cross-section calculated by NULIF for stainless steel.
For investi gation of reactivity effects due to uncertainties (e.g.,
mechanical and fabrication tolerances), the CASMO code was used to calculate the small incremental reactivity changes. Diffusion/blackness theory calcula-were used to estimate the reactivity effects of abnormal/accident con-ditions.
2.4.2 Calculational Bias and Uncertainty The infinite multiplication factor (k) for the Turkey Point spent fuel storage rack i~ased upon CASMO calculations with fuel of the highest antici-pated reactivity. Supporting calculations with both the 123-group and 27-group AMPX-KENO code package confirm that the CASMO calculations are conservative. A diffusion/blackness theory calculation (NULIF-PDgp7) provides further confirmation of the conservatism in the reference CASMO calculation.
To illustrate the inherent conservatism of CASMO calculations, one case (57.72 grams U-235 per axial centimeter with the minimum cell box inside dimension) was selected for intercompari son. Results of criticality calculations with the four independent methods of analysis for this case are as follows:
Method Calculated k Bias Corrected k CASMO 0.9404 0.9404 123-gp AMPX-KENO 0.9052 + 0.0037* 0.0315 0.9367 + 0.0037*
(150,000 histories) (95$ /95$ )
27-gp AMPX-KENO 0.9095 + 0.0063* 0.012 0.9215 + 0.0063*
(50,000 histories) (95$ /95$ )
Diffusion/Blackness Theory 0.9340 0.9340 The 123-group AMPX-KENO calculational model has been benchmarked 7 against critical experiments as nearly representative as possible of the Turkey Point fuel racks. These benchmark calculations indicate a nominal bias of 0.000 +
0.003 (955/95$ ) plus a correction for the water gap thickness between storage cells. Linear extrapolation of the trend identified in reference 7 to the 4.7-in. water gap of the Turkey Point spent fuel rack results in a bias cor-rection of 0.0315 bk, although linear extrapolation probably overestimates the water gap correction. Similar benchmark calculations of the 27-group AMPX-
- With one-sided tolerance factor for 955 probability at 955 confidence level.
KENO calculational. model, reported in references 9 and 10 (and confirmed by independent calculations), indicate a bias of 0.012 bk, with no apparent trend with water gap thickness observed at least for the largest ~ater gap (2.58-in.) measured in the critical experiments. Thus, the storage rack kde-scri bed above, as cal cul ated by the Honte Carl o techni que, probably 1 i es between 0.9215 and 0.9367. Both values confirm the conservatism of the CASMO calculation (k = 0.9404).
Diffusion/blackness theory cal cul ati ons woul d normally be expected to provide a reliable estimate of the rack k, since the stainless-steel wall of the storage cells is a relatively weak absorbing medium. The value calculated by di f fusion/blackness theory (kof 0.9340) tends to further confirm the conservatism of the CASHO calculated value (0.9404). CASMO benchmark calcula-tions on critical experiments representative of the Turkey Point racks {refer-ence 2, Section.2.1.3) suggest an overprediction of 0.006 hk which, if applied to the reference CASMO calculation', would indicate good agreement with the diffusion/blackness theory value. Nevertheless, for conservatism and to assure the true reactivity is less than the calculated value, the higher k as calculated by CASMO, was used as the reference value for the Turkey Point.
spent fuel storage rack. All three of the alternate methods of analysis indicate a lower value and therefore confirm that the reference CASHO value is conservative.
3.0 CRITICALITY ANALYSIS OF SPENT FUEL STORGE RACKS 3.1 Summar of Criticalit Anal ses Criticality analyses for the Turkey 'Point spent fuel storage racks were performed for several fuel densities and U-235 loadings, (in grams per axial centimeter of fuel assembly) and plotted as shown in Fig. 2. These data show that, for a given U-235 axial loading, the higher reactivity results for the lower assumed U02 stack density of 10.08 grams per axial centimeter. To this latter curve was added the total uncertainty in kassoci ated with manufacturing tolerances, 0.0243 hk, as summarized in Table 2, to generate the upper curve in Fig. 2 identified as the "maximum with uncertainties." This curve was then read to determine the limiting U-235 loading of 52.40 grams per axial centimeter corresponding to the maximum allowble k of 0.95.
Table 2
SUMMARY
OF UNCERTAINTIES IN k DUE TO TOLERANCES Reference Uncertainties Fuel enrichment 40.0008 Section 3.2.1 Fuel density +0.0006 Section 3.2.1 Cell inside dimension %0.0020 Section 3.2.2 Lattice spacing %0.0230 Section 3.2.3 SS tolerance 40.0015 Section 3.2.4 Statistical combination +0.0232 Eccentric positioning +0.0011 Section 3.2.5
- tionss, Maximum uncertainty +0.0243 The design basi s temperature for these calculations was 65.6'C (150'F) which, as shown in Fig. 3 is the coolant temperature of the highest neutron multiplication factor. Lower pool temperatures expected for normal opera-as well as neutron leakage from the finite size racks, provide addi-tional margin in keff below the limiting value of 0.95 for k used in the analysis.
10
0 98 0 96 8
0.94 0 93 A~LAATQM 0 92 51 52 53 54 55 58 57 58 GRAMS. U-235 PER, AXIAL CM Fig. 2 Infinite multip'lication factor of spent fuel storage rack for various U-235 loadings in fuel assembly.
11
0 Cl hl tL o -IQ 8 20
-30 20 30 $0 50 60 70 80 90 I 00 I I 0 I20 TEMPERATURE, o C Fig. 3 fffect of coolant temperature on reactivity of spent fuel storage racks.
Thus, to assure a maximum keff including uncertainties of less than 0.95, a U-235 axial loading of 52.40 grams per axial centimeter is the maximum which may be accommodated in the Turkey Point spent fuel storage racks. This load-ing may be- realized by fuel of 4.261% U-235 enrichment at 10.08 grams per axial centimeter U02 stack density (93$ T.D. ) or lower enrichments at higher U02 densities {e.g 4.085% enrichment at a U02 stack density of 10.514 grams per axial centimeter, 97% T.D.).
Credible abnormal or accident conditions will not result in exceeding the 1 iml ti ng kef f of 0. 95, with credit for the Presence of sol ubl e Poi son (nominally 1950 ppm boron).
3.2 Uncertainties Due to Manufacturin Tolerances 3.2. 1 Fuel Enrichment and Density Variations densities The maximum loading of 52.40 grams U-235 per axial centimeter can be realized over a range of enrichments and U02 stack . Table 3 identi-fies the range considered and gives the k values for three combinations of enrichment and density.
Table 3 INFINITE MULTIPLICATION FACTORS OVER ANTICIPATED RANGE OF FUEL ENRICHMENTS AND DENSITIES Density Enrichment, Axi al Loading U02
% U-235 ~l" k
.(CASMO) 97 10.514 4.085 52. 40 0.9232 95 10.297 4.173 52.40 0.9244 93 10.080 . 4.261 52.40 0.9256 These data show that the hi ghest koccurs at a U02 densi ty of 93$ of theo-retical and an enrichment of 4.261 wt.$ U-235. Thus, the low density case has been assumed as the design basis for the criticality safety evaluation.
Higher U02 densities, at the same axial loading in grams U-235 per axial centimeter, will always yield a lower k 13
0 In addition, 'there is a certain level of confidence to which the fuel enrichment and density are known. To evaluate the reactivity uncertainty, is assumed that the U02 density is known to +0.05 g/cm and enrichment to 10.02 wt.$ 4J-235. Evaluating the uncertainty for these tolerance limits (by di fferential CASMO cal cul ations) yields an uncertainty of +0.0006 hk for density and +0.0008 ~k for enrichment.
3.2.2 Inside Cell Dimensional Tolerance The stainless-steel inner box dimension, 8.790 + 0.125 in., defines the inner water thickness between the fuel and the inside box. For the tolerance of +0.125 in. on the box inside dimension, the calculated
\
uncertainty in k is +0.0020 ~k, with k increasing as the inner stainless-steel box dimen-sion increases.
3.2.3 Storage Cell Lattice Spacing Variation The storage cell lattice spacing between fuel assemblies is nominally 13.659 in., positioned by a lattice of support grids intended to provide a nominal water gap between cells of 4.369 in. Receipt inspection of the racks confirmed that the water gap between adjacent storage cells is greater than 3.781 in. for all locations, indicating a toleranace of %0.588 in. in lattice spacing. Calculations with this minimum spacing between cells resulted in an uncertainty in k of +0.0230 ~k due to the tolerance in lattice spacing.
3.2.4 Stainless-Steel Thickness Variations The nominal stainless-steel . box thickness is 0.25 in. The maximum posi-
'tive effect on k of the expected stainless-steel thickness tolerance varia-tion (+0.01 in.) was calculated to be +0.0015 bk.
3.2.5 Eccentric Positioning of Fuel Assembly within Storage Cell The fuel assembly is normally located in the center of the storage cell. Nevertheless, calculations were made with adjacent fuel assemblies 14
moved into the corner of the storage cell (four-assembly cluster at closest approach), resul ting in a small posi ti ve effect on k(0.0011 hk) . Fuel assembly bowing will produce a smal l er posi ti ve reacti vi ty e ffeet locally-;
The calculated reactivity increment due to eccentric positioning is considered an additive allowance, although eccentric positioning (if any) would normally be expected to be randomly distributed throughout the storage rack.
3.3 Abnormal and Accident Conditions 3.3.1 Temperature and Water Density Effects Increasing or decreasing temperature from the nominal temperature of 150'F (65.56'C) is calculated to decrease k in unborated water as indicated in Fig. 3 (reactivity effects calculated by CASMO). At 120'C (248'F), intro-.
ducing voids in the water internal to the storage cell (to simulate boiling) further reduced k indicating a negative void coefficient of reactivity at the boiling temperature. Voids due to boiling will not occur in the outer (flux-trap) water region.
3.3.2 Fuel Assembly Abnormally Located Outside Storage Rack To investigate the possible effect of a fuel assembly abnormally located outside the rack, diffusion calculations were made for unpoisoned assemblies separated only by water. Figure 4 shows the results of these calculations.
From these data, the infinite multiplication factor will be less than 0.95 for any fuel assembly spacing greater than -15 in. in the absence of any soluble poison or neutron-absorbing material other than water between assemblies.
For a drop on top of the rack, the fuel assembly will come to rest hori-zontally on top of the rack with a minimum separation greater than 15 in.
( -24 in.). Consequently, fuel assembly drop accidents will not result in an increase in reactivity above that calculated for the infinite nominal design storage rack.
15
An extraneous fuel assembly cannot physically be positioned outside the rack between the rack and pool wall or between rack modules. However, it is possible, although not likely, to position an extra fuel assembly adjacent to the rack in the region of the cask area. Two-dimensional PDl} calculations show that a fresh fuel assembly positioned adjacent to the storage rack can increase the reactivity to 0.957 in the absence of soluble poison. However, soluble boron of 1950 ppm is normally present in the spent fuel pool (for which credit is permitted under accident conditions)* and would reduce the maximum k to substantially less than 0.95. Consequently, it is concluded that the postulated accident conditions will not adversely affect the criticality safety of the Turkey Point spent fuel storage racks.
An implementing Technical Specification for 1950 ppm soluble boron has been submitted via L-84-71, dated Viarch 14, 1984.
16
I j
1.O8 1.04 I I
I e
I 1.02 1.00 0.98 0 98 0.94 0 92 12 13 14 16 18 1'7 18 19 20 21 FUEL, ASSEMBLY SPACING INCHES Fig. 4 Variation in k~ with fuel assembly spacing (infinite lattice without stainless steel).
4.0 FRESH FUEL STORAGE RACKS The fresh fuel storage racks for the Turkey Point Plant consist of an "L"-shaped ..array of storage cells containing 54 cells on a 21-in. atti ce 1 spacing. Figure 5 illustrates the fresh fuel storage cell arrangement and shows the geometry used in the criticality analysis. Although fuel is nor-mal ly stored in the dry condition, the cri ti cali ty analysis considered flooding with clean, unborated water ranging in density from 1.0 to very low hypothetical values (e.g., fog, mist, or foam) . Preliminary survey calcu-lations with diffusion theory suggested a second maximum in reactivity peaking at a hypothetical water density of -10$ -15$ .
The criticality safety of the new fuel storage racks was evaluated for fuel of 57. 72 grams U-235 per axi al centimeter of assembly ( - 4.5%
enrichment). Since 'diffusion theory is known to be inadequate in very dry lattices, three-dimensional AMPX-KENO cal cul ations were used in the low-moderator-density region to define the maximum keff under optimum moderating conditi ons. For these cal cul ati ons, the array of fuel storage cell s, as indicated in Fig. 5, was assumed to be reflected by full-density water on the outer boundaries and on both top and bottom of the array. Low-density water was used within the storage boxes and between the array of storage cells. The XSDRNPM routine in the AMPX code package was used to homogeni ze the fuel assemblies for each moderator density calculated and to generate the weighted cross-section set for use in KENO.
Fi gure 6 shows the cal cul ated ke ff values as a functi on of moderator density within and between the storage cells. These calculations indicate a low-density maximum reactivity of -0.925 occurring at a water density of 0.10
'g/cm . This low-density maximum reactivity is approximately the same as that for the fully flooded condi ti on. In ei ther event; the reacti vi ty i s sub-stantially less than the limiting reactivity of 0.98 specified in SRP 9.1.1 under optimum moderating conditions. Hence, it is concluded that unirradiated fuel with a loading of 57.72 grams U-235 per axial centimeter (4.5% enrich-ment) may be safely stored in the new fuel racks of the Turkey Point Plant.
18
~21 12 WATER REFLECTOR I I I2 I2 WATER WATER LOW DENSITY MODERATOR 8 ETWE EN ASSEMBLIE S I 2 WATER Fig. 5 Fresh fuel storage rack configuration and analytical model.
I I.O ~ ~
qual
.9 I
1 0
.7
.6
.2 .3 .4 .5 .6 .7 .8 WATER DENSITY ~ gms/cc Fig. 6 Reactivity effect of low-density moderator in fresh,'fuel storage rack.
with fuel of 4.5X enrichment.
0 REFERENCES A. Ahlin and M. Edenius, CASMO - A Fast Transport Theory Depletion Code for LWR Analysis, ANS Transactions, Vol. 26, p. 604, 1977.
CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual, Rev.
A, Control Data Corporation, 1982.
- 2. E. E. Pilat, Methods For the Analysis of Boiling Water Reactors (Lattice.
Physics), YAEC-1232, Yankee Atomic Electric Co., December 1980.
- 3. Green, Luci ous, Petri e, Ford, White, Wri ght, PSR-63/AMPX-1 (code package), AMPX Modul ar Code System for Generating Coupled Mul group ti Neutron-Gamma Libraries from ENDF/B, ORNL-TM-3706, Oak Ridge National Laboratory, March 1976.
L. M. Petrie and N. F. Cross, KENO-IV, An Improved Monte Carlo Critical-ity Program, ORNL-4938, Oak Ridge National Laboratory, November 1975.
- 5. S. R. Bierman al., Critical Separation Between Subcritical g Enriched Clusters of 4.29 wt% U UO Rods in Water with Fixed Neutron Poisons, NUREG/CR-0073, Battelle Pacific Northwest Laboratories, May 1978, with errata sheet issued by the USNRC August 14, 1979.
- 6. M. N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock 8 Wilcox Company, July 1979.
- 7. S. E. Turner and M. K. Gurley, Evaluation of AMPX-KENO Benchmark Calcu-
~Ei i, lations for High Density Spent Fuel Storage Racks, Nuclear Science and 80(2): 230- 37, t' 1 w.
- 8. R. M. Westfall et al ., SCALE: A Modul ar Code System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, 1979.
- 9. R. M. Westfall and J. R. Knight; Scale System Cross-Section Validation wi th Shi ppi ng-Cask Critical Experiments, ANS Transact i ons, Vol . 33,
- p. 368, November 1979.
- 10. B. F. Cooney, T. R. Freeman, and M. H. Lipner, Comparisons of Experiments and Cal cul ati ons for LWR Storage Geometri es, ANS Transactions, Vol. 39, p. 531, 1981.
W. A. Wittkopf, NULIF - Neutron Spectrum Generator, Few-Group Constant Generator and Fuel Depletion Code, BAW-426, The Babcock 5 Wilcox Company, August 1976.
- 12. '. R. Cadwell, PDg-7 Reference Manual, WAPD-TM-678, Bettis Atomic Power Laboratory, January 1967.