Text

Criticality Analysis of Farley Units 1 & 2 Fresh & Spent Fuel Racks
	ML20217C419
Person / Time
Site:	Farley
Issue date:	03/31/1991
From:	Bell R, Durston C, Fecteau M; GENERAL ELECTRIC CO., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20217C378	List: ML20217C373; ML20217C388; ML20217C419;
References
	NUDOCS 9107150260
	Download: ML20217C419 (50)
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l CRITICALITY ANALYSIS OF THE
{ FARLEY UNITS 1 & 2 FRESH AND SPENT FUEL RACKS 4
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March 1991 I
I M. W. Fecteau R. M. Bell C. Durston F. K. Torres I
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TABLE OF CONTENTS I 1.0 Introduction 1.1 Design Description 1.2 Design Criteria 1
2
.. ... ... ..... .,,..... .. ...... 2 2.0 Analytical Methods .. . ........... ... .................. 3 2.1 Criticality Calculation Methodology ... ..,.... ............ 3 2.2 IFBA Credit Reactivity Equivalencing ..... .. .............. 4 3.0 Spent Fuel Rack Criticality Analysis .............. ....... . .... 5 I', 3.1 KENO Reactivity Calculations ........ ,, . . .. . .... 5 l 3.2 IFBA Cr:dit Reactivity Equivalencing ..... ...... ...,. .. 7 3.2.1 Reactivity Equivalencing Analysis ..,.,,,,............ 8 l I 3.2.2 infinite Multiplication Factor 3.3 Sc'1sitivity Analysis 3.4 Postulated Accidents
. 10 10 l
.... ... ... .. ............ 11 l l
4.0 Criticality Analysis of Fresh Fuel Racks .... ..., .. . . ... , . 13 4.1 Full Density Moderation Analysis .., ,,...,,......, , .. 13 I 4.2 Lov' Density Optimum Moderation Analysis 4.3 Postulated Accidents ............,..... ,,,..........
...,,. ....... . 14 15 ;
5.0 Application of Spent Fuel Rack IFBA Credit Limit to Fresh Fuel Rack . 17 6.3 Summary of Criticality Results ...........,.... . . .......,... 18 i
Bibliography ............................... ....... .......... 34 I
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Table of Contents I i
-- . . . . - . - - . - . - - - - . . - - . . - . - . . - - ~ - . . . .-~- - - .. --.
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g j- LIST OF TABLES r
'g Table 1. Fuel Parameters Employed in Criticality Analysis ... . .... 19 R Table 2. Benchmark Critical Experiments [5.6) ............... . 20 Table 3. Comparison of PHOENIX lsotopics Predictions to Yankee Core 5 g Measurements .............. ....... .......... 21 W Table 4. Benchmark Critical Experiments PHOENIX Comparison ... .. 22 Table 5. Data for U Metal and UO: Critical Experiments ..... ..... 23 I Table 6. Farley Units 1 & 2 Spent Fuel Storage Minimum IFBA Requirements ..... .......... ................. 25 I
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, mst o,1ab,es -
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l LIST OF ILLUSTRATIONS
! Figure 1. Farley Units 1 & 2 Spent Fuel Pool Storage Cell Nominal Dimensions . . ... .... ....... ....... ... 26 Figure 2. Farley Units 1 & 2 Fresh Fuel Rack Storage Cell Nominal I Figure Figure Dimensions . . . ....
3. Ferley Units 1 & 2 Fresh Fuel Rack Layout
4. Farley Units 1 & 2 Spent Fuel Storage Minimum IFBA 27 28 I Figure Requirements . . ,. ..
5. Sensitivity of K.ee to Enrichment in the Farley Units 1 & 2 Spent Fuel Racks . . . . . ...
.... 29 30 Figure 6. Sensitivity of K.#, to Center-to-Center Spacing in the Farley Units 1 & 2 Spent Fuel Racks, ,
. . . .. .... ...... 31 Figure 7. Sensiti 'ty of Keet to 8 ~ Loading in the Farley Units 1 & 2 Spent I Figure Fuel Ri -ks
8. Sensiti y Fr'.,sh F 01 Rocks of K.## to Water Density in the Farley Units 1 & 2
.. 32 33 I
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List of Illustrations sii
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1.0 INTRODUCTION
I This report presents the results of the criticality re-analysis of the Farley Units 1 & 2 fresh fuel and spent fuel storage racks.
The fresh fuel rack design considered herein is an existing array of unpoisoned j 1 racks, previously qualified for storage of Westinghouse 17x17 Standard (STD) I fuel assemblies with enrichments up to 4.3 w/o U* The spent fuel rack design analyzed herein is an existing array of poisoned racks, also previously qualified I for storage of Westinghouse 17x17 Standard (STO) fuel assemblies with enrichments up to 4.3 w/o U". This report does not impact the previous li- l censing basis of STD fuel in the spent fuel racks.
The Farley Units 1 & 2 fresh and spent fuel racks are being reanalyzed to allow storage of Westinghouse 17x17 OFA and VANTAGE 5 fuel assemblies at the nighest possible enrichment. The spent fuel rack analysis will show that Westinghouse 17x17 OFA and VANTAGE 5 fuel assemblies with nominal enrichments up to 3.90 w/o can be safely stored in the spent fuel rack utilizing I all locations. Furthermore, the analysis will show that storage of assemblies with nominal enrichments above 3.90 w/o and up to 5.00 w/o U* is also ac-ceptable in the spent fuel rack by taking credit for Integral Fuel Burnable Absorbers (IFBAs). The fuel assembly IFBAs consist of neutron absorbing ma-terial applied as a thin ZrB2 coating on the outside of the UO2 fuel pellet. As a result, the neutron absorbing material is a non-removable or integral part of the fuel assembly once it is manuf actured.
The Parley Units 1 & 2 fresh fuel rack analysis will show that Westinghouse I 17x17 STD, OFA and VANTAGE 5 fuel assamblies with nominal enrichments up to 4.00 w/o can be safely stored in the fresh fuel rack utilizing all locations.
Storage of fuel assemblies with nominal enrichments above 4.80 w/o and up to I 5.00 w/o U* will also be.shown to be acceptable by taking credit for the same IFBAs which are required to satisfy the spent fuel rack limit.
This Farley Units 1 & 2 spent fuel rack criticality analysis is based on main-I taining K.,e 5 0 6" for storage of Westinghouse 17x17 OFA and VANTAGE 5 fuel assemblies. The tresh fuel rack analysis is based on maintaining Kve 5 0.95 for I storage of Westinghouse 17x17 STD, OFA, and VANTAGE 5 assemblies under full water density conditions and 5 0.98 under low water density (optimum moderation) conditions.
The fuel parameters relevant to this analysis are given in Table 1 on page 19.
I.
I introduction 1
1 1
I l 1.1- DESIGN DESCRIPTION The Farley Units 1 & 2 spent fuel storage cell design is depicted schematically
, in Figure 1 on page 26 with nominal dimensions given on the figure. The fresh fuel storage cell design is depicted schematically on Figure 2 on page 27 and the fresh fuel rack layout -is shown on Figure 3 on page 28.
1.2 DESIGN CRITERIA g
Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the
'I minimum separation between fuel assemblies and inserting neutron poison be-tween fuel assemblies.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the ef fective neutron multiplication f actor, K.fe, of the fuel assembly array will be less than 0.95 as recommended by ANSI 57.?-1983, ANSI 57.3-1983 and Reference 1. The 0.95 K.ee limit applies to both the fra
fuel racks under all conditions, except for the fresh fuel rack under low m er density (optimum moderation) conditions, where the K.it limit is 0.98 as recommended by NUREG-0800.
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Introduction 2
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I 2.0 ANALYTICAL METHODS l 2.1 CRITICALITY CALCULATION METHODOLOGY The criticality calculation metnod and cross-section values are verified by comparison with critical experiment data for fuel assemblies similar to those for which the racks are designed. This benchmarking data is suf ficiently diverse to establish that the method bias and uncertainty will apply to rack conditions I which include strong neutron absorbers, large water gaps and low moderator densities.
I The design method which insures the criticality safety of fuel assemblies in the spent fuel storage rach uses the AMPX '
generation and KENO IV for reactivity system of codes for cross-section determination.
The 227 energy group cross-section library that is the common starting point Tor all cross-sections used for the benchmarks and the storage rack is generated I from ENDFiB-V' data. The NITAWL program includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial I we.ghting of cross-sections is performed by the XSDRN'M* program which is a one-dimensional Sa transport theory are then used as input to KENO IV
^
code. These multigioup cross-section sets which is a three dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 33 critical experiments has been analyzed using the above method to demonstrate its applicability to criticaljty analysis and to establish the method I bias and uncertainty. The experiments range from water moderated, oxide fuel arrays separated by various materials (84C, steel, water, etc) that simulate LWR (Light Water Reactor) fuel shipping and storage conditions to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials (Plexiglas and air) that demonstrate the wide range of applicability of the method. Table 2 on page 20 summarizes these experiments.
The average K.n of the benchmarks is 0.992 , The standard deviation of the bias value is 0.0008 Ak. The 95/95 one sided tolerance limit f actor for 33 values is 2.19. Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not great ar than 0.0018 ok.
l 2.2 IFBA CREDIT REACTIVITY EQUIVALENCING Storage of spent fuel assemblies with initial enrichments higher than that al-lowed by the methodology described in Section 2.1 is achievable by means of Analytical Methods I
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l the concept of reactivity equivalencing. Reactivity equivalencing is predicated f an upon the reactivity decrease associated with the addition of IFBA fuel rods and fuel depletion. A series of reactivity calculations is performed to generate a I set of IFBA-rod-number versus enrichment ordered pairs which all yield an equivalent K.o when the fuel is stored in the Farley Units 1 & 2 spent fuel racks.
The data points on the reactivity equivalence curve are generated with a trans-I port theory computer code, PHOENIX PHOENIX is a depletable, two-di.nensional, multigroup, discrete ordinates, transport theory code. A 25 energy I group nuclear date library based on a modified version of the British WIMS library is used with PHOENIX.
A study was done to examine fuel reactivity as a function of time following I discharge from the reactor.
CINDER Fission product decay was accounted for using CINDER is a point-depletion computer coc'e used to determine fission product activities. The fission products were permitted to decay for 30 years I af ter discharge. The fuel reactivity was found to reach a maximum at approxi-mately 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months af trr discharge. At this time, the major fission product poi-son, Xe , has nearly completely decayed away. Furthermore, the fuel reactivity was found to decrease continuously from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 30 years following dis-charge. Therefore, the most reactive time for a fuel assemoly af ter discharge from the reactor can be conservatively approximated by removing the Xe I The PHOENIX code has been validated by comparisons with experiments where the isotopic fuel composition has been examined following discharge from a reactor, in addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX, Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 3 on page 21.
The measurements were made on fuel discharged from Yankee Core 5". The data in Table 3 on page 21 shows that the agreement between PHOENIX pred-ictions and measured isotopic compositions is good.
The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in Table 4 on page 22. Key I parameters describing each of the 81 experiments are given in Table 5 on page
23. These reactivity comparisons again show good agreement between exper-iment and PHOENIX ca culations.
An uncertainty associated with the IFB A dependent reactivity computed with PHOENIX is accounted for in the development of the IFBA requirements as discussed in Section 3.2.1.
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, I Analytical Methods 4 l I
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I 3.0 SPENT FUEL RACK CRITICALITY ANALYSIS l I This section develops and describes the analytical techniques and models em-ployed to perform the criticality and IFBA reactivity equivalencing analyses for storage of 17x17. OFA and VANTAGE 5 Westinghouse fuel assemblics in the ,
Farley Units 1 & 2 spent fuel racks. )
Section 3.1 describes the KENO reactivity calculations which show that storage I of 17x17 OFA and VANTAGE 5 fuel assemblies in the spent fuel rack is ac-ceptable with norrinal enrichments up to 3.90 w/o U ". Section 3.2 describes the PHOENIX reactivity equivalencing analysis which establishes the IFBA re-I quirements for assemblies with nominal enrichments above 3.90 w/o and up to 5.00 w/o U Section 3.3 presents the results of the PHOENIX sensitivity cal-culations for enrichment, cell center-to-center spacing, and poison loading, and Section 3.4 discusses spent fuel rack postulated accidents.
l 3.1 KENO REACTIVITY CALCULATIONS The following assumptions are used to develop the nominal case KENO model for storage of fuel assemblies in the Farley Units 1 & 2 spent fuel rack.
1. Evaluation of the Westinghouse 17x17 fuel assemblies shows that the 17x17 I OFA assembly is the most reactive fuel type when all assemblies have the same enrichment. The 17x17 VANTAGE 5 fuel design parameters relevant to the criticality analysis are the same as the OFA parameters and will yield I equivalent results. Therefore, only the Westinghouse 17x17 OFA fuel as-sembly is analyzed (See Table 1 on page 19 for fuel parameters).
All fuel rods contain u anium dioxide at an enrichment of 3.90 w/o (nominal)
I 2.
and 3.95 w/o (worst case) over the entire length of each rod.
3. The fuel pellets are modeled at 96% of theoretical density without dishing or chamfering to bound the maximum fuel assembly loading.
4. No credit is taken for any U or U in the fuel, nor is any credit taken for the build up of fission product poison material.
5. No credit is taken for any spacer grids or spacer sleeves.
6. No credit is taken for any burnable absorber in the fuel rods or any natural enrichment axial blankets.
7. The moderator is pure water (no boron) at a temperature of 69'F. A con-servative value of 1.0 gm/cm is used for the density of water.
I Spent Fuel Rack Criticality Analysis 5
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8. The minimum poison material loading of 0.015 grams B ' per square centi-meter is used throughout the array.
9. The array is infinite in lateral (x and y) extent and finite in axial (vertical)
I. extent which allows neutron leakage from only the axial direction.
10. All available storage cells are loaded with fuel assemblies.
The KENO calculation for the nominal case resulted in a K.'s of 0.9101 with a 95 percent probabilityl95 percent confidence levet uncertainty of 10.0057. The nominal case result can be compared to the worst case result to determine the relative impact of applying worst case assumptions. The nominal case is also used as the center point for the sensitivity analysis discussed in Section 3 3. j I The maximum K.ee under normal conditions arises froin consideration of me-chanical and material thickness tolerances resulting from the manu%ct aring I process in addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells have shown that symmetrically placed f uel assemblies yield equal I or conservative results in rack K.ei. For development of a " worst case" model, sheet metal and Boraflex absorber tolerances are considered along with con-struction tolerances related to the cell 1.D. and cell center-to-center spacing.
I This results in a " worst case" model of the spent fuel rack with minimum center-to-center and cell-to-cell water gap (flux trap) spacing, as follows:
I Fuel Rack Dimension Nominal (inches)
Tolerance (inches)
Worst Case (inches)
Cell Center-to-Center 10.750 + /-0.060 10.690 Cell 1.D. 8.900 + /-0.045 8.945 Cell Wall Thk 0.135 + /-0.012 0.147 Cell Wrapper Gap 0.070 + /-0.00 5 0.075 Cell Wrapper Thk I Cell-to-Cell Gap 0.024 1.392
+ /-0.00 3 n/a' 0.027 1.247 Boraflex Thk 0.065 +/-0.010 0.055 Boraflex Width 8.000 + /-0.060 7.940 Boraflex Length I Boraflex Length 147.00 0%
+/-0.2 50 n/a#
146.75 3%
Shrinkage I ,
-. _..._.s_....._........._.._.-.c._...
3% Bo..el.s snr ink.g. chos.a .s .d di tion.1 conse. vat.sm Spent Fuel Rack Criticality Analysis I
6
I i I The Boraflex absorber material in the " worst case" model is assumed to shrink 3% in the axial direction. For conservatism, all shrinkage is taken from one end of the poison panel, thereby exposino 3.65 inches of active fuel. At the oppo-j j
site end, no credit is taken for poison material extending beyond the active fuel
~
I'i length. Additionally,- fuel enrichment is increased by 0.05 w/o U to conservatively account for enrichment variability. Thus, the " worst case" KENO I
model of the spent fuel storage racks contains minimum center-to-center and cell to-cell gap spacings, Boraflex shrinkage, and symmetrically placed f uel as- i m ,
semblies at a maximum enrichment of 3.95 wto U .
Based on the analysis described above, the following equation is used to de-velop the maximum K.o for the Farley Units 1 & 2 spent fuel racks:
K." = Kwo,si + Bm.moe + Bp.o + (( (k s)'.oru + (ks)ieroo ]
K.ori = worst case KENO Ken that includes material, mechanical and enrichment tolerances and Boraflex absorber shrinkage Sm.inoo = method bias determineo from benchmark critical comparisons B o.o = bias to account for Boraflex poi'.on particle self-shielding.
k s.or it = 95/95 uncertainty in the worst case KENO K."
k sm.inoo = 95i55 9ncertainty in the method bias Substituting calculated values in the orc.er listed above, the result is:
K." = 0.9312 + 0.0083 + 0.0022 + (((0.0052)# + (0.0018)# ] = 0.9472 Since K." is less than 0.95 including uncertainties at a 95!M probability / confidence level, the acceptance criteria for criticality is met with fuel enriched to a nominal 3.90 w/o U*.
I l 3.2 IFBA CREDIT REACTIVITY EQUIVALENCING Storage of fuel assemblies with nominal enrichments greater than 3.90 w/o U" in the Farley Units 1 & 2 spent fuel stora;;e racks is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equiv-alencing is predicated upon the reactivity decrease associated with the addition of IFBA fuel rods and fuel depletion.
I Two analytical techniques are used to establish the criticality criteria for the j storage of IFBA fuel in the fuel racks. Th6 first method uses reactivity equiv-l aiencing to establish the poison material loading required to meet the criticality limits. The poison material considered in this analysis is a zircorGum diboride (ZrB2) coating manuf actured by Westinghouse. The second method uses the fuel Spent Fuel Rack Criticality Analysis 7
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I assembly infinite multiplication f actor to establish a reference reactivity. The reference reactivity point is compared to the fuel assembly peak reactivity to determine its acceptability f or storage in the spent fuel racks, I
3.2.1 REACTIVITY EQUlVALENCING ANALYSIS A series of reactivity calculations are perf nrmed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the equivalent K.fi when I the fuel is stored in the spent fuel racks. The fuel burnup used in the reactivity calculation is that burnup which yields the highest equivslent Ken when the fuel
'i s stored in the spent fuel racks. Fuel assWy depletions performed in I PHOENIX show that for the number of IFBA rods per assembly considered in this analysis, tne maximum reactivity for rack geometry occurs at zero burnup.
Although the boron concentration in the IFBA rods decreases with fuel depletion.
I the fuel assembly reactivity decreases more rapidly, resulting in a maximum fuel rock reactivity at zero burnup.
The following assumptions were used for the IFBA rod assemblies in the
-I PHOENIX models:
1. Evaluation of the Westinghouse 17x17 fuel assemblies shows that the 17x17 OFA assembly is the most reactive fuel type when all assemblies have the same enrichment. The 17x17 VANTAGE 5 fuel design parameters relevant to the criticality analysis are the ssme as the OFA parameters and will yield equivalent results. Therefore, only the Westinghouse 17x17 OFA fuel as-sembly is analyzed (See Table 1 on page 19 for fuel parameters).
I 2. The fuel pellets are modeled at 9t,% of theoretical density without dishing or chamfering to bound the maximum fuel assembly loading.
3. No credit is taken for any spacer grids or spacer sleeves.
4. No credit is taken for any natural enrichment axial blankets.
5. The IFBA absorber material is a zirconium diboride (ZrB: ) coating on the fuel pellet. Each IFBA rod has a nominal poison material loading of 1.50 m
milligrams B per inch which is reduced by 5 percent to conservatively account for manuf acturing tolerances.
6. The B loading is additionally reduced by 25 percent in each IFBA rod to conservatively model a minimum poison length of 108 inches.
7. The moderator is pure water (no boron) at a temperature of 68*F. A con-servative value of 1.0 gm/cm' is used for the density of water.
8. The array is infinite in lateral (x and y) and axial (lateral) extent which pre-cludes any neutron leakage from the array.
Figure 4 on page 29 shows the constant K.o contour generated for the Farley Units 1 & 2 spent f uel racks. Note the endpoint at 0 IFBA rods where the Spent Fuel Rack Criticality Analysis 8
I I nominal enrichment is 3.00 wio and at 80 IFBA rods where the nominal enrichment is 5.00 w/o. The interpretation of the endpoint data is as follows:
the reactivity of the spent fuel racks containing fuel with 80 IFBA rods which l
I has an initial t'ominal enrichment of 5.00 w/o is equivalent to the reactivity of the spent fuel racks containing fresh fuel having an initial nominal enrichment of 3.90 w/o. The data in Figure 4 on page 29 is also provided on Table 6 on page 25.
It is important to recognite that the curve in Figure 4 on oage 29 is based on reactivity equivalence calculations for the specific enrichment and IFBA combi- l nations in actual fuel rack geometry (and not just on simple comparisons of individual fuel assembly infinite multiplication f actors). In this way, the envi-ronment of the storage rack and its influence on assembly reactivity is implicitly considered.
The IFBA requirements of Figure 4 on page 29 include the reactivity ef fects of I 'FBA rod repositioning within an assembly. The worth of individual IFBA rods can change depending on position within the assembly due to local variations in thermal f Nx. Studies were performed to evaluate the reactivity ef fects of I lFBA tod repositioning and a conservative reactivity margin was included in the development of the final IFBA requirements to account for this ef fect, in ad-dition, to conservatively account for calculational uncertainties, the IFB A re-I quirements of Figure 4 on page 29 also include a conservatism of approximately 10% on total IFBA rods at the 5.0 w/o end.
I Additional IFBA credit calculations were performed to examine the reactivity effects of higher IFBA linear B" loadings (2.25 and 3.00 mglin). These calcu-lations confirm that assembly reactivity remains constant provided the net B I material per assembly is preserved. Therefore, with higher IFBA B loadings, the required number of IFBA rods per assenihly can be reduced by the ratio of the higher loao, g to the nominal 1.50 mglin loading. For example, using 3.00 I mg-B'/in IFBA in 5.0 w/o fuel assenablies would allow a reduction in the IFBA rod requirement from 80 IFBA rJds per assembly to 40 IFBA rods per assembly (80 civided by the the ratio 3.00/1.50).
The equivalent K.et for the storage of fuel in the Farley Units 1 & 2 spent fun racks is determined using the methods described in Section 3.1 of this report.
The reference conditions for this are defined by the zero IFBA intercept point
-I in Figure 4 on page 29. The KENO-IV computer code was used to calculate the storage rack multiplication factor with an equivalent fresh fuel nominal enrichment of 3.90 w/o and no IFBAs. The KENO calculation for the nominal I case resulted in a K.ve of 0.9101 with a 95 percent probability /95 percent confi-dence level uncertainty of 10.0057.
The maximum K.fr under normal conditions was determined with a " worst case" KENO model which includad mechanical and material tolerances, symmetric po-sitioning of fuel assemblies within the storage cells, Boraflex shrinkage, and increased fuel assembly enrichment to account for enrichment variability. The maximum K ve for the Fuley Units 1 & 2 spent fuel storage racks was 0.9472 I
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I I including method biases and uncertainties at a 95/95 probability / confidence level, This analysis is discussed in detail in Section 3.1.
I 3.2.2 INFINITE MULTIPLICATION FACTOR I The infinite multiplication f actor, Km , is used as a reference criticality reactivity point, and of fers an alternative method for determining the acceptability of fuel assembly storage in the Farley Units 1 & 2 spent fuel racks. The reference Ka is determined for a nominal fresh 3.90 w/o fuel assembly.
The fuel assembly Km calculations are performed using the Westinghouse li-I censed core design code PHOENIX-P . The following assumptions were used to develop the infinite multiplication f actor model:
A Westinghouse 17x17 OFA fuel assembly was analyzed (see Table 1 on I
1.
page 19 f or fuel parameters). The fuel assembly is modeled at its most reactive point in life and no credit is taken for any burnable absorbers in the assembly.
2. All fuel rods contain uranium dioxide at an enrichment of 3.90 w/o U* (OFA or VANTAGE 5) over the infinite length of each rod (this is the maximum nominal enrichment that can be placed in the spent fuel racks without IFBA ro d s).
I 3. The fuel array model is based on a unit assembly configuration (infinite in the lateral and axial extent) in Farley Units 1 & 2 reactor geometry (no rack).
I 4 The moderator is pure water (no boron) at a temperatu e of 68' F.
conservative value of 1.0 gm/cm' is used for the density of water.
A Calculation of the infinite multiplication f actor resulted in a reference Km of I 1.455. This includes a 1% AK reactivity bias to conservatively account for calculational uncertainties. This bias is consistent with the standard conserva-I t:sm included in the Farley Units 1 & 2 core design refueling shutdown margin calculations. All OFA and VANTAGE 5 fuel assemblies placed in the Farley Units 1 & 2 spent fuel racks must comply with the enrichment versus number of IFBA I rods curve in Figure 4 on page 29 or have a reference reactivity less than or equal to the above value. The maximum spent fuel rack reactivity will then be less than 0.95, as shown in Section 3.1, and the acceptance criteria for criticality ch will be met for fuel that has reference Km less than or equal to 1.455.
l 3.3 SENSITIVITY ANALYSIS To show the dependence of K.o on fuel and storage cells parameters as re-quested by the NRC , the variation of the K." with respect to the following parameters was developed using the PHOENIX computer code:
I Lpant Fuel Rack Criticality Analysis I 10
I
_g 1. Fuel enrichment, with a 0.50 w/o U" delta about the nominal case B enrichment.
Center-to-center spacing of storage cells, with a half inch delta about the I 2.
nc minal case center-to-center spacing.
Poison loading, with a 0.01 gm-B /cm delta about the nominal case poison I
3.
loading.
Results of the sensitivity analysis are shown in Figure 5 on page 30 through Figure 7 on page 32.
3,4 POSTULATED ACCIDENTS I Most accident conditions will not result in an increase in K.o of the rack. Ex-amples are:
I Loss of cooling systems Reactivity decreases since loss of cooling causes an increase in temperature, which causes a decrease in water density, which results in decreased reactivity.
Dropping a fuel The rack structure pertinent for criticality is not assembly on top of rack excessively deformed and the dropped assembly I has more (nan ten inches of water separating it from the active fuel height of stored assemblies which precludes interaction, Dropping a fuel Design of spent fuel rack is such that it assembly between rack precludes the insertion of fuel assembly in other modules or between I
than prescribed locations.
rack module and well I However, two accidents can be postulated which would increase reactivity:
mistoading an assembly with an enrichment and IFBA combination outside of the acceptable area of Figure 4 on page 29. or dropping an assembly into an I already loaded cell, The mistoading enrichment-IFBA accident is not considered credible for the Farley Units 1 & 2 spent fuel rack since the same rack design and limits apply throughout the entire spent fuel pool (one-region spent fuel rack design). With a one-region design, the possibility of misloading a Region 1 assembly into a Region 2 area of the rack does not exist. Furthermere, the requirements of the spent fuel rack IFBA limit will become a design constraint on future Farley reload core designs, anct fuel vendor quality assurance controls I on design, manuf acturing, and shipment will eliminate the possibility of ever having a potentially violating fuel assembly delivered to the site.
For the accident of dropping of a fuel assembly into an already loaded cell the upward axial leakage of that cell will be reduced however the overall ef f ect on rack reactivity will be insignificant. This is because the total axial leakage Spent Fuel Rock Criticality Analysis 11
, - _ . - . - . - _ - ~ . - - - . - - . - ~ .- - _.- - - - - - -
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I in tioth the upward and downward airections f or the entire f pent fuel attay (over 1400 cells) is worth only 0.30 percent Ak. Thus, minimiting the upward-only leakage of just a single cell will not cause any significant increase in rack re-iI activity (much lebs than 0.15 percent Ak.). Furthermore, the neutronic coupling between the dropped assembly and the already loaded assembly will be very low due to the several inches of asss mbly nottle f,tructure which would separate the active f uel regions, I
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I l 4.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS I This section describes the analytical techniques and models employed to per-form the criticality analysis for storage of fresh fuel in the Farley Units 1 & 2 fresh fuel storage racks.
Sin;e the fresh fuel tacks are maintained in a dry condition, the criticality J analysis will show that the rack Kee, is less than 0.95 for the full water density )
condition and less than 0.98 for the low water density ioptimum modeiation) ,
cond.tions. The criticality methodology ernployed in this analysis is discussed l in Section 2 of this report.
I The following assumptions were used to develop the KENO model f or the stot-age of fresh fuct in the Farley Units 1 & 2 fresh fuel storage rack under full density and low density optimum moderation conditions:
l
1. The fuel assembly is modeled at its most reactive point in life, and no credit is taken for any burnable absorber in the fuci rods or any natural enrichment axial blankets.
2, All fuel rods contain uranium dioxide at a maximum enrichment of 4.85 w/o over the en ire length of each rod.
3. The fuel pellets are modeled at 96% of theoretical density without dishing or chamfering to bound the maximum fuel assembly loading.
4. No credit is taken for any U* or U* in the fuel, nor is any credit taken for the build up of fission product poison material.
5. No credit is taken for any spacer grids or spacer sleeves.
6. The fuel rack center-to-center spacing is nasumed to be 21 inches and all available storage cells are loaded with fuel assemblies.
7. For both the full density and optimum moderation casos, there is no boron in the water, g 4.1 FULL DENSITY MODERATION ANALYSIS in the KENO model for the full density moderation analysis, the moderator is I pure wates at a temperature of 68'F. A conservative value of 1.0 gm/cm' is used for the density of water. The fuel array is infinite in lateral (x and y) and axial (vertical) extent which precludes any neutron leakage from the array.
Evaluation of the Westinghouse 17x17 fuel assemblies shows that the 17x17 OFA assembly is the most reactive fuel type under full density moderation conditions when all assernblies have the same enrichment. The 17x17 VANTAGE 5 fuel Criticality Analysis of Fresh Fuel Racks 13 I i
I design parameters relevant to the criticality analysis are the same as the OF A parameters and will yield equivalent results. Therefore, only the Westinghouse 17x17 OFA fuel assembly is analyzod under f ull density moderator conditions (see Table 1 un page 10 f or fuel parameterst
} The maximum K." under normal conditions arises from consideration of me-i chenical and material thickness tolerances resulting from the manuf acturing I process in addition to asymmetric positioning of fuel assemblies within the l storage cells. Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield equal or conservative results in rack K.H.
] Due to the relatively large cell spacing, the small tolerances on the cell I.D.
and center-to center spacing are not considered since they will have an insig-j nificant ef fect on the fuel rack reactivity. However, the corner angle iron di-mensions (thick ness of 0.25 +/- 0.03 and width of 2.00 + /- 0.2 5 inches) are l reduced to their minimum tolerances to minimize neutron absorption in the rack j structure. Thus, the most conservative, or " worst case" KENO model of the fresh fuel storage racks contains minimum corner angle iron dimensions with i
symmetrically $4 aced fuel assemblies at 4.85 w/o U* (corresponds to 4.80 w/o nominal).
i
'g Bast.d o. the analysis described above, the f ollowing equation is used to de-3 velop the maximum K.o for the Farley Units 1 & 2 fresh fuel storage racks:
, K.n = K. .n + Bm moe + /((ksf.o,n + (ks)$.moa )
where:
K .., n s worst case KENO K." with full dencity water Om.moa a method bias determined from benchmark critical comparisons k s.. n = 95/95 uncertainty in the worst case KENO K.o k sm.mos = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K.o = 0.9191 + 0.0083 + (((0.0070)# + (0.0018)' )= 0346 I Since K." is less than 0.95 including uncertainties at a 95/95 probability confi-dence level, the acceptance criteria for criticality is met.
4.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS g
For the low density optimum moderation analysis, the fuel array is modeled as I an infinitely long double row of fuel assemblies. The " worst case" cell con-figuration from the full density analysis is used in modeling the actual fresh fuel rack a rr a y. Concrete walls and floor are modeled. Under low water density I conditions, the presence of concrete is conservative because neutron" are re-Criticality Analysis of Fresh Fuel Racks 14
I I flected blck into the fuel array more ef ficiently than they would be with just low densit/ water. The area above the fresh fuel rock is filled with water at the optimum moderation density.
The Westinghouse 17x17 STO f uel assembly was analvred under low density optimum moderation conditions. Previous Westinghouse studies have shown the 17x17 STD f uel assembly to be more reactive than the other 17x17 fuel as-sembly types under optimum moderation conditions. This is because the STD fuel assembly contains a higher uranium loading than the other assemblies, and when optimum moderation conditions are prebent, higher load;ngs rt ault in higher reactivity (see Table 1 on page 19 for fuel parameters).
rigure 8 on page 33 shows the calculated Farley Units 1 & 2 fresh fuel rack I reactivity data points hncluding biases and 95/95 uncertainties) as a f unction of water density. A bounding curve has been drawn through the limiting data points to encompass all the calculated reactivity points. Based on this c ut ve, the I mamimum rack K." under low density moderation conditions is shown to occur at approximately 0.095 gm/cm' water density. At this condition, the calculated K.o of the fresh fuel tack array at 0.095 gm!cm' water density is estimated to be 0.8030 (not including biases and uncertainties) with a 95 percent probability and 95 percent confidence level uncertainty of !0.0075.
I The following equation is used to develop the maximum K.o f or the Farley Units 1 & 2 fresh fuel storage racks under low water density optimum moderation Conditions:
K.o =Ke... + 8e.moe + ~ ((k s)'e. . + (ks)hemoa )
where:
I K u.. . = maximum KENO Ken with low density optimum moderation Bm. moo = method bias determined from benchmark critical comparisons k s e... = 95/95 uncertainty in the base case KENO K.o k s m.mo a = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:
K." = 0.8030 + 0.0083 + (((0.0075)' + (0.0018)' ] = 0.8190 Since K.o is less than 0.98 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met.
I 4.3 POSTULATED ACCIDENTS Under normal conditions, the f resh f uel racks are maintained in a dry environ-ment. The introduction of water into the f resh fuel rack area is the worst case accident scenario. The full density and low density optimum modera' ion cases
, c,i,,camy na,ys,s o, ,,esh ,uema s se
I lg are bounding accident situations which result in the most conservative fuel rack l 3 ts e n.
Other accidents can be postulated which would cause some reactivity increase
, (i.e., dropping a fuel assembly between the rack and wall or on top of the rack).
l For these other accident conditions, the double contingency principle is applied, This states that one is not requited to assume two unlikely, independent, con-l I current events to ensure protection against a criticality accident. Thus, f or these other accideht conditions, the absence of a moderator in the fresh fuel storage j
rack s can be assumed as a reahstic initial condition sjnce assuming its presence would be a second unlikely event.
Westinghouse generic studies have shown the maximum reactivity increase for l postulated accidents ($ gh .* those entioned above) will be less than 10 per-cent AL, Furthermore, tcc nort. dt, 3ry fresh f uel tack reactivity is less than 0.70.
As a ret, ult, for postulato mu.idents, the maximum rock K," will be less than
0.95.
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I Criticality Analysis of Fresh Fuel Racks 16
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I 5.0 APPLICATION OF SPENT FUEL RACK IFBA CREDIT LIMIT TO FRESH FUEL RACK )
I Section 3 of this report shows that the acceptance criteria f or criticality is satisfied for the storage of Westinghouse 17x17 OF A and VANTAGE 5 fuel as-semblics with nominal entschments up to 3.90 w/o in the spent fuel storage racks. To allow storage of fuel assembhes with higher enrichments, Figure 4 on page 29 was developed based on a reactivity equivalencing analysis. When fuel assemblies which satisfy the enrichment-IFBA combinations of Figure 4 on page 29 are stored in the spent fuel rock area, the spent fuel tack reactivity (K.n) will be less than the re activity of the spent fuel rack filled with 3.90 w/o, non-IFB A fuel assemblies. l For the fresh fuel storage racks, Section 4 of this report shows that the ac- 1 I ceptance criteria for criticality is satisfied for the storage of Westinghouse 17x17 STD. OFA, and VANTAGE 5 f uel assemblies with nominal enrichments up to 4.80 w/o.
Of the two storage rack limits (fresh and spent), the spent fuel rack limit is the more restrictive since fuel assemblies are limited to the equivalent reactivity of a nominally enriched 3.90 w/o fuel assembly. Therefore, any fuel assembly
_I which satisfies the requirements for storage in the spent fuel rack area will be less reactive than the nominal 4.80 w/o fuel assembly which can be safely stored in the fresh fuel rack area. Thus, it is conservative to use the spent fuel I rack enrichment lFBA limit for the fresh fuel storage rsck.
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Application of Spent Fuel Rack IFB A Credit Limit to Fresh Fuel Rack lI 17 1
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l 6,0
SUMMARY
OF CRITICALITY RESULTS The acceptance criteria f or criticality requires the ef fective neutron multiplication I factor, Kett, to be less than or equal to 0.95, including uncertainties, under all conditions for the storage of fuel assemblies in the spent fuel storage racks.
For the storage of fuel assemblies in the fresh fuel storage racks, the Ken must be less than or equal to 0.95, including uncertainties, under flooded conditions, i
.and less than or equal to 0.98, including uncertainties, under optimum moderation I conditions.
This report shows that the acceptance criteria for criticality is met f or the Farley Units 1 & 2 spent fuel storage racks for the storage of Westinghouse 17x17 I OFA and VANTAGE 5 fuel assemblies, and for the fresh fuel storage racks for the storage of Westinghouse 17x17 STD, OFA and VANTAGE 5 fuel assemblies, with the f ollowing enrichment limits:
Spent Fuel Rack s 5.00 w/o U" (nominal), with IFB A required f or assembly enrichments greater than 3.90 wio l I (nominal) as shown in Figure 4 on page 29. The previous licensing basis for standard fuel up to 4.30 w/o (maximum) in the spent fuel racks is not impacted by this report.
Fresh Fuel Rack s 5.00 w/o U* (nominal), with IFB A required for assembly enrichments greater than 4.80 w/o (nominal). The use of Figure 4 on page 29 for establishing IFBA requirements for enrichments above 4.80 w/o is conservative.
I The analytical methods employed herein conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants,'
Section 5.7, Fuel Handling System; ANSI 57.2-1983, " Design Objectives f or LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety";
NRC Standard Review Plan, Section 9.1.2, " Spent Fuel St orage"; and ANSI 57.3-1983, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."
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Summary of Criticality Results 18
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I Table 1. Fuel Parameters Employed in Criticality Analysis 0
Parameter W 17x17 OFA W 17x17 STD l
& VANTAGE 5 l l
Number of fuel Rods I per Assembly 26k 264 Rod Zirc-4 Clad 0.0. (i nc h) 0 360 0 374 l Clad Thickness (inch) 0.0225 0.0225 I Fuel Pe l le t 0.D. (inch) 0 3088 0 3225 l
Fuel Pellet Density 1
(% of Theoretical) 96 96 !
Fuel Pellet Dishing Factor 0.0 0.0 Rod Pitch (i nc h) 0.496 0.496
Number of Zirc-4 Guide Tubes 24 24 Guide Tube 0.D. (inch) 0.474 0.482 Guide Tube Thickness (inch) 0.016 0.016 Number of i n s t rerr.en t Tubes 1 I I I instrument Tube 0.0. (inch) 0.474 0.482 instrument Tube Thickness (i nc h) 0.016 0.016 I
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I Surnmary of Criticality Results 19
I I Table 2. Benchmark Critical Experiments (5.6)
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Deneral I
Enetchment Seoarating Soluble Desertption w/o U235
...................................................................ppe Reflector Material Boron koff l
1. 002 rod lattico 2.46 water water
2. U02 rorJ 1 s t t lee 2.46 water water O O.9857 +/- .0028 l
3. U02 rod lattice 2.46 water water 1037 764 0.9906 */* .0018 4 U02 rod 1sttice 2.46 water 84C pins 0.9896 */* .0019 S. U02 rod lattice 2.46 watee 54C pins O O.9914 */* .0025
6. U02 rod lattice 2.46 water 84C pins O O.9891 +/. .0026
7. U02 rod lattice 2.44 O O.9955 */* .0020 I
water 84C pine O O.9449 +/* .0027
8. U02 rod lattico 2.46 water 84C pins
9. U02 rod lattice 2.46 water water O O.fe983 */* .0025
10. UO2 rod lattice 2.46 water water O O.8931 +/* .0028
11. U02 rod lattice 2.44 water 143 0.9928 +/* .0025
12. U02 rod lattico 2.46 water stainless steel Sid 0.9967 */* .0020 l I 0.9943 +/* .0019 stainless steel 217
13. U02 rod lattice 2.44 water borated aluminum 15 0.9892 +/* .0023 i
14 U02 rod lettIce 2.46 water
19. U02 rod lattice 2.46 water borated aluminum 92 395 0.9484 +/* .0023
14. 002 rod lattico 2.46 water borated aluminum 0.0032 +/* .0021
17. U02 rod lattlee 2.46 water borated aluminum 121 0.9848 +/* .0024 borated aluminum 407 0.9895 +/* .0020 I 18. 002 rod lattice
19. U02 rod lattice
20. UO2 rod lettico
21. UO2 rod lattice 2.46 2.45 2.46 2.46 water water water water berated borated borated borated aluminum aluminum aluminum aluminum 197 634 320 72 0.9805 +/* .0022 0.9921 +/- .0019 0.9920 */* .0020 0.9939 +/* .0020
22. U metal cylinders 93.2 bare ate O O.9905 +/- .0020
23. U metal cylthders 93.2 bare ate O O.9976 +/- .0020 24 U metal cyltnders 93.2 bare
29. U metal cylinders 93.2 bare ate O O.9947 +/- .0025
26. U metal cylinders 93.2 bare air O O.9928 +/- .0019
27. U metal cyltnoere 93.2 bare air O O.9922 +/- .0026 afr O O.9950 +/* .0027 I 20. U metal
29. U mets)
30. U metal cyttnders cyllnoers cylinders 31 U metal cyltnoers 93.2 93.2 93.2 93.2 bare paraffin bare paraffin pleutglass pleutglass plextglass pleutglass O
O O
O O.9941 */* .0030 O.9928 +/- .0041 O.9968 +/* .0018 t 0042 */* .0019
32. U metal cylinders 93.2- paraffin pleutglass O O.9963 +/- .0030
33. U metal cylinders 93.2 parafftn pleutglass O O.9919 +/- .0032 I
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I Table 3. Comparison of PHOENIX lsotoples Predictions to Yankee Core 5 Measurements Quantity (Atom Ratio) '6 Difference U235/U -0.67 U2361U -0.26 ,
U238/U -0.03 4
i Pu239/U Pu240/U
+ 3.2 7
+ 3.63 Pu241/U -7.01 Pu242/U -0.20 Pu239/U238 + 3.24 Mass (Pu/U) + 1.41 FISS-Pu/ TOT-Pu -0.02 I
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I Table 4.
1 Benchmark Critical Experiments PHOENIX Comparison
= Description of Number of PHOENIX k." Using Experiment Experiments Experiments Bucklings Al clad 14 0.9947 SS clad 19 0.9944 80f ated Hr0 7 0.9940 Subtotal 40 0.9044 U-Metal Al clad 41 1.0012 TOTAL 81 0.9978 I
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T8bl9 5. D8t8 for U MetSI and UO Crit lCSI Experiments (Part 1 of 2)
I I fuel Pellet C16d Clod Case Colt A/O H20/U Oenesty Otometer Materiet 00 Lattice Number Type U 235 Satto (0/CC) (CM) Cleo Thicknese Ptten Boron (CMI (CM) (CM) PPM t Hema 1.328 3.02 7.83 I 2 3
4 6
Hess Home Hema Hera 1.328 1.328 1.328 1.328 3.95 4.95 3.92 4.89 7.53 7.83 7.S2 7.52 1.5265 1.S265 1.5269
.98S5
.9855 Aluminum 1.6816 Aluminum 1.6916 Aluminum 1.6916 Aluminum t.1506
.07910
.07110
.07110
.07110 2.20S0 2.3590 2.5120 1.5580 0.0 0.0 0.0 0.0 6 Hess 1.328 2.88 10.83 Aluminum t.1506 .07110 1.6S20 I .9728 Aluminum 1.1506 0.0 7 Here 4.328 3.58 10.53 .9728 .07110 1.5580 0.0 8 Hems t.328 4 83 10.53 Aluminum t.1506 .07110 1.6820 0.0 9 .9728 Afuntnum 1.1806 .07110 1.8060 Souare 2.734 2.18 10.18 .7620 55 304 .8594 0.0 to Square 2.734 2.92 10.18 04085 1.0287 0.0
.7620 55*304 .8594 0408S 1.1049 ft Savaro 2.734 3.86 10.18 .7620 55-304 .8594 0.0 12 Square 2.734 7.02 10.18 .04085 1.1938 0.0
.7620 55 304 .8904 04089 1 13 14 Square 2.734 Souare 2.734 8.49 10.38 10.18 10.18
.7620
.7620 55*304 55*304
.8694
.8594 04085 1.4594 1.5621 0.0 0.0 15 Souare 2.734 10.18 .04085 1.6091 0.0 2.50 .7620 55*304 .5594 .0408S 1.0617 16 luusre 2.734 4.99 10.18 .7620 0.0 I 17 18 19 20 Souare 3.746 Sauere 3.745 Square 3.749 Square 3.749 2.60 4.81 4.St 4.81 10.27 10.37 10.37 10.37
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.04060 1.2922 1.2522 1,2522 1.2522 709,0 1260.0 1334.0 1477.0 Square 4.069 2.56 9.46 1.1278 55*304 1. Sit 3 0.0 1.2090 .04040 I 27 28 29 30 31 Sque e Square Square 4.069 2.490 3.037 Ecuaro 3.037 Square 4.069 2.14 2.84 2.64 8.16 9.46 10.24 9.28 9.28 1.1274 1.0297 1.1268 1.1268 55*304 55*304 55*304 1.2090 .04060 Aluminum 1.2060 .08130 1.1701 1.2701
.07163
.07153 1.8113 1.4500 1.8113 1.5860 2.1980 3392.0 0.0 0.0 0.0 2.59 9.45 1.1268 55*304 1.2701 0.0 I 32 Square 4.069 3.93 9,45 .07163 f.9550 0.0 33 Square 4.069 1.1268 55 304 1.2701 .07163 1.6840 8,02 9.49 1.1268 55*304 0.0 34 Square 4.069 1.2701 .07163 2,1980 35 Square 2.490 9.90 9.45 1.1268 55-304 1.2701 .07163 2.3810 0.0 36 2.64 10.24 1.0297 Aluminum 1.2040 .98t30 0.0 Hema 2.096 2.06 10.38 1.5240 1.8113 1677.0 37 Aluminum 1.6916 .07112 2.1737 I 2.096 Hema 1.09 10.38 1.5240 0.0 38 Here 2.096 4.12 10.38 Aluminum t.6916 .07112 2.40S2 0.0 39 Hema 2,096 1.5240 Aluminum t.6916 .07112 2.6162 6.14 10.38 1.5240 0.0 40 Hema 2.096 8.20 10.38 1.5240 Aluminum 1.6916 .07112 2.9891 0.0 el Hems 1.307 Aluminum 1.6916 .07112 3.32SS 42 1.01 18.90 1.5240 aluminum 1.6916 .07112 0-0.
Hema 1.307 1.81 18.90 1.5240 2.1742 0.0 43 Hema 1.307 2.02 18.90 Aluminum 1.6816 .07112 2.4054 0.0 1.5240 Aluminum 1.6916 .07112 2.6162 0,0 I
I 23 I
Table 5. Data for U Metal and UO: CritlCal Experiments (Part 2 Of 2)
I I
Fuel Pellet clea clad tattice Case Call A/O H20/0 Donetty Diameter Mateetal 0D T hick ne s e PitCN . Boron Numeer Type U 236 Batto (0/CC) (CM) Clea (CM) (CN) (CM) PPM 44 Hexa 1.307 3,01 18.90 1.5240 Aluminum t.6916 .07192 2.9896 0.0 I 49 46 47 48 de te m a Hews Hess Hess Haus t.307 1.160 1.160 1.160 1.160 4.02 1.01 1.51 2.02 3.01 18.90 18.90 18.90 18.90 18.D0 1.9240 1.9240 1.5240 1.8240 1.6240 Aluminum 1.6916 .07t12 Aluminum 1.6916 .07112 Aluminum 1.6916 .07192 Aluminum 1.6916 Aluminum 1.6916 .07112 07112 3.3249 2.1742 2.4054 2.6162 0.0 0.0 0.0 0.0 2.9896 0.0 I 4.02 1.9240 SO Here 1.160 18.90 Aluminum 1.6916 07112 3.3249 0.0 51 Hema 1.040 1.01 18.90 1.9240 aluminum 1.6918 .07112 2.1742 0.0 S2 Hess 1.040 1.St 18.90 f.6240 Aluminum 1.6918 .07112 2.4064 0.0 l
53 Heus 1.040 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 64 Hess 1.040 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 1.040 18.90 I
SS Hema 4.02 1.6240 Aluminum 1.6916 .07012 3.3249 0.0 SS Hema 1.307 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 57 Hema 1.307 1.52 18.90 .9830 1.1506 .07112 l Aluminum 1.5926 0.0 i 98 Hosa 1.307 2.02 18.90 .9830 Aluminum 1.1506 07112 t.7247 0.0 )
99 Hema 1.307 3.02 18.D0 .9830 Aluminum .1.1606 .07112 1.9609 0.0 60 1.307 18.90 .9830 I
Heum 4.02 aluminum 1.1906 .07112 2.1742 0.0 61 Heus 1.160 1.52 18.90 .9830 aluminum 1.1506 .07112 1.5926 0;0 62 Hema 1.160 2.02 18.90 .9830 Aluminum 1.1606 .07112 1.7247 0.0 63 Heus 1.160 3.02 18.90 .9830 Aluminum 1.1906 .07112 1.9609 0.0 64 Hema 1.160 4.02 18.90 .9830 Aluminum 1.1606 .07112 2.1742 0.0 65 Home t.160 1.00 18.90 .9830 I
Aluminum 1.1606 .07tt2 t,4412 0.0 46 Hess 1.160 1.52 10.90 .9830 aluminum 1.1906 .07112 1.5926 0.0 67 Hema 1.160 2.02 18.90 .9830 Aluminum 1,1506 .'7112 0 1.7247 0.0 48 Heus 1.160 3.02 18.D0 .9830 Aluminum 1.1806 .07112 1.9609 00 SS Hema 1.150 4.02 18.90 .9830 Alumirsus 1.1506 .07112 2.1742 0.0 70 Hess 1.040 1.33 18.90 19.0S0 Aluminum 2.0$74 .07620 2.8687 0.0 I 71 72 73 74 Heus Hema Hess Hess 1.040 1.040 1.040 1.040 1.58 18.90 1.83 2.33 2.83 18.90 18.bo 18.90 19.050 19.0S0 19.050 10.050 aluminum 2.0574 Aluminum 2.0974 Aluminum 2.0574 Aluminum 2.0574
.07620
.07620 07620
.07620 3.0006 3.142$
3.3942 3.6284 0.0 0.0 0.0 0.0 75 Hess 1.040 3.83 18.90 19.050 Aluminum 2.0574 .07620 4.0S66 0.0 I 76 77 78 73 Hesa Hess Hema Heus 1.310 1.310 1.159 1.159 2.02 3.01 2.02 3.01 18.88 18.88 18.88 18.88 1.8240 1.5240 1.5240 f.5240 Aluminum 1.6916 Aluminum 1.6918 Aluminum t.6916 Aluminum 1.6916
.07112
.07112
.07112
.07412 2.6160 2.9900 2.6160 2.9900 0.0 0.0 0.0 00 to Hema 1.312 2.03 18.84 .9830 a l um i num 1.1606 .07112 1.7200 0.0 I St Hema 1.312 3.02 15.88 .9830 A l us t rLas 1.1506 .07112 1.9610 0,0 I
I I
24
Table 6. Farley Units 1 & 2 Spent Fuel Storage Minimum IFBA Requirements initial U* IFBA Rods I Enrichment in Assembly g 39 0 4.0 8 j k.2 22 44 37 46 51 4.8 66 5.0 80 I
I Note: The IFBA rod requirements shown in this table are based on a- IFB A linear BIO loading of 1.50 mg-B10/ inch. For higher IFBA linear I B10 loadings, the required number of IFBA rods per assembly can be reduced by the ratio of the increased 810 loading to the nominal 1.50 mg-B10/ inch loading.
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Note: The IFBA rod requirements shown in this figure are based on an IFBA linear B10 loading of 1.50 mg-B10/in. For higher IFBA linear B10 loadings, the required number of IFBA rods per assembly can be reduced by the ratio of the increased B10 loading to the nominal 1.50 mg-B10/in loading.
Figure 4 Farley Units 1 & 2 Spent Fuel Storage Minimum IFBA Requirements 29 I .
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Figure 6 Sensitivity of K.n to Center-to-Conter Spacing in the Farley Units 1
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Figure 8 Sensitivity of K.ee to Water Density in the Farley Units 1 & 2 Fresh Fuel Racks 33 I
j I l BIBLIOGRAPHY
1. Nuclear Regulatory Commission. Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent fuel Storage and Handling Applications,. Apr\\ 14, 1978.
lI
2. W. E. Ford ill, CSRL-V: Processed ENDf!B- V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNL/CSOlTM-160, June 1982.
3. N. M. Greene, AMPX: A Modular Code System for Generating Coupled Multigroup Neutron Gamma Libraries from ENDFIB, ORNLIT M-3706, March 1976.
4. L. M. Petrie and N. F. Cross, KENO IV-- An Improved Monte Carlo Criticality Program, ORNL-4938, November 1975.
5. M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor fue/, B AW-1484-7, July 1979.
6. J. T. Thomas, Critical Three Dimensional Arrays of Ul93.2) Metal Cylinders, Nuclear Science and Engineering, Volume 52, pages 350-359,1973.
7. D. E. Mueller, W. A. Boyd, and M. W. Fecteau (Westinghouse NFD), Qualification of KENO Calculations with ENDflB V Cross Sections, American Nuclear Society Transactions, Volume 56, pages 321-323. June I 1988.
8. A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors, WCAP-10106, June 1982.
9. Askew, J. R., Fayers, F. J., and Kemshell, P. B., A General Description of the I lattice Code WIMS, Journal of British Nuclear Energy Society, 564-584, 1966.
5, pp.
10. England. T. R., CINDER - A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.
11. Melehan, J. B., Yankee Core Evaluation Program Final Report, WC AP-3017-6094, January 1971.
12. Nquyen, T. O., et al., Qualification of the PHOENIX-PIANC Nuclear Design I Sysrem for Pressurized Water Reector Cores, WC AP-11597-A, November 1987.
I I Bibliography 34 I - _
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Attachment 3 Significant llazards Evaluation Pursuant to 10CTR50.92 ;
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JOSEPH M. FARLEY UNITS 1 & 2 CRITICALITY ANALYSIS FOR NEW AND SPENT FUEL RACKS SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS PROPOSED CHANGE l
It is proposed to update Technical Specification 5.3 (Reactor Core) and Technical Specification 5.6 (Fuel Storage) to increase maximum allowable enrichments to 5.05 w/o maximum enrichment for Optimized fuel Assemblies i (OFA) and for VANTAGE 5 fuel assemblies taking credit for the presence of Integral Fuel Burnable Absorber (IFBA). The current licensing basis of 4.3 w/o maximum enrichment for Low Parasitic (LOPAR) fuel remains i unchanged.
BACKGROUND Currently, Joseph H. Farley Nuclear Plant Units 1 and 2 use Westinghouse 17x17 Low Parasitic (LOPAR) fuel. As part of a long term fuel management strategy for Farley Nuclear Plant Units 1 and 2, Alabama Power Company plans to use the Westinghouse VANTAGE-5 fuel design in both units starting with Unit 2 Cycle 9 and Unit 1 Cycle 12. In support of this effort, plant-specific evaluations and analyses have been performed to qualify the new fuel and spent fuel storage racks for storage of VANTAGE-5 fuel. These evaluations and analyses also qualify the use of 0FA fuel in the new and spent fuel racks for possible future use of 0FA fuel. These evaluationt and analyses include OFA and VANTAGE-5 fuel enrichments of up to 5.05 w/o (includes a 0.05 w/o manufacturing uncertainty) in order to provide the maximum flexibility in core designs. Support for implementing the reanalysis is based on a plant-specific criticality analysis of the spent fuel racks taking credit for the presence of Integral fuel Burnable Absorber (IFBA) in 0FA or VANTAGE-5 fuel. Thus, Alabama Power Company proposes that the Farley Nuclear Plant Units 1 and 2 Technical
, Specifications be amended by revising the design fcatures requirements to allow for up to 5.05 w/o (includes 0.05 w/o manufacturing uncertainty)
U-235 0FA and VANTAGE-5 fuel in the reactor core and spent fuel and new fuel pit storage racks. The enrichment limit for LOPAR fuel will remain unchanged at 4.3 w/o with no requirement for IFBA.
A criticality analysis of the spent fuel racks at the Joseph H. Farley Nuclear Plant Units 1 & 2 was performed taking credit for the presence of Integral Fuel Burnable Absorber (IFBA) in 0FA and VANTAGE-5 fuel. The analysis of the spent fuel racks shows that Westinghouse 17x17 0FA and i 17x17 VANTAGE-5 fuel assemblies with enrichments of up to 3.95 w/o can be safely stored in the Joseph H. Farley spent fuel racks utilizing all locations. OFA and VANTAGE-5 fuel with enrichments above 3.95 w/o and up to 5.05 w/o can also be stored in the spent fuel racks provided credit for IFBA is taken. The fuel assembly IFBAs consist of neutron absorbing Page 1 of 9
material applied as a thin ZrD2 coating on the outside of the U02 fuel pellet. As a result, the neutron absorbing material is a non removable or integral part of the fuel assembly once it is manufactured. The current enrichment limit of 4.3 w/o remains applicable to LOPAR fuel. No IFBAs are required to support the LOPAR enrichment limit. It has also been demonstrated that Westinghouse 17x17 OfA and 17x17 VANTAGE-5 fuel assemblies with enrichments of up to 4.85 w/o can be safely stored in the Joseph M. f arley new fuel racks utilizing all locations. Storage of OfA ;
and VANTAGE 5 fuel assemblies in the new fuel racks with enrichments above '
4.85 w/o and up to 5.05 w/o U 235 is shown to be acceptable by taking credit for the same IFBAs which are required to satisfy the spent fuel rack limit. The LOPAR enrichment limit of 4.3 w/o will remain unchanged with no requirement for IFBA. As a result of the above enrichment change, the Joseph M. Farley Units 1 & 2 Technical Specifications Section 5.3, Reactor ;
Core (Fuel Assemblies) and Section 5.6, fuel Storage (Criticality) are '
proposed to br updated.
All enrichment values discussed in this document include a 0.05 w/o manufacturing uncertainty. Nominal enrichments, when increased by this uncertainty, yield maximum enrichment limits.
The Joseph M. Farley Units 1 & 2 spent fuel rack criticality analysis is based on maintaining Keff less than or equal to 0.95 for storage of fuel assemblies. The new fuel rack analysis is based on maintaining Keff )
less than or equal to 0.95 for storage of fuel assemblies under full water l density conditions and less then or equal to 0.98 under low water density l (optimum moderation) conditions.
The new and s)ent fuel rack criticality analyses form the analysis basis for the enric1 ment limit in the " Reactor Core" section (Section 5.3.1) of the Technical Specifications. Enrichment is not a key safety parameter with respect to core operation. For core operation, reload designs (including fuel enrichments) are evaluated to confirm that the cycle core design adheres to the safety limits that exist in the current accident i analysis and plant Technical Specifications. This is the only impact of enrichment on core operation. Thus, it is proposed that this section is changed to reflect the new rack criticality analyses.
As required by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that a proposed license amendment to implement the criticality analysis for new and spent fuel racks and to modify the reactor core enrichment limit for Joseph M. Farley Nuclear Plant, represents a no significant hazards consideration, in accordance with the three-factor test of 10 CFR 50.92(c), implementation of the proposed license amendment was analyzed using the following standards and found not to 1) involve a significant increase in the probability or consequences for an accident previously evaluated, 2) create the possibility of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of safety.
Page 2 of 9
I ANALs!S i
Criticality of fuel assemblies in a fuel storage rack is prevented by the ,
design of the rack which limits fuel assembly interaction. This is done by l establishing the minimum separation between fuel assemblies and, for the l spent fuel racks, inserting neutron poison between fuel assemblies, l The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a P5 percent confidence level ttat the effective neutron rditiplication factor, Keff, of the fuel assembly array will be less than 0.95 as recommended by ANSI 57.2 1983, ANSI 57.3 1983 and the NRC letter to power reactor licensees,
" Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978. The 0.95 Keff limit applies _to both the new and spent fuel racks under all conditions, except for the new fuel rack under low water density-(optimum moderation) conditions, where the Keff limit is 0.98 as recommended by NUREG 0800.
-Criticality Calculation Methodolooy l The criticality calculation method and cross section values are verified by comparison with critical experiment data for fuel assemblies similar to those for whicn the racks are designed. This benchmarking data is i sufficiently ofverse to establish that the method bias and uncertainty will apply to rack conditlons which include strong neutron absorbers, large water gaps, and low moderator densities.
IFBA Credit Reactivity Eauivalencino !
A base case analysis was performed tq determine the maximum 0FA and VANTAGE-5 fuel enrichment that may be stored while meeting the requirements above. Storage of s)ent OFA and VANTAGE-5 fuel assemblies with initial enrichments higher tian that obtained in the base case is achievable by means of the concept of reactivity equivalencing. Reactivity equivalencing ,
is predicated upon the reactivity decrease associated with the addition of IFBA fuel rods and fuel depletion. A series of reactivity calculations were performed to generate _a set of IFBA-rod-number versus enrichment i ordered pairs which all yield an equivalent Keff when the 0FA or VANTAGE-5 fuel is stored in the Joseph ~ M. Farley Units l'& 2 spent fuel racks.
Criticality Analysis of Spent Fuel Rack Storage of 17x17 0FA and VANTAGE-5 fuel assemblies in the spent fuel racx is acceptable with enrichments of up to 3.95 w/o U 235, and storage of 17x17 0FA and VANTAGE-5 IFBA assemblies is acceptable with enrichments above 3.95 w/o and up to 5.05 w/o U-235. The analyr' included sensitivity -
calculations for enrichment, cell center-to center spacing, and poison '
loading. The current enrichment limit of 4.3 w/o remains applicable for.17x17 LOPAR fuel with no requirement for IFBA.
The following accident conditions will not result in an increase in Keff of the rack: loss of cooling systems (reactivity decreases with decreasing ,
water density), dropping a fuel assembly on top of the rack (the rack i Page 3 of 9 ,
J
structure pertinent for criticality is not excessively deformed and the dropped assembly has more than ten inches of Hater separating it from the active fuel height of stored assemblies which precludes interaction), and dropping a fuel assembly between rack modules or between a rack module and wall (the design of the spent fuel rack is such that it precludes the insertion of a fuel assembly into other than prescribed locaticns).
However, accidents can be postulated which would increase reactivity (i.e.,
misplacing an assembly in an unqualified position in the spent fuel rack, misloading an assembly with an enrichment and IFBA combination outside of the acceptable limits, or dropping a fuel assembly into an aircady loaded cell). Hisplacing an assembly in the spent fuel rack is not considered credible for the Joseph H. Farley 1 & 2 ssent fuel racks, since the same rack design and limits apply throughout t ie entire spent fuel pool. The requirements of the spent fuel rack IFBA limit for OfA and VANTAGE-5 fuel will become a design constraint on future Joseph H. Farley reload core designs, and multi-layered fuel vendor cuality assurance controls on design, manufacturing, and shipment provide assurances that a potentially violating fuel assembly will not be delivered to the site, for the accident of dropping of a fuel assembly into an already loaded cell, the upward axial leakage of that cell will be reduced; however, the overall effect of rack reactivity will be insignificant.
Criticality Analysis of New fuel Racks Since the new fuel racks are maintained in a dry condition, the criticality analysis shows that the rack Keff is less than 0.95 for the full water density condition and less than 0.98 for the low water density (optimum moderation) conditions.
Under normal conditions, the new fuel racks are maintained in a dry environment. 1he introduction of water into the new fuel rack area is the worst case accident scenario. The full density and low density optimum moderation cases are bounding accident situations which result in the most conservative fuel rack Keff, for both cases, Keff remains within the acceptance limits.
Other accidents can be postulated which would increase reactivity (i.e.,
dropping a fuel assembly between the rack and wall or on top of the rack).
For these accident conditions, the double contingency principle is applied.
This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.
Thus, for these accident conditions, the absence of a moderator in the new fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.
Westinghouse generic studies have shown the maximum reactivity increase for postulated accidents (such as those mentioned above) would be less than 10 percent delta-k. Furthermore, the normal, dry, new fuel rack reactivity is less than 0.70. As a result, for postulated accidents, the maximum rack Keff will be less than 0.95.
Page 4 of 9
Anolication of Sp.ent fuel Rack IFBA Credit limik%_litw fuel Rack The acceptance criteria for criticality is sati' %' for the storage of Westinghouse 17x17 0FA and VANTAGE-5 fuel assem 4 . with enrichments of up to 3.95 w/o in the spent fuel 'torage racks. TL 4 -
aw storage of 0FA and VANTAGE-5 fuel assemblies with higher enrichments, minimum IFBA requirements were developed based on a reactivity equivalencing analysis.
For the new fuel storage racks, the acceptance criteria for criticality is satisfied for the storage of Westinghouse 17x17 0FA and VANTAGE-5 fuel assemblies with enrichments up to 4.85 w/o. This limit can be increased to enrichments of 5.05 w/o for 0FA ano VANTAGE-5 fuel in the new fuel racks by taking credit for the same IFBA that is present and taken credit for in the spent fuel racks.
Of the two storage rack limits (new and spent), the spent fuel rack limit is the more restrict've .ince OFA and VANTAGE-5 fuel assemblies are limited to the equivalent reactivity of an 0FA or VANTAGE-5 3.95 w/o fuel assembly.
Hence, it is conservative to use the spent fuel rack enrichment-IFBA limit for the new fuel storage rack.
Other Considerations The storage of 0FA or VANTAGE-5 fuel in the new and spent fuel racks will not adversely affect the seismic response of the racks. 17x17 LOPAR, 0FA, and VANTAGE-5 fuel assemblies are structurally equivalent for the fuel storage rack seismic analyses.
Heat load calculations performed for 0FA and VANTAGE-5 fuel show that the spent fuel pool decay heat load assumed in the current analysis remains bounding for 0FA and VANTAGE-5 fuel. The clad temperature will increase by at most 10F due to increased heat flux with the smaller rod diameter. This small increase will not challenge fuel integrity.
Storage of 0FA and VANTAGE-5 fuel in the new and spent fuel racks does not involve any changes that would affect the releases of radiolcgical effluent during normal operation nor the radiological consequences of postulated accidents reported in the FSAR. The VANTAGE-5 fuel design doas include features to accommodate extended fuel burnup; however, extended fuel burnup has previously been incorporated at Farley Nuclear Plant Units 1 & 2.
Extended fuel burnup was evaluated for both units, and fuel containing extended burnup features was incorporated into operation in Cycle 10 of Unit 1 and Cycle 7 of Unit 2.
Implementation of extended fuel burnup on Farley Units 1 & 2 considered the potential impact on accident source terms. The topical report, " Extended Burnup Evaluation of Westinghouse fuel," (WCAP-10125-P-A) was used as the evaluation basis. This report, which was reviewed and approved by the NRC, demonstrated that there is very little effect on the source terms and that radiological consequences of accidents are not significantly affected. The NRC, in their SER regarding this report, added discussion specifically in regard to the fuel handling accident.
Page 5 of 9 l
w
There will be no significant increase in dose rate as a result of the use of VANTAGE-5 or OfA fuel of up to 5.05 w/o enrichment. Evaluations have shown that no change to existing radiation controls is required.
Conformance of the proposed amendments to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three-factor test) is shown in the following:
A. Increased fuel Enrichment for Reactor Core
1) Operation of Joseph M. farley Units 1 & 2 in accordance with the proposed license amendment does not involve a significtst '.ncrease in the probability or consequences of an accident pr viausly evaluated because the applicable safety limits are entnia ti a bounds previously established. Neither actuation of tafety synems nor accident mitigating capabilities are adversely af ferted by operation of the plant in accordance with the p oposed license amendment. The analysis demonstrates that the procoser amendment does not pose a challenge to installed safety eystem . T h.e re f ore ,
no new performance requirements are being impor d on any syst-m or component important to safety such that any desfg cru a ja will be exceeded. The implementation of the criticality mealysis is not an initiator for any of the postulated FSAR acciderte analyzed.
This analysis does not impact accident analyse; y plht accident scenarios. The analysis does not impact the accuni ., as analyzed in the FSAR. All accident acceptance criteria continue to be met.
The analysis demonstrates that the proposed amendmen .eets the acceptance criteria for criticality for speat fuel storage racks and new fuel storage racks. Operation of the plant in accordance with the proposed license amendment will not impact accident analyses or plant accident scenarios as analyzed in the FSAR.
2) The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated because no changes are Seing made to fuel which affect fuel handling methods. No change o the plant other than that described for fuel is being made. Dus, no new failure modes are being introduced. Operation of the plant in accordance with the proposed license amendment will not create any initiators for accidents, including any accidents that may be different than already evaluated in the FSAR.
3) The proposed license amendment does not involve a significant reduction in a margin of safety because increasing the fuel enrichment does not change the conclusions of the accident analyses or safety limits of the plant. This analysis does not decrease the margin of safety as described in the bases to any Technical Specification. Tne analysis does not adversely affect the operation of the fuel.
Page 6 of 9 l
B. Increased Allowable Enrichment of New Fuel in the New Fuel Storace Racks
1) Operation of Joseph H. Farley Units 1 & 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated with respect to new fuel in the new fuel storage racks because the applicable safety limits do not change and are within the bounds previously established. Neither actuation of safety systems nor accident mitigating capabilities are adversely affected by operation of the plant in accordance with the proposed license amendment. The analysis demonstrates that the proposed amendment does not pose a ch911enge to installed safety systems. Therefore, no new performance requirements are being imposed on any system or component important to safety such that any design criteria will be exceeded. The implementation of the criticality reanalysis is not an initiator for any of the postulated FSAR accidents analyzed.
This analysis does not impact accident analyses or plant accident scenarios. The analysis does not impact the accidents as analyzed in the FSAR. All accident acceptance criteria continue to be met.
In addition, the Keff design limits of 0.95 for the full water density condition and 0.98 for the optimum moderation condition are not exceedcJ. Therefore, dose calculations are not affected by this reanalysis. The ability to mitigate the consequences of any accidents analyzed in the FSAR is not hdversely affected by the implementation of the criticality reanalysis. As such, the conclusions presented ir, the FSAR remain valid such that no increase in radiological consequences will result.
2) The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated with respect to new fuel in the new fuel storage racks because no changes are being made to fuel which affect fuel handling methods. No change to the plant other than that described for fuel is being made. Thus, no new failure modes are being introduced. Operation of the plant in accordance with the proposed license amendment will not create any initiators for accidents, including any accidents that may be different than already evaluated in the FSAR.
3) The proposed license amendment does not involve a significant reduction in a margin of safety with respect to new fuel in the new fuel storage racks because increasing the fuel enrichment does not change the conclusions of the accident analyses or safety limits of the plant. The Keff design limits of 0.95 for the full water density condition and 0.98 for the optimum moderation condition continue to be met.
Page 7 of 9
l C. Increased Allowable Enrichment of fuel in the Soent fuel Storaae Radi
1) Operation of Joseph M. farley Units 1 & 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated with respect to fuel in the spent fuel storage racks because the applicable safety limits do not change and are within the bounds previously established. Neither actuation of safety systems nor accident mitigating capabilities are adversely affected by operation of the plant in accordance with the proposed license amendment. The analysis demonstrates that the proposed amendment does not pose a challenge to installed safety systems, lherefore, no new performance requirements are being imposed on any system or component important to safety such that any design criteria will be exceeded. The implementation of the criticality reanalysis is not an initiator for any of the postulated FSAR accidents analyzed.
This analysis does not impact accident analyses or plant accident scenarios. The analysis does not impact the accidents as analyzed in the FSAR. All accident acceptance criteria continue to be met.
In addition, the Keff design li .it of 0.95 is not exceeded.
Therefore, dose calculations are not affected by this reanalysis.
The ability to mitigate the con 9quences of any accidents analyzed in the FSAR is not adversely affected by the implementation of the criticality reanalysis. As such, the conclusions presented in the FSAR remain valid such that no increase in radiological consequences will result.
2) The proposed license amendment dues not cr3 ate the possibility of a new or different kind of accident from any accicent previously evaluated with respect to fuel in the spent fuel storage ry.ks because no changes are being made to fuel which af fect Nei handling methods. No change ts the plant other than that described for fuel is being made. Thus, no new failure modes are being introduced. Operation of the plant in accordance with the proposed license amendment will not create any initiators for accidents, including any accidents that may be different than already evaluated in the FSAR.
3) The proposed license amendment does not involve a significant reduction in a margin of safety with respect to fuel in the spent fuel storage racks because increasing the fuel enrichment does not change the conclusions of the accident analyses or safety limits of the plant. The Keff design limit of 0.95 continues to be met.
Page 8 of 9
CONCLUSION Based on the preceding analysis it is concluded that operation of Joseph M.
Farley Units 1 & 2 in accordance with the proposed amendment does not result in the creation of a significant hazards consideration, a significant increase in the probability of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, nor involve a significant reduction in any margins to plant safety. Therefore, the license amendment does not involve a Significant Hazards Consideration as defined in 10 CFR 50.92.
Page 9 of 9
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ML20217C419

Text

1.0 INTRODUCTION

SUMMARY

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