ML18106B054

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Criticality Analysis of Salem Units 1 & 2 Fresh Fuels Racks.
ML18106B054
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Site: Salem  PSEG icon.png
Issue date: 01/31/1994
From: Bradfute J, Melko P, John Miller
Public Service Enterprise Group
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NUDOCS 9902090149
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Criticality Analysis of the Salem Units 1 and 2 Fresh Fuel Racks January 1994 J. L. Bradfute J. W. Miller '

P. J. Melko, Jr.

W. D. Newmyer Verified: (' * ~,vJ~. lk.

S. Srinilta Core Design B Approved: C/.~

C. R. Savage, Manager Core Design B 9902090149 990202 ! ;

PDR ADOCK 050002721 p *- .

PDR*,

Table of Contents 1.0 Introduction .............................................................................................................. I

1. l Design Description..................................................................................................... 1 1.2 Design Criteria ........................................................................................................... l 2.0 Analytical Methods .....*.....*..........*...........................................................................2 2.1 Criticality Calculation Methodology .......................... ~ .............................................. 2 2.2 Reactivity Equivalencing for IFBA Credit .................................................. :............. 3 3.0 Criticality Analysis of the Fresh Fuel Racks ......................................................... 4 3.1 Full Density Moderation Analysis ............................................................................. 5 3.2 Low Density Optimum Moderation Analysis ............................................................ 6 3.3 IFBA Credit Reactivity Equivalencing .....................................................................?

4.0 Discussion of Postulated Accidents ......................................................................... 11 4.1 Fresh Fuel Storage Racks........................................................................................... 11 5.0 Summary of Criticality Results .............................................................................. 12 Bibliography ............................................................................................................ .23 Table of Contents

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ii

List of Tables Table 1. Fuel Parameters Employed in the Criticality Analysis .................................. 13 Table 2. Benchmark Critical Experiments ................................................................... 14 Table 3. Benchmark Critical Experiments PHOENIX Comparison ............................ 15 Table 4. Data for U Metal and U0 2 Critical Experiments (Part 1 of 2) ...................... 16 Table 4. Data for U Metal and U0 2 Critical Experiments (Part 2 of 2) ...................... 17 Table 5. Salem Units 1 and 2 Fresh Fuel Rack IFBA Requirement. .......................... 18 List of Tables m

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List of Illustrations Figure l.

Salem Units l and 2 Fr~sh Fuel Rack Array LayoTit.. .................................. 19 Figure 2.

Salem Units I and 2 Fresh Fuel Rack Cell Layout ....................................... 20 Figure 3.

Sensitivity of l<eff to Water Density in the Salem Units I and 2 Fresh Fuel Racks ...................................................................................................... 21 Figure 4.

Salem Units I and 2 Fresh Fuel Rack Minimum IFBA Requirement.. ........ 22 List of Illustrations v

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vi

e 1.0 Introduction This report presents the resul~ of a cri~cality analysis of the PSE&G S::ilem- tJnits- l and 2 fresh fuel storage raclcs. The fresh fuel rack design considered herein is an existing array of fuel racks, previously qualified for storage of 17x17 fuel assemblies with enrichments up to 4.5 w/o u2J5.

The Salem fresh fuel storage racks are being reanalyzed to allow storage of the Westinghouse l 7x 17 V5H fuel assembly type. The following storage configuration and enrichment limits are considered in this analysis:

Fresh Fuel Racks Storage of fresh fuel assemblies with nominal enrichments up to 4.65 235 w/o U utilizing all available storage cells. Fresh fuel assemblies with higher initial enrichments (up to 5.0 w/o U235 ) can also be stored in these racks provided a minimum number of Integral Fuel Burnable Absorbers (IFBA) are present in each fuel assembly. IFBAs consist of neutron absorbing material applied as a thin zirconium diboride (ZrB 2) coating on the outside of the U0 2 fuel pellet. As a result. the neutron absorbing material is a non-removable or integral pan of the fuel assembly once it is manufactured.

The Salem fresh fuel rack analysis is based on maintaining Keff S 0.95 for storage of 17x17 fuel assemblies under full water density conditions and S 0.98 under low water density (optimum moderation) conditions. The optimum moderation condition applies only to fresh fuel racks since these racks are used to store fuel in a dry environment 1.1 Design Description The Salem fresh fuel rack layout is shown in Figure 1 on page 19. The Salem fresh fuel rack storage cell design is shown in Figure 2 on page 20.

The fuel parameters relevant to this analysis are given in Table 1 on page 13. With the simplifying assumptions employed in this analysis (no grids, sleeves, axial blankets, etc.), the fuel parameters given in Table 1 on page 13 are appropriate for Westinghouse 17xl7 V5H, Vantage+, and Performance+ assembly types. All calculations performed in.this report consider storage of these fuel assembly types in the Salem fresh fuel racks. For convenience, a reference only to the Westinghouse l 7xl 7 VSH will be used throughout this report.

1.2 Desip* Criteria 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 minimum separation between 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 effective neutron multiplication 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 NRC guidanceCO, and less than 0.98 under low water density (optimum moderation) conditions as recommended by NUREG-0800.

Salem Units 1 and 2 Fresh Fuel Racks

e 2.0 Analytical Methods

~.1 Lrn1ca1ny - - ~

La1cu1anon_ - - .. ... .

-*- -*- .. *- 1v1etnodoiugy T~~ criticalit_Y calculation method and cross-section values are verified by comparison with cnucal expenment data for fuel assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncenainty will apply to rack conditions which include strong neutron absorbers. large water gaps and low moderator densities.

The design method which insures the criticality safety of fuel assemblies in the fuel storage rack 23 uses the AMPX< * > system of codes for cross-section generation and KENO vaC 4 l for reactivity determination.

The 227 energy group cross-section library that is the common starting point for all cross-sections used for the benchmarks of KENO Va and the KENO Va storage rack calculations is generated from ENDF/B-v< 2> data. The NITAWL( 3) 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 weighting of cross-sections is performed by the XSDRNPM(3)program which is a one-dimensional S transport theory code. These multi.group cross-section sets arc then used as input to KENO viiI which is a three dimensional Monte Carlo theory program designed for reactivity calculations.

A set of 44 critical experiments< 5*6*7*8*9) has been analyzed *using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and uncertainty. The benchmark experiments cover a wide range of geometries, materials. and enrichments, ranging from relatively low enriched (2.35, 2.46, and 4.31 w/o), water moderated.

oxide fuel arrays separated by various materials (B 4C, aluminum, steel, water, etc.) that simulate LWR fuel shipping and storage conditions to dry, harder spectrum, uranium metal cylinder arrays at high enrichments (93.2 w/o) with various interspersed materials (Plexiglass and air).

Comparison with these experiments demonstrates the wide range of applicability of the method.

Table 2 on page 14 summarizes these experiments.

The highly enriched benchmarks show that the criticality code sequence can correctly predict the reactivity of a hard spectrum environment, such as the optimum moderation condition often considered in fresh rack and shipping cask analyses. However, the results of the 12 highly enriched bencbmlrks are not incorporated into the criticality method bias because the enrichments are well above any encountered in commercial nuclear power applications. Basing the method bias solely on the 32 low enriched benchmarks results in a more appropriate and more conservative bias.

The 32 low enriched, water moderated experiments result in an average KENO Va Kerr of 0.9933.

Comparison with the average measured experimental Keff of 1.0007 results in a ~ethod bias of 0.0074. The standard deviation of the bias value is 0.0013 &. The 95/95 one-sided tolerance limit factor for 32 values is 2.20. Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.0029 2 Salem Units 1 and 2 Fresh Fuel Racks

,1K. This KENO Va bias and uncertainty are consistent with the previous Westinghouse bias and uncertainty calculated for KENO rvOO).

2.2 Reactivity Equivalencing for IFBA Credit Storage of fresh fuel assemblies with initial enrichments higher than that allowed by the methodology described in Section 2.1 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. A series of reactivity calculations is performed to generate a set of enrichment-IFBA ordered pairs which all yield an equivalent Keff when the fuel is stored in the Salem fresh fuel racks.

The data points on the reactivity equivalence curve are generated with the transport theory computer code, PHOENIX{l l). PHOENIX is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMs< 12) library is used with PHOENIX.

The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in Table 3 on page 15. Key parameters describing each of the 81 experiments are given in Table 4 on page 16. These reactivity comparisons again show good agreement between experiment and PHOENIX calculations.

Uncertainties associated with the IFBA dependent reactivities computed with PHOENIX are accounted for in the development of the individual reactivity equivalence limits. For IFBA credit applications, an uncertainty of approximately 10% of the total number of IFBA rods is accounted for in the development of the IFBA requirements. Additional information concerning the specific uncertainties included in the Salem IFBA credit limit is provided in a Section 3.3. l of this report Salem Units I and 2 Fresh Fuel Racks 3

3.0 Criticality Analysis of the Fresh Fuel Racks This section describes the analytical techniques and models employed to perform the criticality analysis for the storage of fresh fuel in the Salem fresh fuel storage racks. The complete criticality analysis of the fresh fuel racks is performed in four separate steps using the methodology outlined in Section 2 of this repon. The four steps of this complete analysis are outlined below:

Step I The fresh fuel racks are analyzed for the full density water flooding condition under nominal and "worst case" scenarios (highest enrichments without IFBA of 4.65 w/o and 4.70 w/o U 235 respectively were considered). The KENO results for the "worst case" model are then used to develop the maximum Kett which is compared to the criticality safety limit of 0.95. Details of this analysis are outlined in Section 3.1 Step 2 The fresh fuel racks are analyzed for optimum moderation water flooding conditions under the "worst case" scenario (5.05 w/o U235 ). The KENO results are then used to develop the maximum Kett which is compared to the criticality safety limit of 0. 98.

Details of this analysis are outlined in Section 3.2 Step 3 The fresh fuel racks are analyzed to determine the number of IFBA rods that must bC present in fuel assemblies with enrichments from 4.65 w/o to 5.0 w/o u235 . The result of this analysis is a curve which determines how many IFBA rods must be present in an

  • assembly to meet criticality safety limits for storage in the fresh fuel racks. Details of this analysis are outlined in Section 3.3.1.

Step 4 To provide an alternative method for determining the acceptability of fuel assembly storage given in Step 3 above, the infinite multiplication factor, Keo. for a nominal fresh 4.65 w/o fuel assembly in cold reactor geometry is determined. Details of this analysis are outlined in Section 3.3.2 Since the fresh fuel racks are normally maintained in a dry condition, the criticality analysis will show that the rack Kett is less than 0.95 for the accidental full water density flooding scenario and less than 0.98 for the accidental low water density (optimum moderation) .flooding scenario.

The following assumptions were used to develop the KENO model for the storage of fresh fuel in the Salem fresh fuel storage racks under full density and low density optimum moderation

, conditions:

l. The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 17xl7 V5H design (sec Table I on page 13 for fuel parameters).
2. The fuel assembly is modeled at its most reactive point in life.
3. All fuel rods contain uranium dioxide at an enrichment of 4.65 w/o (nominal) and 4.70 w/o

("worst case") over the entire length of each rod for the full density flood scenario.

4. All fuel rods contain uranium dioxide at an enrichment of 5.05 w/o ("worst case") over the entire length of each rod for the optimum moderation flood scenario.
5. The fuel pellets arc modeled assuming a value of 95% of theoretical density and 1.2074%

dishing fraction for nominal conditions, and a value of 96% theoretical density and no dishing fraction for "worst case" conditions.

4 Salem Units I and 2 Fresh Fuel Racks

6. No credit is taken for any natural or reduced enrichment axial blankets.

. . tak- f .. ..,. ,'1.,.

7. No eredlt is en or any U .. or U.._,v in the fuel. nor is any credit taken for the buildup cf J ..

fission product poison material. -

8. No credit is taken for any spacer grids or spacer sleeves.
9. No credit is taken for any burnable absorber in the fuel rods.

l 0. For both the full density and optimum moderation cases, there is no boron in the water.

11. Fuel rods are modeled with a fuel stack height which is infinitely long for- the full density moderation scenario and 144 inches long for the optimum moderation scenario.

3.1 Full Density Moderation Analysis ln the KENO model for the full density moderation analysis, the moderator is pure water at a conservative density of 1.0 gm/cm3. The fuel array is infinite in lateral (x and y) and axial (vertical) extent which precludes any neutron leak.age from the array.

The KENO calculation for the nominal case resulted in a 'Keff of 0.9296 with a 95 percent probability/95 percent confidence level uncertainty of +/-o.0057 LiK.

The maximum 'Keff for the full density flooding scenario arises from consideration of mechanical and material thickness tolerances resulting from the manufacturing 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 fuel assemblies yield equal or conservative results in rack Keff*

Due to the relatively large cell spacing, the small tolerances on center-to-center spacing are not considered since they will have an insignificant effect on the fuel rack reactivity. Furthermore, fuel parameters of 4.70 w/o u235 , 96% theoretical density and 0% dishing fraction are assumed to conservatively account for fuel parameter variability. Thus, the most conservative, or "worst case" KENO model of the fresh fuel storage racks contains symmetrically placed fuel assemblies at 4.70 w/o u235

  • The KENO calculation for the "worst case" full density flooding scenario resulted in a 'Keff of 0.9324 with a 95 percent probability/95 percent confidence level uncertainty of +/-o.0064 LiK.

The following equation is used to develop the maximum Kett for the Salem fresh fuel storage racks under full density water conditions:

where:

Kfull = maximum KENO Ket! with full density water Salem Units 1 and 2 Fresh Fuel Racks 5

8 method

= method bias detennined from benchmark critical comparisons ksfull = 95/95 uncenainty in the full density water KENO KetI ks method = 95/95 uncenainty in the method bias Substituting calculated values in the order listed above, the result is:

Keff = 0.9324 + 0.0074 + J0.00642 + 0.00~92 = 0.9468 Since l<eff is less than 0.95 including uncertainties at a 95/95 probability/confidence level, the acceptance criteria for criticality is met for the Salem fresh fuel storage racks under full density water flooding conditions for storage of Westinghouse 17x 17 V5H fuel assemblies with nominal enrichments up to 4.65 w/o u235 .

3.2 Low Density Optimum Moderation Analysis For the low density optimum moderation analysis, the fuel array is finite in all directions. "Wor~t case" assumptions on U02 material properties are used in modeling the entire fresh fuel rack array. All available fresh fuel storage cells are utilized as depicted in Figure 1 on page 19. Fuel rods are modeled 'with a nominal active fuel length of 144 inches. Concrete walls and floor are modeled. Under low water density conditions, the presence of concrete is conservative because neutrons are reflected back into the fuel array more efficiently than they would be with just low density water. The area above the fresh fuel racks is filled with water at the optimum moderation density.

Analysis of the Salem fresh fuel racks shows that the maximum rack l<eff under low density moderation conditions occurs at 0.05 gm/cm3 water density. The Ketr of the fresh fuel racks at 0.05 gm/cm3 water density is 0.8974 with a 95 percent probability/95 percent confidence level uncertainty of +/-o.0066 &c. Figure 3 on page 21 shows the fresh fuel rack reactivity as a function of water density in the range where the optimum moderation peak occurs. Over the remainder of the density range between 0.10 gm/cm3 and 1.00 gm/cm3 (fully flooded), the rack reactivity will be less than the values calculated for the optimum moderation peak or for the full density condition.

The following equation is used to develop the maximum Ketr for the Salem fresh fuel storage racks under low water density optimum moderation conditions:

2 Keff = Koptimum + B;,,ethod + Jksoptimum +ks melhol where:

Koptimum = maximum KENO Keff with low density optimum moderation 6 Salem Units I and 2 Fresh Fuel Racks

8 method = method bias determined from benchmark critical comparisons ks optimum

. = 95/95 uncenainty in the low density optimum moderation KENO Kerr ks method = 95/95 uncenainty in the method bias Substituting calculated values in the order listed above, the result is:

Keff = 0.8974 + 0.0074 + J0.00662 + 0.00292 = 0.9120 Since Kerr is less than 0.98 including uncertainties at a 95/95 probability/confidence level, the acceptance criteria for criticality is met.

3.3 IFBA Credit Reactivity Equivalencing Storage of fuel assemblies with nominal enrichments greater than 4.65 w/o U 235 in the Salem fresh fuel storage racks is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with the*

addition of Integral Fuel Burnable Absorbers (IFBAi 13>. IFBAs consist of neutron absorbing material applied as a thin zirconium diboride (ZrBi) 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.

Two analytical techniques are used to establish the criticality criteria for the storage of IFBA fuel in the fuel storage racks. The first method uses reactivity equivalencing to establish the poison material loading required to meet the criticality limits. The poison material considered in this analysis is a ZrB2 coating manufactured by Westinghouse. The second method uses the fuel assembly infinite multiplication factor to establish a reference reactivity. The reference reactivity point is compared to the fuel assembly peak reactivity to determine its acceptability for storage in the fuel rack.

3.3.1 IFBA Requirement Determination A series of reactivity calculations arc pctfonned to generate a set of ~A rod number versus enrichment orden:d pairs which all yield the equivalent l<eff when the fuel is stored in the fresh fuel racks. In this analysis, only fresh .(no bumup) fuel assemblies are considered since burned assemblies cannot be stored in the fresh fuel racks.

The following assumptions were used for the IFBA rod assemblies in the PHOENIX models:

1. The fuel assembly parameters relevant to the criticality analysis are based on the Westing-house 17x17 V5H design (sec Table 1 on page 13 for fuel parameters).
2. The fuel assembly is modeled at its most reactive point in life which for this case is 0 MWD/

MTUbumup.

Salem Units 1 and 2 Fresh Fuel Racks 7

3. The fuel pellets are modeled assuming values of 96% of theoretical density and no dishing fraction. *
4. No credit is taken for any natural enrichment or reduced enrichment axial blankets.
5. No credit is taken for any U 234 or U236 in the fuel. nor is any credit taken for the buildup of fission product poison material.
6. No credit is taken for any spacer grids or spacer sleeves.
7. The IFBA absorber material is a zirconium diboride (ZrB 2) coating on the fuel pellet. E*ach IFBA rod has a nominal poison material loading of 1.57 milligrams B 10 per mch, which is the minimum standard loading offered by Westinghouse for l 7x 17 V5H fuel assemblies.
8. The IFBA B 10 loading is reduced by 10% to conservatively account for manufacturing toler-ances and then by an additional 25% to conservatively model a minimum poison length of l 08 inches.
9. The moderator is pure water (no boron) with a density of 1.0 gm/cm3.
10. The array is infinite in lateral (x and y) and axial (vertical) extent This precludes any neutron leakage from the array.

Figure 4 on page 22 shows the constant Kett contour generated for the Salem fresh fuel racks:

Note the endpoint at 0 IFBA rods where the nominal enrichment is 4.65 w/o and at 24(1.0X)

IFBA rods where the nominal enrichment is 5.00 w/o. The interpretation of the endpoint data is as follows: the reactivity of the fuel rack array when filled with fuel assemblies enriched to a nominal 5.00 w/o U 235 with each containing 24(1.0X) IFBA rods is equivalent to the reactivity of the racks when filled with fuel assemblies enriched to a nominal 4.65 w/o and containing no IFBAs. The data in Figure 4 on page 22 is also provided on Table 5 on page 18 for both LOX and l .5X IFBA rods.

It is important to recognize that the curve in Figure 4 on page 22 is based on reacnv1ty equivalence calculations for the specific enrichment and IFBA combinations in actual rack geometry (and not just on simple comparisons of individual fuel assembly infinite multiplication factors). In this way, the environment of the storage racks and its influence on assembly reactivity is implicitly considered.

The IFBA requirements of Figure 4 on page 22 were developed based on the standard IFBA patterns used by Westinghouse. However, since the worth of individual IFBA rods can change depending on position within the assembly (due to local variations in thermal flux), studies were performed to evaluate this effect and a conservative reactivity margin was included in the development of the IFBA requirement to account for this effect. This assures that the IFBA requirement remains valid at intennediate enrichments where standard IFBA patterns may not be available. In addition, to conservatively account for calculational uncertainties, the IFBA requirements of Figure 4 on page 22 also include a conservatism of approximately 10% on the total number of IFBA rods at the 5.00 w/o end (i.e., about 2 extra IFBA rods for a 5.00 w/o fuel assembly).

Additional IFBA credit calculations were performed to examine the reactivity effects of higher IFBA linear a 10 loadings (l.5X and 2.0X). These calculations confirm that assembly reactivity 8 Salem Units 1and2 Fresh Fuel Racks

remains constant provided the net B 10 material per assembly is preserved. Therefore. with higher lFBA B 10 loadings. the required number of IFBA rods per assembly can be reduced by the ratio of the higher loading to the nominal 1.0X loading. For example, using 2.0X IFBA in 5.00 w/o fuel assemblies allows a reduction in the IFBA rod requirement from 24 IFBA rods per assembly to 12 IFBA rods per assembly (24 divided by the ratio 2.0X/1.0X).

The equivalent Keff for the storage of fuel in the Salem fresh fuel racks is determined using the methods described in Section 2.1 of this report. The reference conditions for this are defined by the zero IFBA intercept point in Figure 4 on page 22. The KENO Va computer code was used to calculate the storage rack multiplication factor with an equivalent nominal enrichment of 4.65 w/

o and no IFBAs. The reference KENO calculation for this case resulted in a nominal Keff of

0. 9324 with* a 95 percent probability/95 percent confidence level uncertainty of +/-o.0064 ~K.

After summation with appropriate biases and uncertainties the final Keff for the Salem fresh fuel racks was shown to satisfy the 0.95 Keff limit.

3.3.2 Infinite Multiplication Factor The IFBA credit curve given in Figure 4 on page 22 was established to identify the minimum number of IFBA rods per assembly needed to allow fuel storage in the fresh fuel racks. However.

this curve has several restrictions built into it. These restrictions are tied directly to the assumptions and parameters used in the analysis to develop the curve. These parameters, detailed in Section 3.3.1. include the fuel type, IFBA 8 10 loading, minimum IFBA length, and IFBA rod configuration. Some allowance for variation in all of these parameters has been included in the development of the IFBA credit curve and as a result may turn out to be overly conservative. The infinite multiplication factor, Koo. is used as a reference criticality reactivity point, and offers an alternative method for determining the acceptability of fuel assembly storage in the Salem fresh fuel racks. This alternative method allows for flexibility in future fuel designs.

The fuel assembly Koo calculations are performed using the Westinghouse licensed core design code PHOENIX-p( 14). The following assumptions were used to develop the infinite multiplication factor model:

l. The Westinghouse 17xl7 V5H fuel assembly was analyzed (see Table 1 on page 13 for 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 4.65 w/o U 235 over the entire length of each rod (this is the maximum nominal enrichment that can be stored in the fresh fuel racks without IFBA rods).
3. The fuel array model is based on a unit assembly configuration (infinite in the lateral and axial extent) in Salem reactor geometry (no rack).
4. The moderator is pure water (no boron) at a temperature of 68° F with a density of 1.0 gm/

cm3.

Calculation of the infinite multiplication factor for the Westinghouse 17xl7 V5H fuel assembly in Salem core geometry resulted in a reference Koo of 1.480. This includes a 1% ~K reactivity bias to conservatively account for calculational uncertainties. This bias is consistent with the standard Salem Units 1 and 2 Fresh Fuel Racks 9

conservatism included in the Salem core design refueling shutdown margin calculations.

For IFBA credit, all 17x 17 V5H fuel assemblies placed in the Salem fresh fuel racks must comply with the enrichmcnt-IFBA requirements of Figure 4 on page 22 or have an assembly ~ less than or equal to the reference ~ of l .480. By meeting either of these conditions, the maximum rack reactivity will then be less than 0. 95, as shown in Section 3.1.

10 Salem Units 1and2 Fresh Fuel Racks

4.0 Discussion of Postulated Accidents 4.1 Fresh Fuel Storage Racks Under normal conditions, the fresh fuel racks are maintained in a dry environment The

. introduction of water into the fresh fuel rack area is the worst case accident scenario. The water flooding cases analyzed in this repon are bounding accident situations which result in the most conservative fuel rack l<eff*

Other accidents can be postulated which would cause some reactivity increase (i.e., dropping a fuel assembly between the rack and wall, or dropping an assembly on top of the rack). For these other 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 other accident conditions, the absence of a moderator in the fresh fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event Experience has shown that the.maximum reactivity increase associated with postulated accidents (dropping a fuel assembly between the rack and wall, or dropping an assembly on top of the rack) is less than 10 percent ~K.

Therefore, since the normal, dry fresh fuel rack reactivity is less than 0.65, and the maximum reactivity increase for the postulated accidents is less than 10 percent ~K. the maximum rack l<eff under these other postulated accidents conditions will be less than 0.95.

Salem Units 1 and 2 Fresh Fuel Racks 11

5.0 Summary of Criticality Results The acceptance criteria for criticality requires the effective neutron multiplication factor, Keff* in the fresh fuel storage racks to be less than or equal to 0. 95, including uncertainties, under flooded conditions and less than or equal to 0.98. including uncertainties, under optimum moderation conditions.

This repon shows that the acceptance criteria for criticality is met for the Salem fresh fuel storage racks for the storage of Westinghouse 17x 17 V5H fuel assemblies with the following configuration and enrichment limits:

Fresh Fuel Racks Storage of fresh fuel assemblies with nominal enrichments up to 4.65 235 w/o u utilizing all available storage cells. Fresh fuel assemblies with higher initial enrichments (up to 5.0 w/o u235 ) can also be stored in this rack provided a minimum number of IFBAs are present in each fuel assembly.

The analytical methods employed herein conform with ANSI Nl8.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5. 7, Fuel Handling System; ANSI 57 .3-1983, "Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants"; and ANSI Nl6.9-1975, "Validation of Calculational Methods for Nuclear Criticality Safety".

12 Salem Units 1and2 Fresh Fuel Racks

Table l. Fuel Parameters Employed in the Criticality Analysis Parameter W 17x17 VSH Number of Fuel Rods per Assembly 264 Rod Zirc-4 Clad O.D. (inch) 0.3740 Clad Thickness (inch) 0.0225 Fuel Pellet 0.D.(inch) 0.3225 Fuel Pellet Density (% of Theoretical) 95 Fuel Pellet Dishing Factor (%) 1.2074 Rod Pitch (inch) 0.496 Number of Zirc-4 Guide Tubes 24 Guide Tube O.D. (inch) 0.474 Guide Tube Thickness (inch) 0.016 Number of Instrument Tubes 1 Instrument Tube O.D. (inch) 0.474 Instrument Tube Thickness (inch) 0.016 Salem Units 1 and 2 Fresh Fuel Racks 13

Table 2. Benchmark Critical Experiments Critical General Enrichment Soluble Measured KENO Reactivity Reflector Separating Material

...... Number Description U235 (w/o) Boron (ppm) Kerr (Kerr +/- One Sigma)

""" 1 U02 Rod Lattice 2.46 water water 0 1.0002 0.9966 +/- 0.0024 2 U02 Rod Lattice 2.46 water water 1037 1.0001 0.9914 +/- 0.0019 3 U02 Rod Lattice 2.46 water water 764 1.0000 0.9943 +/- 0.0019 4 U02 Rod Lattice 2.46 water B4Cpins 0 0.9999 0.9871 +/- 0.0022 5 U02 Rod Lattice 2.46 water B4Cpins 0 1.0000 0.9902 +/- 0.0022 6 U02 Rod Lattice 2.46 water B4Cpins 0 1.0097 0.9948 +/- 0.0021 7 U02 Rod Lattice 2.46 water B4Cpins 0 0.9998 0.9886 +/- 0.0021 8 U02 Rod Lattice 2.46 water B4C pins 0 1.0083 0.9973 +/- 0.0021 9 U02 Rod Lattice 2.46 water water 0 1.0030 0.9966 +/- 0.0021 10 U02 Rod Lattice 2.46 water water 143 1.0001 0.9973 +/- 0.0021 11 U02 Rod Lattice 2.46 water stainless steel 514 1.0000 0.9992 +/- 0.0020 12 U02 Rod Lattice 2.46 water stainless steel 217 1.0000 1.0031 +/- 0.0021 13 U02 Rod Lattice 2.46 water borated aluminum 15 1.0000 0.9939 +/- 0.0022 14 U02 Rod Lattice 2.46 water borated aluminum 92 1.0001 0.9882 +/- 0.0022 15 U02 Rod Lattice 2.46 water borated aluminum 395 0.9998 0.9854 +/- 0.0021 16 U02 Rod Lattice . 2.46 water borated aluminum 121 1.0001 0.9848 +/- O.CX)22 17 U02 Rod Lattice 2.46 water borated aluminum 487 1.0000 0.9892 +/- 0.0021 18 U02 Rod Lattice 2.46 water borated aluminum 197 1.0002 0.9944 +/- 0.0022 19 U02 Rod Lattice 2.46 water borated aluminum 634 l.CXl02 0.9956 +/- ().()()20 20 U02 Rod Lattice 2.46 water borated aluminum 320 1.0003 o.9893 +/- o.m22 21 U02 Rod Lauice 2.46 water borated aluminum 72 0.9997 0.9900 +/- 0.(l022 22 U02 Rod Lauice 2.35 water borated aluminum 0 1.0000 0.9980 +/- 0.0024 23 U02 Rod Lattice 2.35 water stainless steel 0 .<XXlO 0.9933 +/- O.<l022 24 002 Rod Lattice 2.35 water water 0 .0000 0.9920 +/- 0.(l024 25 U02 Rod Lattice 2.35 water stainless steel 0 .0000 0.9877 +/- 0.0022 26 U02 Rod Lattice 2.35 water borated aluminum 0 .0000 0.9912 +/- 0.()022 27 U02 Rod Lattice 2.35 water B4C 0 .0000 0.9921+/-0.0021 28 U02 Rod Lattice 4.31 water stainless steel 0 .0000 0.9968 +/- 0.0023 29 U02 Rod Lattice 4.31 water water 0 .0000 0.9963 +/- 0.0025 30 U02 Rod Lattice 4.31 water stainless steel 0 .0000 0.9950 +/- O.<Xl26 31 U02 Rod Lauice 4.31 water borated aluminum 0 .0000 0.9952 +/- 0.0025 32 U02 Rod Lattice 4.31' water borated aluminum 0 .0000 1.0006 +/- 0.0024 33 U-metal Cylinders 93.2 bare air 0 .0000 0.9968 +/- 0.0023 34 U-metal Cylinders 93.2 bare air 0 .<XXlO 1.0082 +/- 0.0025 35 U-metal Cylinders 93.2 bare air 0 .0000 0.9935 +/-.Cl.0024 36 U-metal Cylinders 93.2 bare air 0 .0000 0.9982 +/- 0.0028 37 U-metal Cylinders 93.2 bare air 0 .0000 0.9916 +/- 0.0025 38 U-metal Cylinders 93.2 bare air 0 .0000 . 0.9922 +/- 0.0025 39 U-metal Cylinders 93.2 bare pie xi glass 0 .0000 0.9972 +/- 0.0025 40 U-metal Cylinders 93.2 paraffin plexiglass 0 .<KXJO 0.9973 +/- O.IKJ2lJ 41 U-metal Cylinders 93.2 bare plexiglass {) .<KKKJ UXH lJ +/- 0.0027 42 U-metal Cylinders 93.2 paraffin plexiglass () .IKXJO 1.0 un +/- 0.0025 4.~ U-mctal Cylinders 93.2 paraffin plcxiglass 0 .IKIOO 1.0021 +/- 0.111126 44 U-metal Cylinders 93.2 para Hin plexiglass () .OIKKJ I .IHJ22 +/- 0.111129 J

Table 3. Benchmark Critical Experiments PHOENIX Comparison Description of Number of PHOENIX Keff Using Experiments Experiments Experiment Buckling U0 2.

Al clad 14 0.9947 SS clad 19 0.9944 Borated H20 7 0.9940 Subtotal 40 0.9944 U-Metal Al clad 41 1.0012 TOTAL 81 0.9978 15 Salem Units 1 and 2 Fresh Fuel Racks

Table 4. Data for U Metal and U0 2 Critical Experiments (Part I of 2)

Fuel Pellet Clad Lattice Soluble Case Cell A/0 H20/U Material Clad OD Number Type U-235 Ratio Density (glee)

Diameter (cm)

Clad (cm)

Thickness (cm)

Pitch (cm)

Boron (ppm)

I Hexa 1.328 3.02 7.53 1.5265 Aluminum 1.6916 .07110 2.2050 0.0 2 Hex a 1.328 3.95 7.53 1.5265 Aluminum 1.6916 .07110 2.3590 0.0 3 Hex a 1.328 4.95 7.53 1.5265 Aluminum 1.6916 .07110 2.5120 0.0 4 Hexa 1.328 3.92 7.52 .9855 Aluminum 1.1506 .07110 1.5580 0.0 5 Hex a 1.328 4.89 7.52 .9855 Aluminum 1.1506 .07110 1.6520 0.0 6 Hex a 1.328 2.88 10.53 .9728 Aluminum 1.1506 .07110 1.5580 0.0 7 Hex a 1.328 3.58 10.53 .9728 Aluminum 1.1506 .07110 .6520 0.0 8 Hex a 1.328 4.83 10.53 .9728 Aluminum 1.1506 .07110 .8060 0.0 9 Square 2.734 2.18 10.18 .7620 SS-304 .8594 .04085 .0287 0.0 10 Square 2.734 2.92 10.18 .7620 SS-304 .8594 .04085 .1049 0.0 II Square 2.734 3.86 10.18 .7620 SS-304 .8594 .04085 .1938 0.0 12 Square 2.734 7.02 10.18 .7620 SS-304 .8594 .04085 .4554 0.0 13 Square 2.734 8.49 10.18 .7620 SS-304 .8594 .04085 .5621 0.0 14 Squar~ 2.734 10.38 10.18 .7620 SS-304 .8594 .04085 .6891 0.0 15 Square . 2.734 2.50 10.18 .7620 SS-304 .8594 .04085 .0617 0.0 16 Square 2.734 4.51 10.18 .7620 SS-304 .8594 .04085 .2522 0.0 17 Square 3.745 2.50 10.27 .7544 SS-304 .8600 .04060 .0617 0.0 18 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 .2522 0.0 19 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 .2522 0.0 20 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 .2522 456.0 21 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 709.0 22 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1260.0 23 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1334.0 24 Square 3.745 4.51 10.37 .7544 SS-304 .8600 .04060 1.2522 1477.0 25 Square 4.069 2.55 9.46 1.1278 SS-304 1.2090 .04060 1.5113 0.0 26 Square 4.069 2.55 9.46 1.1278 SS-304 1.2090 .04060 1.5113 3392.0 27 Square 4.069 2.14 9.46 1.1278 SS-304 1.2090 .04060 '1.4500 0.0 28 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 0.0 29 Square 3.037 2.64 9.28 1.1268 SS-304 1.1701 .07163 1.5550 0.0 30 Square 3.037 8.16 9.28 1.1268 SS-304 1.2701 .07163 2.1980 0.0 31 Square 4.069 2.59 9.45 1.1268 SS-304 1.2701 .07163 1.5550 0.0 32 Square 4.069 3.53 9.45 1.1268 SS-304 1.2701 .07163 1.6840 0.0 33 Square 4.069 8.02 9.45 1.1268 SS-304 1.2701 .07163 2.1980 0.0 34 Square 4.069 9.90 9.45 1.1268 SS-304 1.2701 .07163 2.3810 0.0 35 Square 2.490 2.84 10.24 1.0297 Aluminum 1.2060 .08130 1.5113 1677.0 36 Hex a 2.096 2.06 10.38 1.5240 Aluminum 1.6916 .07112 2.1737 0.0 37 Hexa 2.096 3.09 10.38 1.5240 Aluminum 1.6916 .07112 2.4052 0.0 38 Hex a 2.096 4.12 10.38 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 39 Hex a 2.096 6.14 10.38 1.5240 Aluminum i.6916 .(17112 2.9891 0.0 40 Hex a 2.096 8.20 I0.38 1.5240 Aluminum 1.6916 .07112 3.3255 ( ).()

41 Hex a 1.307 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 ().()

42 Hex a 1.307 1.51 18.90 1.5240 Aluminum 1.6916 .117112 2.4054 ( l.()

Table 4. Data for U Metal and U0 2 Critical Experiments (Part 2 of 2)

Fuel Pellet Clad Lattice Soluble Case Cell A/0 H20/U Material Clad OD Density Diameter Thickness Pitch Boron Number Type U-235 Ratio Clad (cm)

(glee) (cm) (cm) (cm) (ppm) 43 Hexa 1.307 2.02 18.90 1.5240 Aluminum 1.6916 .07112 2.6162 0.0 44 Hexa 1.307 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 45 Hexa 1.307 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 46 Hex a 1.160 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 47 Hexa 1.160 1.51 18.90 1.5240 Aluminum 1.6916 .07112 2.4054 0.0 48 Hexa 1.160 2.02 18.90 1.5240. Aluminum 1.6916 .07112 2.6162 0.0 49 Hexa l.160 3.01 18.90 1.5240 Aluminum 1.6916 .07112 2.9896 0.0 50 Hexa 1.160 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 Hexa 1.01 18.90 1.5240 Aluminum 1.6916 .07112 2.1742 0.0 51 52 53 Hexa Hexa 1.040 1.040 1.040 1.51 2.02 18.90 18.90 1.5240 1.5240 1.5240 Aluminum Aluminum Aluminum 1.6916 1.6916 1.6916

.07112

.07112

.07112 2.4054 2.6162 2.9896 0.0 0.0 0.0 e

54 Hexa 1.040 3.01 18.90 55 Hexa 1.040 4.02 18.90 1.5240 Aluminum 1.6916 .07112 3.3249 0.0 56 Hexa 1.307 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 57 Hexa l.307 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 58 Hexa 1.307 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 59 Hexa 1.307 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 60 Hexa 1.307 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 61 Hexa 1.160 1.52 18.90 .9830 AlumiQum 1.1506 .07112 1.5926 0.0 62 Hex a 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 63 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 64 Hexa 1.160 4.02 18.90 .9830 Aluminum 1.1506 .07112 2.1742 0.0 65 Hexa 1.160 1.00 18.90 .9830 Aluminum 1.1506 .07112 1.4412 0.0 66 Hexa 1.160 1.52 18.90 .9830 Aluminum 1.1506 .07112 1.5926 0.0 67 Hexa 1.160 2.02 18.90 .9830 Aluminum 1.1506 .07112 1.7247 0.0 68 Hexa 1.160 3.02 18.90 .9830 Aluminum 1.1506 .07112 1.9609 0.0 2.1742 0.0 69 70 71 Hexa Hexa Hexa 1.160 1.040 1.040 4.02 1.33 1.58 18.90 18.90 18.90 18.90

.9830 19.050 19.050 19.050 Aluminum Aluminum Aluminum Aluminum 1.1506 2.0574 2.0574 2.0574

.07112

.07620

.07620

.07620 2.8687 3.0086 3.1425 0.0 0.0 0.0 e

72 Hexa 1.040 1.83 73 Hexa 1.040 2.33. 18.90 19.050 Aluminum 2.0574 .07620 3.3942 0.0 74 Hexa 1.040 2.83 18.90 19.050 Aluminum 2.0574 .07620 3.6284 0.0 75 Hexa 1.040 3.83 18.90 19.050 Aluminum 2.0574 .07620 4.0566 0.11 76 Hexa 1.310 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 0.11 77 Hexa 1.310 ' 3.01 . 18.88 1.5240 Aluminum 1.6916 .07112 2.9900 0.(1 78 Hexa 1.159 2.02 18.88 1.5240 Aluminum 1.6916 .07112 2.6160 (),()

79 Hexa 1.159 3.01 18.88 1.5240 Aluminum 1.6916 .07112 2.9900 II.II 80 Hexa 1.312 2.03 18.88 .9830 Aluminum 1.1506 .07112 1.7250 ().()

,....... 81 Hexa 1.312 3.02 18.88 .9830 Aluminum 1.1506 .07112 l.96IO (}.()

-....J

Table S. Salem Units 1 and 2 Fresh Fuel Rack IFBA Requirement Enrichment (w/o) IX (l.57 mg/in) IFBA l.SX (2.J6 mg/in) IFBA Rods In Assembly Rods In ~embly 4.65 0 0 4.83 12 8 5.00 24 16 18 Salem Units 1 and 2 Fresh Fuel Racks

Q'J

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. I

!Sl D D D D D D D D - - -

I I

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Ill SI J1 tO

,, JJ D D D D D D D D E Ill Ul .

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. Q)

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L D D D D D D D D D LL

(\j x

D D D D D D D D D (\j D D D D D D D D I*"°

@]

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(\J Figure 1. Salem Units 1 and 2 Fresh Fuel Rack Array Layout Salem Units 1and2 Fresh Fuel Racks 19

.....------------------------ --------1 21 .IZllZl - (CENTER TO CENTERi WESTINGHOUSE 17XI7 V5H FUEL ASSEMBLY 0 0 0 0 0 21.IZllZl" 8.432" 0

L__ _____________________________ _J Figure 2. Salem Units I and 2 Fresh Fuel Rack Cell Layout 20 Salem Units 1 and 2 Fresh Fuel Racks

0.95 0.9 0 -- ...

-- -~ ~ ,

0.8 5 0.8 0 CD t3

~ 0.7 5 UJ

~

I 0.7 0 0.6 5 0.6 0 0.55 0.00 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.10 Water Density (glee)

Error Bars Represent Plus or Minus 95/95 Sigma About KENO Calculated Kett Figure 3. Sensitivity of Kew to Water Density in the Salem Units 1 and 2 Fresh Fuel Racks Salem Units 1and2 Fresh Fuel Racks 21

24 I

,/

,/

I

/

ACCEPTABLE /

I

>- 16 / ,

15 I/

E

/

j Q)

Cl)

Cl)

<( J .,

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a..

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r ,*

i"

                  • 1.SX IFBA Loadinc J ,*

0

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4.65 4. 70 4. 75 4.80 4.85 4.90 4.95 5.00 U235 Enrichment (w/o)

Figure 4. Salem Units I and 2 Fresh Fuel Rack Minimum IFBA Requirement 22 Salem Units 1 and 2 Fresh Fuel Racks

Bibliography

l. 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.

April 14, 1978.

2. W. E. Ford III, CSRL-V: Processed ENDF!B-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, ORNL/CSDrrM-160, June 1982. *
3. N. M. Greene, AMPX: A Modular Code System for Generating Coupled Multigroup Neu-tron-Gamma Librariesfrom ENDFIB. ORNL!fM-3706, March 1976.
4. L. M. Petrie and N. F. Landers, KENO Va--An Improved Monte Carlo Criticality Program With Supergrouping, NUREG/CR-0200, December 1984.
5. M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, July 1979.
6. S. R. Bierman and E. D. Clayton, Criticality Separation Between Subcritical Clusters of 2.35 wt% 235U Enriched U02 Rods in Water with Fixed Neutron Poisons, PNL-2438, October 1977.
7. S. R. Bierman and E. D. Gayton, Criticality Separation Between Subcritical Clusters of 4.29 wt%235U Enriched U02 Rods in Water with Fixed Neutron Poisons, PNL-2615, August 1979.
8. S. R. Bierman and E. D. Gayton, Criticality Experiments with Subcritical Clusters of 2.35 wt% and 4.3 I wt% 235U Enriched U02 Rods in Water at a Water-to-Fuel Volume Ratio of 1.6, PNL-3314, July 1980.
9. J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, Nuclear Sci-ence and Engineering, Volume 52, pages 350-359, 1973.
10. D. E. Mueller, W. A. Boyd, and M. W. Fecteau (Westinghouse NFD), Qualification of KENO Calculations with ENDFIB-V Cross Sections, American Nuclear Society Transac-tions, Volume 56, pages 321-323, June 1988.
11. A. J. Hanis. A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors, WCAP-10106, June 1982.
12. Askew, J. R., Fayers, F. I., and Kemshell, P. B., A General Description of the Lattice Code WIMS, Journal of British Nuclear Energy Society, 5, pp. 564-584, 1966.
13. Davidson, S.L., et al, VANTAGE 5 Fuel Assembly Reference Core Repon, Addendum 1, WCAP-10444-P-A, March 1986.
14. Nquyen, T. Q., et al, Qualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor Cores, WCAP-11597-A, November 1987.

Bibliography-23

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