ML20211C064

Rev 0 to Final Rept Criticality Safety Analysis Hb Robinson Spent Fuel Storage Racks (Unpoisoned,Low Density) W/4.2% Enriched 15x15 Fuel Assemblies

ML20211C064
Person / Time
Site:	Robinson
Issue date:	09/09/1986
From:	Gerrald L, Pieper J, Skogen F SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML14190A959	List: ML20211C049 ML20211C064
References
XN-NF-86-107, XN-NF-86-107-R, XN-NF-86-107-R00, NUDOCS 8610210261
Download: ML20211C064 (38)

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I XN-NF-86-107, Rev. O Issue Date 9/9/86 8

I FINAL REPORT CRITICALITY SAFETY ANALYSIS H. B. ROBINSON SPENT FUEL STORAGE RACKS (UNPOISONED, LOW DENSITY)

WITH 4.2% ENRICHED 15xIS FUEL ASSEMBLIES -

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lI XN-NF-86-107, Rev. O lssue Date: 9/9/86 5 .

FINAL REPORT CRITICALITY SAFETY ANALYSIS H. B. ROBINSON SPENT FUEL STORAGE RACKS (UNPOISONED, LOW DENSITY)

WITH 4.2% ENRICHED 15xl5 FUEL ASSEMBLIES

! AUGUST 1986 I .

I Prepared by. v, dk/S

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I . D. Gerrold, Senior Engineer Corporate Licensing Date

• )/2 /f 6 Reviewed by: m J. JT. Pieper, Engine (r Date Cdrporate Licensing

), $.S . $fbfh l F. B. Skogen, Managey Date 3 PWR Neutronics Approved by: , 7 2- [

C. W. Molody, Mahoger~ y / Oate Corporate Licensi g Y

T. V. Potten, Manager ik//ls Dat e' I Neutronics and Fuel Managernent jrs i

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Ij CUSTOMER OISCLAIMER I

IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Exxon Nucteer Company's warranties and representations concoming the subject matter of this document are those set form in the Agreement between Exxon Nuclear Company, Inc. and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provedad in sucfi Agreement, norther Exxon Nuclear Company, Inc. nor any person acting on its behalf makes any werranty or representatien, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method or process disclosed in this document will not infnnge privetely owned rights; or assumes any liabilities with respect to the use of any information, apparatus, method or process disclosed in this doctrnent.

The information contained herein is for the sole use of Customer.

In order to avoid impeirment of rights of Exxon Nuclear Company, Inc.

in potents or inventions which may be included in the information contained in this document, the receient, by its acceptance of this document agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in wnting by Exxon Nuclear Company, Inc.

or until after six (6) months following termination or expiration of the aforessed Agreement and any extensaan thereof, unless otherwise expressly provided in the Agreement. No rights or licenses in or to any pater ts are implied by the fumishing of this doczament.

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r ii XN-NF-86-107, Rev. O I FINAL REPORT I CRITICALITY SAFETY ANALYSIS H. B. ROBINSON SPENT FUEL STORAGE RACKS (UNPOISONED, LOW DENSITY) l WITH 4.2% ENRICHED 15xl5 FUEL ASSEMBLIES II AUGUST 1986 I TABLE OF CONTENTS SECTION Page

1.0 INTRODUCTION

I i l.1 2.0 Summary FUEL PARAMETERS 1

2 3.0 STORAGE RACK GEOMETRY 4 4.0 CALCULATION METHODS 5 I 4.1 4.1.1 4.l.2 Normal and Credible Abnormal Conditions General Requirements Credible Abnormal Occurrences 5

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I 4.1.3 4.l.4 4.1.5 Double Contingency Fuel Characteristics Rock Characteristics 6

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4.1.6 Fuel Handling Accidents 7 4.1.7 Maximum Allowable Multiplication Factor 7 1 4.1.8 Burnup Effects 7 Fuel Characteristic Effects 7 4.1.9 4.1.10 Fuel Assembly Arrangement Effects 9 1 4.1.1 i Other Assembly Spacing Effects 9 10 4.1.12 Fixed Absorbers 4.1.13 Soluble Absorbers 10 5.0 MODERATION EFFECTS Ii 5.1 Fully Flooded Conditions 1I I 5.2 5.2.1 Low Density Moderator Effects K-infinite Calculations 13 14 16 I

6.0 6.1 SENSITIVITY STUDIES Bundle Offset Effects 17 18 7.0 FUEL HANDLING ACCIDENTS

" Bundle Drop" Accidents 18 7.1 8 7.2 Bundle-Bundle Interactiors During Handling 18 I

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B iii XN-NF-86-107, Rev. O TABLE OF CONTENTS (Cont.d)

I SECTION Page I

8.0 METHODS VERIFICATION 21 Il 8.1 Reference 2 Experiements 21 3 8.2 Reference 3 Data 22 8.3 Referer:ce 4 Dato 23 5'

8.4 Reference 5 Data 24 8.5 Acceptability Limit 26 8.6 Cross Section Comparisons 28 10.0 COMPUTER INPUT LISTINGS 29 '

10.1 KENO input: K-infinite Case at Full Flooding 29 Il.0 REFERENCES 30 Ii i

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f iv XN-NF-86-107, Rev. O i

LIST OF TABLES I Table Page 2.0 Bundle Parameters 2

5. l a Fuel Assembly Composition 11 5.lb Rod Pitch (Moderation) Ef fects Full Density Water Moderation XSDRNPM Results 12 5.lc Generic Bundle Parameters 13 5.2. l a KENO Results for K-Infinite Calculations Knight-4 Modified Hansen-Roach Cross Sections Interspersed Moderation Effects Bundles on 20.875 inch Centers 15 6.0 XSDRNPM Results Bundle Spacing Effects 100%

Interspersed Water Density 16 8.2 Fuel Handling Interactions Array / Spacing Effects 19 Bundle Handling Interactions Dissolved Boron Effects

, Two Edge-Edge BunJies 20 8.1 Benchmark Results Dato of Reference 2 22 8.2 Benchmark Data From Referac e 3 23 8.3 Reference 4 Dato KENO-IV with Hansen-Roach Cross Sections 24 8.4a Fuel Design Parameters 25 I 8.4b Reference 5 Data Low Density Moderation Between

. Bundles KENO Results 26 LIST OF FIGURES Figure Page 2.0 Rod Arrangement Within Bundle 3 I

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XN-NF-86-107, Rev. O Page i I FINAL REPORT CRITICALITY SAFETY ANALYSIS H. B. ROBINSON SPENT FUEL STORAGE RACKS (UNPOISONED, LOW DENSITY)

WITH 4.2% ENRICHED 15xl5 FUEL ASSEMBLIES AUGUST 1986 I

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l.0 INTRODUCTION I The criticality safety of the unpoisoned, low density spent fuel storage racks with 4.2% enriched ISxl5 bundles is demonstrated in accordance with NUREG-0800 and a ANSI /ANS-57.2-1983.

l.I Summary The spent fuel storage rocks meet the applicable criticality safety criteria subject to the limits and controls given below.

m 1. Fuel design - as specified in Table 2.0.

2. 500 ppm Boron (minimum) in pool water before starting fuel movements.

3. Fuel handling inch minimum edge-edge spacing between assemblies will assure subcriticality with zero dissolved boron.

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I XN-NF-86-107, Rev. O Page 2 I

2.0 FUEL PARAMETERS Key bundle design parameters used in these calculations are listed in Table 2.0.

All fuel rods in the model were 4.20% enriched. The arrangement of the fuel rods and the instrument / guide tubes is shown in Figure 2.0.

I Table 2.0 Bundle Parameters I Parameter Design Value Model Value I Enrichment (wt% U-235)

Pellet Diameter (inch)

Pellet Density (%TD) 4.20 (max.)

0.3565 94.0 + 1.5 4.20 0.3565 95.0 Pellet Dish Volume (%) 1.0 - 0 Stock Length (inch) 132 (enr) + 12 (not) 144 (min.)

i Clad ID/OD (inch) 0.364/0.424 0.364/0.424 Rod Pitch (inch) 0.563 0.563 Gd 023 Content Variable None i Fuel Rods per Bundle 204 (max.) 204 l Guide / Inst. Tube ID/OD (inch) 0.511/0.544 0.511/0.544 l

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I XN-NF-86-107, Rev. O Page 3 I

ROW / COL l 2 3 4 5 6 7 8 9 10 Il 12 13 14 15 i F F F F F F F F F F F F F F F g 2 F F F F F F F F F F F F F F F 3 3 F F G F F G F F F G F F G F F 4 F F F F F F F G F F F F F F F F F F F G F F F F F G F F F F 5

F F G F F l

6 F F G F F F F F F F 7 F F F F F F F F F F F F F F F 8 F F F G F F F 1 F F F G F F F E 9 F F F F F F F F F F F F F F F g 10 F F G F F F F F F F F F G F F 11 F F F F G F F F F F G F F F F g F F F F F F F G F F F F F F F 12 13 F F G F F G F F F G F F G F F I

14 F F F F F F F F F F F F F F F 15 F F F F F F F F F F F F F F F Key: F = Fuel Rod l G = Guide Tube 3 i = Instrument Tube Figure 2.0 Rod Arrangement 'Nithin Bundle I

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E I XN-NF-86-107, Rev. O Page 4 I

3.0 STORAGE RACK GEOFETRY The spent fuel storage rocks were conservatively modeled as an infinite x infinite array of infinite length bundles.

! The bundles are on a nominal 21-inch square pitch within the rocks. The modeled pitch was 20.857 inches for all locations.

All rock materials of construction were neglected in the models.

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I XN-NF-86-107, Rev. O Page 5 I

4.0 CALCULATION METHODS All computer codes and cross sections used are part of the SCALE (l) system.

I The neutron multiplication factors, k inf and keff, were calculated using KENO-IV or I XSDRNPM.

The 16 group Knight-modified, Hansen-Roach cross section library was used for most calculations. Reference cases were also calculated using the 27 group library (subset of 218 group libraries based on ENDF/B-IV) and with the 123 group GAM-THERMOS library. For the reference calculations, the resonance, self-shielding calculations were performed using NITA"!L.

I All codes and cross sections have been extensively benchmarked against critical experiment data.

I Evidence of methods verification is presented later in this document.

4.1 Normal and Credible Abnormal Conditions I The conditions specified in Section 6.4.2 of ANSI /ANS-S7.2-1983 have been evaluated.

I 4.1.1 General Requirements Sections 6.4.2.1.1 through 6.4.2.1.2 of ANSI /ANS-S7.2 state that handling, transfer, and storage of spent fuel assemblies be demonstrated as subcritical.

4.1.2 Credible Abnormal Occurrences Section 6.4.2.1.3 of ANSI /ANS-S 7.2 specify the following credible abnormal I conditions for evaluation:

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XN-NF-86-107, Rev. O Page 6 I

1. Tipping 'or falling of an assembly.

2. Tipping of a storage rock.

3. Misplacement of a spent fuel assembly.

4. Fuel drop accidents.

5. Stuck fuel assembly / crane uplifting forces.

6. Horizontal movement of fuel during removal from rocks.

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7. Placing assembly along rock perimeter.

8. Falling objects.

9. Missiles generated by f ailure of rotating machinery or natural phenomena.

All credible effects are evaluated in Section 6 of this report. .

4.1.3 Double Contingency .

Section 6.4.2.1.4 of the standard requires that no single credible accident condition will result in criticality. This requirement is satisfied.

4.l.4 Fuel Characteristics Section 6.4.2.1.5 requires that the fuel parameters controlling reactivity be identified.

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XN-NF-86-107, Rev. O Page 7 I

Fuel parameters are in Section 2 of this report. The key fuel design parameters are the enrichment and the ef fective moderation level. The enrichment is a controlled parameter. Moderation effects are detailed in Section S.

I 4.l.S Rock Characteristics I Section 6.4.2.1.6 requires that the controlling rock parameters be identified.

The key parameter is the cell pitch. See Section 7 for details.

4.1.6 Fuel Handling Accidents Section 6.4.2.2 requires that hypothetical fuel handling accidents be evaluated.

Please refer to Section 7 of this report.

I 4.1.7 Maximum Allowable Multiplication Factor Sections 6.4.2.2.1 through 6.4.2.2.3 specify the methods for establishing the maximum allowable kef f. Please refer to Section 8 of this report.

I 4.1.8 Burnup Effects I Section 6.4.2.2.4 requires that the fuel be evaluated at its most reactive burnup.

Since the modeled design contains no burnable poison, the zero burnup case is most reactive.

4.1.9 Fuel Characteristic Effects Section 6.4.2.2.5 of the standard requires that the following fuel characteristics be I evaluated to assure that the most reactive credible fuel has been modeled.

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XN-NF-86-107, Rev. O Page 8 I

1. Maximum fissile loading. This analysis covers a fuel design described in Section 2 of this report. Based on a 95.5 percent TD pellet density and a 1.0 .

vol% dish, the average pellet stock density is 94.55% TD maximum while the modeled value is 95% TD. The modeled pellet stack length is 12 inches greater than the design value. The model contains the maximum credible number of f uel rods. The effect of pellet diameter tolerance is negligible.

Therefore, the maximum fissile loading is conservatively modeled.

2. Fuel rod diameter. The ef fect of the 0.0015 inch diameter tolerance is negligible.

3. Cladding effects. The clad and the pellet-clad gap were explicitly modeled at I

the nominal dimensions. Tolerance effects on the order of 0.002 inch are negligible. The clad was modeled as pure zirconium at 100 percent TD.

4. Pellet density. As stated earlier, the modeled pellet density conservatively simulates the mcximum credible pellet density.

5. Fuel rod pDcf cod number of rods in assembly. The modeled pitch is the I

nominal value. The modeled number of fuel rods is the maximum credible value. The ef fects of removing fuel rods from a flooded assembly are detailed in Section 5.l.

6. Absence of fuel rods in certain locations. The instrument / guide tubes were explicitly modeled. Other effects are given in Section 5.1.

7. Burnable poison. None.

8. Distribution of fissile content. All rods were 4.2 percent enriched. The locations are shown in Figure 2.0.

Since the fuel is at zero exposure, there is no plutonium present.

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I XN-NF-86-107, Rev. O Page 9 I

4.1.10 Fuel Assembly Arrangement Effects Section 6.4.2.2.6 of the standard specifies that the most reactive credible arrange-ment be evaluated.

I 1. Spacing between assemblies. All storage cells are on a 21-inch nominal square pitch. The cells were conservatively modeled on 20.875-inch centers.

I Sensitivity study results for bundle spacing effects are in Section 7.0 of this report.

I 2. Moderation between assemblies. An infinite rock array is acceptable for all interspersed water densities greater than cbout 20 percent. Thus, all credible moderation conditions have been considered.

I 3. Fixed neutron absorbers. None.

4. l . I l Other Assembly Spacing Effects Section 6.4.2.2.7 of the standard specifies that the fo'!owing of fects be evaluated:

Eccentricity of assembly locations and other spacing effects. Evaluation in I 1.

Section 7.0 of this report.

2. Dimensional tolerances. The bundle pitch was conservatively modeled as 20.875 inches for all locations. In addition, the results in Section 7.0 indicate that closer spacings would be acceptable.

! 3. Construction materials. The effects of neutron absorptions in the materials of construction were conservatively neglected.

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4. Moderator / fuel density effects. Already ccvered.

5. Burnable poison. None.

6. Cell wall materials. None.

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4.1.12 Fixed Absorbers Section 6.4.2.2.7 of the standard allows credit for fixed absorbers. None were modeled in this analysis.

4. f .13 Soluble Absorbers I

Section 6.4.2.2.9 specifies that soluble poison effects not be included except for PC IV and V faults.

Soluble poison ef fects were not included in the evaluation except for two remote probability accident conditions. These hypothetical accidents allow two bundles to be placed closely together during fuel handling or by placing a bundle at the edge of the rack.

Administrative controls will be implemented to assure that the minimum specified dissolved boron concentration be maintained during fuel handling. These administra-tive controls will satisfy the double contingency requirement.

All criticality safety aspects of ANSI /ANS-57.2 have been covered.

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XN-NF-86-107, Rev. O Page lI I

5.0 MODERATION EFFECTS 5.1 Fully-Flooded Conditions I If fuel rods are removed from an assembly, the keff for flooded assemblies may be effected.

I The modeled fuel design (i.e., 204 fuel rods and 21 instrument / guide tubes) is composed as noted in Table 5.la.

I Table 5.la Fuel Assembly Composition Material Volume Percent Fuel 28.55 Pellet-Clad Gap I.21 I Clad (Zr)

Water I l.43 58.81 Total 100.0 I The average water / fuel volume ratio (Vw/Vf) is 2.06. If the entire 15xl5 array was fuel rods, the Vw/Vf would be 1.76.

The ef fect of changes in Vw/Vf on keff was evaluated for generic assemblies using XSDRNPM. The generic assemblies were modeled with the nominal pellet and clad dimensions, but with rod pitches to yield Vw/Vf ratios in the range 1.0 to 4.0. The calculations were made as follows:

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1. Self-shielded 16 group crcss sections were generated using BONAMI/NITAWL.

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XN-NF 107, Rev. O Page 12 I

2. The unit rod cell was cell-weighted using XSDRNPM. The k i ng for the infinite rod lattice is listed in Table 5.lb. .

3. These cell-weighted (homogeneous) cross sections were then used in a XSDRNPM model of 8.445" x 8.445" x infinite bundles on 20.875 inch centers.

Full density water was within and between bundles. The assembly and the water between assemblies were modeled as concentric cylindrical regions. The king for the infinite bundle lattice is listed in Table 5.lb.

I Table 5.lb Rod Pitch (Moderation) Effects Full Density Water Moderation XSDRNPM Results Rod k i nf k i nf Pitch (Rod (Bundle Vw/Vf (cm) Lattice) Lattice) 1.0 1.2470 1.3765 0.8333 1.5 1.3700 1.4458 0.8896 2.0 1.4829 1.4791 0.9209 3 2.5 1.5877 1.4939 0.9378 3 3.0 1.686I l.4977 0.9454 3.5 1.7790 1.4933 0.9465 4.0 1.8673 1.4838 0.9432 The optimum Vw/Vf is near 3.5. An infinite array of flooded bundles on 20.875 inch centers is acceptable at any rod pitch. The actual assembly (l.43 cm pitch) is obviously undermoderated even with 21 instrument / guide tubes.

Due to the fixed 8.445 inch square bundle modeled with various rod pitches, the bundles are not 15xl5 rod arrays. The generic bundle parameters are listed in Table 5.lc.

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F I XN-NF-86-107, Rev. O Page 13 I l I Table 5.lc Generic Bundle Parameters Array Vw/Vf (Rods / Edge) Total Rods I.0 17.2 295.9 l.5 15.7 245.1 I 2.0 2.5 3.0 14.5 13.5 12.7 209.2 I82.5 161.8 I 3.5 4.0 12.1 I l .5 145.4 132.0 The peak activity condition (Vw/Vf = 3.5) corresponds to about 80 removed fuel rods; i.e., about 59 in addition to the 21 instrument / guide tubes. The reactivity of the generic bundle filled with rods at optimum pitch is bounding for the actual bundle with removed rods.

I 5.2 Low Density Mcderator Effects Flooding with full density water is the most reactive state for a single bundle, or for on array of bundles placed closely (one rod pitch or less) together.

As the edge-edge spacing between bundles in a flooded array is increased above I zero, the arrav k fr declines to the asymptotic va!ue for a single bundle with full water reflection; 1.e., bund!e-bundle interactions even vally become negligible. This asymptotic keff is 0.917 1 0.005. It will be shown that the 21 inch (20.875 inches) center-center spacing between bundles in the rocks is more than adequate to result in negligible bundle-bundle interaction at full flooding.

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XN-NF-86-107, Rev. O Page 14 as As the water density within the rocks decreases below 100 percent (full density ,

water flooding), the system reactivity declines to a minimum value in the vicinity .

of 40 percent water density and then rises to a peak value near 5 percent water density and then declines to dry conditions.

An infinite system is adequately subcritical for all interspersed water densities ,

greater than about 20 percent. The minimum credible water density within the rocks is certainly much greater than 20 percent. The density of water at 1000C is 0.958 g/cc (95.8 percent density).

5.2.1 K i nf Calculations A single bundle was explicitly modeled using KENO-IV. The bundle was centered in a 20.875" x 20.875" x infinite cuboid of between-bundle water of the same density as the within-bundle water. The fuel zone length was infinite.

With specular reflection at the six faces of the cuboid, the model simulates an infinite x infinite array of infinite length bundles on 20.875 inch centers.

The KENO results at several moderator densities are listed in Table 5.2.la.

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. XN-NF-86-107, Rev. O Page 15 I

I Table 5.2.la KENO Resulty ic,. k;nf Calculations I Knight-Modified Hansen-Roach Cross Sections Interspersed Moderation Effects Bundles on 20.875 inch Centers I Water Density Sig-m (eff) DANCOFF

(%) (U-235/U-238) (Bundle-Average) k i nf 20 1007/44.7 0.564 0.944 + 0.0039 40 1250/56 0.521 0.692 - 0.0082 I 60 80 100 l450/60 1500/65 1593/70.7 0.405 0.325 0.243 0.744 0.917i0.005 0.0076 0.829 I 0.0090 I

These results indicate that on infinite system would be adequately subcritical at all I credible, interspersed water densities.

Cell weighted macroscopic cross sections were prepared for the fully flooded case.

The cell weighting was based on the KEMO fluxes f rom the kinf calculations.

These macroscopic cross sections were usen in subsequent calculations which employed a homogeneous bundle model.

I For reference, the k nfi values for the homogeneous zun& a 1.5069. This is a conservative model.

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6.0 SENSITIVITY STUDIES The assumed temperature for these models is 200C or less. Since keff will tend to decrease with increasing temperature, the model is conservative.

No credit is taken for neutron absorption in the steel / aluminum materials of the I racks. All neutron absorptions occur in the fuel, the moderator, or the reflector.

This is a conservative model.

The key parameter for sensitivity evaluation is the bundle spacing within the rocks.

The modeled pitch between bundles is conservative. For reference, the effect of pitch on k nfi was evaluated for a flooded system. For lower moderator densities, the peak keff will change little, if any, but the optimum water density will increase with decreasing bundle-bundle spacing.

I These calculations were pt s .ned using XSDRNPM. The bundle and water between bundles wer modeled as conventric ef lindr. cal : ions of infinite 'ength. Results are listed in Table u.0.

Table 6.0 I XSDRNPM Results Bundle Spacing Effects 100% interspersed Water Density I Center-Center Spacing (inch) kri i

20.875 0.9155 20.0 0.9 I 6 I I I 9.0 I 7.0 I5.0 0.9172 0.9227 0.9403 13.0 1.0000 I .

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The Table 7.0 results indicate: .

l. The XSDRNPM result at 20.875 inch pitch is statistically identical to the KENO value in Table 2 (0.917 1 0.005).

2. The kinf would be less than 0.95 for all pitches greater than about 15 inches.

6.1 Bundle Offset Effects The bundles are constrained on 21-inch nominal centers in the subject rocks by channels or guides slightly larger than the bundle dimensions.

The nominal guide dimensions are 8.9375 inch squore. Therefore, assuming a 20.875 inch pitch for the guides, if . 2x2 array of bundles were brought together as closely as possible at the near edges of the contiguous guides, the bundles would be on 20.3825 inch center s. Based on the results in Table 6.0, this spacing would be acceptable even for an infinite system. This could not be achieved actually since the spacing is increased on two sides of the guides as the bundles are moved together.

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7.0 FUEL HANDLING ACCIDENTS The accidents considered include:

I 1. A single bundle placed edge-edge to the side or top of the spent fuel rocks

(" Bundle Drop").

2. Two bundles not in the rock placed closely together, I 7.1 " Bundle Drop" Accidents '

I A bundle placed horizontally on the top of the racks will be neutronically isolated from the nearest fuel in the bundles below.

A vertical bundle placed at the edge of the rocks may result in nearly edge-edge I contact of two bundles. Since there is negligible interaction among the bundles within the ract<% ibis is equivalent to the bundle handling accident analyzed in Sectior /J 7.2 Bundle-bundle Interactions During Handling Placing two or more bundles closely together was considered as a hypothetical I accident condition.

All cases assumed fully flooded bundles (12 foot length), with full water reflection (30 cm water), at all six f aces of the smallest cuboid enclosing the system model.

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Table 8.2

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Fuel Handling Interactions Array / Spacing Effects Center-Center Bundle Array Spacing (inch) keff IxI 0.92I + 0.0056 2xl 8.445 1.048 + 0.0053 E 2xl 10.445 1.012 7 0.0058 3 2xl 12.445 0.943 I 0.0056 2x1 I3.445 0.940 7 0.0060 2xl l 4.445 0.923 i 0.0049 2x2 8.445 1.214 + 0.0050 I

These. results indicate:

1. The kef f for a single, fully reflected bundle (12 foot length) is statistically I

identical to that for an infinite x infinite array of infinite length bundles on 20.875-inch centers. As expected, the 12.43 inches of water between adjacent bundle envelopes is adequate to isolate fully-flooded bundles.

2. Two closely placed bundles may become critical if flooded and fully reflected.

The ef f ect of dissolved boron concentration on the keff of two edge-edge bundles I

(8.445-inch centers) was examined. As before, the system was surrounded by 30 cm of water. The water within and around the bundle contained the boron concentration indicated below.

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I Bundle Handling Interactions Dissolved Boron Effects Two Edge-Edge Bundles I Dissolved Boron Concentration { ppm) keff 500 0.900 + 0.0043 I 1000 1500 2000 0.812 + 0.0045 0.733 + 0.0042 0.686 i 0.0038 I

These data indicate that 500 ppm boron is more than adequate to lower the keff to less than 0.95 for this spacing condition.

Therefore, either 500 ppm minimum dissolved baron concentration or a 7-inch minimum edge-edge spacing between bundles will assure criticality safety during fue; handling.

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I XN-NF-86-107, Rev. O Page 21 I

8.0 METHODS VERIFICATION Supplemental benchmarking of the methods employed in this analysis were performed. Critical experiments documented in references 2-5 were modeled using methods identical to those of this report. The critical experiments include bundle I arrays with variable bundle-bundle spacings, and with and without neutron absorber rods / plates between the bundles.

8.1 Reference 2 Experiments Reference 2 experiments include a 3x3 array of 14xl4 bundles. The rods contain 2.46 percent enriched UO2 pellets on a 1.636 cm square pitch. Five of the experiments were selected for this benchmark. These cases contain little, if any, dissolved boron in the moderator (water), and include the desired effects of neutron absorbers. The other cases, not selected for benchmarking, include effects such as I dissolved boron content ar.d slight temperature changes.

The critical moderator height was determined in these experiments. The reported kef f's were normalized to a constant moderator height for each of the two classes of experiments. Therefore, the observed keff's are not all unity. The data are in Table 8.1.

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I XN-NF-86-107, Rev. O Page 22 I

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Table 8.1 ~I

] Benchmark Results Data of Reference 2 keff keff keff g Case Number (Observed) (Calculated) (95% UL) 5 2321 1.0030 + 0.0009 0.997 + 0.005 1.007 2317 1.0083 + 0.0012 1.004 + 0.004 1.012 2378 1.0000 I 0.0010 1.009 7 0.005 1.019 2396 1.0001 7 0.0019 1.004 I 0.004 1.012 2420 0.9997 i 0.0015 1.002i0.004 1.010 Average 1.0022 1 0.0016 1.0032 1 0.0019 1.0120

] -

The 95 percent upper limit on the calculated kef f, which is the parameter used in g

judging acceptability, exceeds the observed keff in every case. The average of the 3 individual biases (calculated minus observed) is 0.00098 + 0.0028.

8.2 Reference 3 Data I

Reference 3 includes data on experiments using 2.35 and 4.31 percent enriched UO2 rods in a lx3 bundle array. Only the 4.31 percent enriched cases were selected for g this benchmark. These cases were eit.her 8xl3 bundles (2.54 cm rod pitch) or 5 16xl2 bundles (1.892 cm pitch). The critical separation between the bundles was determined with various neutron absorbers between the bundles and with various spacings to a thick steel wall.

In these cases, the observed keff's are all 1.000.

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I XN-NF-86-107, Rev. O Page 23 I

I Table 8.2 Benchmark Data From Reference 3 I Distance to Rod Pitch Steel Wall Neutron (cm) (cm) Absorbers keff 2.54 0 0.999 + 0.006 I 2.54 2.54 6.6 26.16 1.001 + 0.005 1.012 [ 0.005 l.892 6.6 0.999 3 0.004 I 1.892 1.892 1.892 13.21 54.05 1.96 Boroflex 0.998 + 0.004 1.008 + 0.005 1.003 7 0.004 1.892 1.96 Boral 0.997 + 0.005 Average 1.0021

= _ - .___

The average bias is 0.0021 1 0.0019. The 95 percent upper limit on the KENO I keff exceeds the observed value in each case.

8.3 Refe:ence 4 Data

( A single, undermoderated 22x22 array of 4.742 percent enriched rbds with various patterns of 25 " water holes" (removed fuel rods) was tested to determine the Since the 15xl5 bundle assumed in the storage rocks I critical moderator height.

contain.21 guide / instrument tubes, th'e reference 4 data are useful to verify the l methods, particularly the homogeneous representation of the bundle.

I Three cases were calculated using KENO with explicit modeling and homogeneous modeling.

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I XN-NF-86-107, Rev. O Page 24 I

Table 8.3 Reference 4 Dato KENO-IV with Hansen-Roach Cross Sections keff keff .

Case (Explicit) (Homogeneous) 1670 0.995 + 0.0054 1.002 + 0.0055 1674 0.996 7 0.0054 1.001 7 0.0054 3 1680 1.000{0.0051 1.005 1 0.0063 3 Average 0.997 1.0027 The bias is -0.00017 1 0.0015 for all six cases.

The homogeneous model results appear to be about 0.005 higher than the explicit model results, but this bias is not significant. All results agree very well with the observed keff of unity.

8.4 Reference 5 Data The rod design here is identical to that of the reference 4 data:

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I XN-NF-86-107, Rev. O Page 25 I

I Table 8.4a Fuel Design Parameters I Rod Diameter (cm) 0.79 Enrichment (% U-235) 4.742 UO2 Density (% TD) 94.71 Fuel Length (cm) 90 Clad Aluminum Clad ID/OD (cm) 0.82/0.94 I

Four flooded 18xl8 bundles were placed in a 2x2 array spaced by various thick-nesses of various between-bundle moderators. These moderators included air, water, expanded polystyrene, polyethylene powder (low density). and polyethylene balls (higher density).

I These experiments were modeled using the SCALE system as documented in reference 6. Selected cases were modeled here for comparison.

I These experiments are useful in validating the optimum moderation calculations.

Experiments with low density moderation within and between bundles is not available.

I The three cases selected had a 10 cm spacing between bundles. This spacing was filled with either air, polystyrene, or polyethylene powder. The corresponding I hydrogen densities were 0, 0.0022, and 0.0464 gm/cc, respectively. The water densities to yield these H densities are 0,1.97 and 41.47 percent, respectively.

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XN-NF-86-107, Rev. O Page 26 I

Table 8.4b Reference 5 Data Low Density Moderation Between Bundles E KENO Results E H Density keff keff (gm/cc) (16 Group) (27 Group) 0 1.012 + 0.0046 0.985 + 0.0053 0.0022 1.012 ! 0.0059 0.0464 1.036 1 0.0045 1.024 3 0.0050 For the three 16-group cases, the bias is 0.020 1 0.008.

The results presented, agree well with the complete result set of reference 6. The 16 group (Hansen-Roach) results appear to be slightly conservative.

E 3

These results indicate that the low density moderation results are occurate or perhaps slightly conservative.

8.5 Acceptability Limit Pooling data f rom references 2, 3 and 4 (flooded cases), the average and standard deviation of the systematic bias are 0.0011 and 0.00ll, respectively. Clearly, there is no significant systematic bias. Based on the limited replication of the low l moderation cases (reference 5), and the complete results in reference 6, the bias at this condition will be conservatively set equal to that for the flooded cases.

Using the criteria of ANSl/ANS-8.17-1984 (and other similar documents), the l maxim em allowable, calculated keff is established as follows.

LIMIT = A - B - C - D l I i

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I XN-NF-86-107, Rev. O Page 27 The terms are defined below.

LIMIT: Maximum acceptable keff.

I A: Mean keff f rom appropriate benchmarks. The assigned value is I l.0011.

B: An allowance for uncertainties in parameter A. The assigned value is 0.00l I.

C: An allowance for uncertainty in kef f calculations. This allowance is variable, and is included in the final kef f result; i.e., the 95 percent upper limit statement. Therefore, the acceptability limit is not adjusted, and C is set to zero.

D: An orbitrary margin to ensure subcriticality. This is set to 0.05, except for optimum moderation conditions where the value is 0.02.

Therefore, the acceptability limit is 0.95, or 0.98 (optimum moderation).

I It should be noted that allowances B and C are of ten pooled before applying a l

l confidence level multiplier (usually about 2.0).

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in this format:

LIMIT = A - D - K * /B + C where K is the confidence level multiplier.

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XN-NF-86-107, Rev. O Page.28 I

Since the KENO standard deviation is typically 0.003-0.006, the sum of squares is g dominated by the KENO variance. The limits calculated by the two methods are -5 very close. The ANS/ ANSI format is more conservative by 0-0.0004 for typical KENO standard deviations.

8.6 Cross Section Comparisons Selected models were calculated using the 27 group ENDF/B-IV and 123 group GAM-THERMOS cross section libraries with NITAWL processing.

For the infinite array of flooded bundles on 20.875-inch centers, the k inf results are 0.906 3 0.0048 (27 group) and 0.913 1 0.0057 (123 group). These results agree well with the 0.917 1 0.005 result using the 16 group Hansen-Roach library.

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XN-NF-86-107, Rev. O Page 29 I 10.0 COMPUTER INPUT LISTINGS Typical KENO input listir.gs are provided for reference.

10.1 KENO Input: K-Infinite Case at Full Flooding This is the explicit model for an infinite array of 15xl5 bundles on 20.875-inch centers.

HB ROBINSON, 4.2% ENR, K-INF, BUNDLES ON 20.875" CENTERS, FLOODED 300.0 83 300 3 16 6 7 3 7 9 2 15 15 1 I 7 I 0 2000 00 1 0 0 0 0 0 00 0 0 6*-1 1 -92508 6.290fE-04 I i 92509 3.5867E-04 1 9281I l.0803E-02 1 92812 1.1443E-02 8100 4.6466E-02 I I 2 502 1.00 3 40100 4.2535-2 BOX l I '

CYCL FUEL ROD 1 0.452755 500.0 -500.0 CYLI 0 0.46228 500.0 -500.0 1

I CYLI 3 0.53848 500.0 -500.0 CUBO 2 0.71501 -0.71501 0.71501 -0.71501 500.0 -500.0 BOX 2

INST / GUIDE TUBE l CYLI 2 0.64897 500.0 -500.0 I CYLI 3 0.69088 500.0 -500.0 CUBO 2 0.71501 -0.71501 0.71501 -0.71501 500.0 -500.0 I

15XIS BUNDLE: 21.4503 CM SQUARE (8.445" SO)

CORE O 10.72515 -10.72515 10.72515 -10.72515 500.0 -500.0 1 CUBO 2 26.5II25 -26.5II25 26.5II25 -26.51125 500.0 -500.0 l 1 1 15 I i 15 I III O

( 2 3 13 10 363 I II 0 2 3 13 10 10 13 3 III 0 2 4 12 8 88I I Il 0 lE 2 5 ll 6 5 ll 6 III 0 15 2 6 10 4 3 13 10 III 0 l 2 881 4124 I II 9 END KENO I

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XN-NF-86-107, Rev. O Page 30 Il.0 REFERENCES (l) " SCALE: A Moduiar Code System for Performing Standardized Computer Analyses for Licensing Evaluation," NUREG/CR-0200.

(2) M. N. Baldwin, et. al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," BAW-1484-7, July 1979.

I (3) -S. R. Bierman and E. D. Clayton, " Criticality Experiments with Subcritical Clusters of 2.35 wt% and 4.31 wt% U-235 Enriched UO2 Rods in Water with Steel Reflecting Walls," NUREG/CR-1784 (PNL-3602), January 1981.

(4) J. C. Manaranche, et. al., " Critical Experiments with Lattices of 4.75% wt%

U-235 Enriched UO2 Rods in Water," ANS Trans, Vol. 28, pp. 302-303.

I (5) J. C. Manoranche, et. al., " Dissolution and Storage Experimental Program with U(4.75)O2 Rods," ANS Trans, Vol. 33, pp. 362-364.

A. M. Hathout, et. al., " SCALE System Cross Section Validation for Criticality I (6)

Safety Analysis," ANS Trans, Vol. 35, pp. 281-283.

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XN-NF-86-107, Rev. O Issue Date: 9/9/86 FINAL REPORT CRITICALITY SAFETY ANALYSIS H. B. ROBINSON SPENT FUEL STORACE RACKS (UNPOISONED, LOW DENSITY)

WITH 4.2% ENRICHED 15xlS FUEL ASSEMBLIES AUGUST 1986 I

DISTRIBUTION

, ' L. D. Cerrold C. W. Malody J. E. Pieper F. B. Skogen I. Z. Stone J. W. Hulsman T. W. Potten L. J. Federico CP&L/H. G. Shaw (SI)

Document Control (S)

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