ML20078F100

Summary of Waterford 3 Criticality Safety Analysis for Fuel Enrichments Above 4.1 W/O U-235 Taking Credit for Fixed Burnable Poisons

ML20078F100
Person / Time
Site:	Waterford
Issue date:	01/31/1995
From:	Gober T ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20078F090	List: ML20078F088 ML20078F092 ML20078F100
References
NEAD-SR-94-075., NEAD-SR-94-075.R1, NEAD-SR-94-75., NEAD-SR-94-75.R1, NUDOCS 9502010318
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Text

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-:- ~~ ENTERGY

SUMMARY

OF THE WATERFORD 3 CRITICALITY SAFETY ANALYSES FOR FUEL ENRICHMENTS ABOVE 4.1 W/O U-235 TAKING CREDIT FOR FlXED BURNABLE POISONS l

i JANUARY 1995 NEAD-SR-94/075.R1 EDC FILE OR 304-37.3 AUTHOR T.G.OBER TECHNICAL REVIEW K.B.MEGEHEE CENTRAL DESIGN ENGINEERING ENTERGY OPERATIONS, INC.

502010318 950127 TO H REQU MENTS OF D 3 0, REV DR p

ADOCK 05000382 PDR

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TABLE OF CONTENTS TABLE OF CONTENTS 2

LIST OF TABLES 3

LIST OF FIGURES 4

l 1.0

SUMMARY

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i 2.0 PIFERENCES 7

3.0 METHODOLOGY 9

3.1 Model Description..

.9 3.2 Spent Fuel Rack..

.15 3.2.1 Base Analyses....

.15 3.2.2 Tolerance Factor Analyses.

.15 3.2.3 Resultant 95/95 Probability / Confidence Reactivities.

.17 i

3.2.4 Modeling Conservatisms.

.18 3.3 Containment Temporary Storage Rack.

. 19 l

3.4 Fuel Handling Accidents.

.19 i

i 4.0 RESULTS 20 4.1 Spent Fuel Rack Results.

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4.1.1 Reactivity vs. Axial Gap Location...

. 20 4.1.2 Tolerance Results.

. 20 4.1.3 Final Reactivities.

. 20 4.2 Containment Temporary Storage Rack Results.

.21 j

4.3 Accident Results.

. 21 APPENDIX A: BENCHMARK OF SCALE 4 24 i

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TABLE OF CONTENTS NEAD-SR-94/075.R1

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LIST OF TABLES

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Table 3.1-1. WATERFORD 3 Dimensional Data.

............... 14 Table 4.3-1. Tolerances....

. 22 Table 4.3-2. Final Reactivities..

. 22 Table A.0-1. Summary of SCALE 4 Benchmark.

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h LIST OF FIGURES P

1 Figure 3.1-1. Axial Geometry...

..I1 Figure 3.1-2. Assembly Geometry..

.12 Figure 3.1-3. Radial Geometry of Analyzed Cell..

.. 13 Figure 4.3-1. KENO K-eff for Eight Coplanar Gaps...

. 23 Figure 4.3-2. Reactivity Equivalent Designs..

. 23 t

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SUMMARY

The Waterford 3 Spent Fuel Racks (SFR's) were originally designed for the storage of fuel containing up to 3.5 w/o U-235. The criticality safety analysis for these racks was later revised to support enrichmentr up to 4.1 w/o. Both of these analyses assumed no fixed poisons were present in the fuel assembly. Subsequently, a Boraflex monitoring program was proposed and approved by the NRC [ Reference 1j. This program includes periodic non-destructive testing of selected Boraflex panels and criticality review. The initial testing and analysis review were successfully completed in 1993 [ Reference 12].

With increasing cycle lengths, fuel management plans require a larger fraction of the reload batch to contain poisons. Because poison rods significantly reduce the reactivity of the fuel assembly, higher fuel enrichments may be loaded and still meet the fuel storage criticality acceptance criteria

[ Reference 11]. The Waterford 3 fuel design for poisoned assemblies contains fixed poison shims and does not permit the poison rods to be readily removed. Therefore, taking credit for the l

reactivity control effects of these poison rods is consistent with the criticality safety requirements

[ Reference 11].

l This' spent fuel pool criticality analysis demonstrates that, in the unborated SFR, a radially enrichment zoned assembly with enrichments of 4.50 and 4.10 w/o containing eight poison rods with 0.016 grams B-10 per inch (the " Base Assembly") meets the NRC acceptance criteria of 0.95

{

k-eff at the 95/95 probability / confidence level. Various other combinations of fuel enriclunents and poison loadings were also analyzed and confirmed to meet the NRC acceptance criteria. These configurations include zoned assemblies with maxunum pin enrichments up to 4.9 w/o U-235.

Acceptable reactivity was achieved at a specific enrichment by varying tne number of poison rods i

and/or concentration of poison material.

]

l This report describes the KENO Monte Carlo calculations which model the Waterford 3 SFR for zoned and shimmed assemblies containing (1) the ABB CE Guardian Grid lower grid design, (2)

)

increased enrichment above 4.1 w/o, and (3) and the presence of fixed burnable poison shims. The criticality analysis conservatively assumes that all Boraflex panels contain 4-1/2 inch coplanar gaps at the most reactive axial location (top of the panel). Uncertainties and biases due to methods and manufacturing tolerances are addressed in determining the 95/95 k-eff values for the assemblies.

The previous criticality analysis [ Reference 13] considered the effects of various accident configurations, and concluded that, when credit is taken for a minimum soluble boron concentration in the SFR water, the NRC acceptance criteria is met. The new criticality analysis demonstrates that the Base Assembly is less reactive than the assembly assumed in the presious 1.0

SUMMARY

NEAD-SR-94/075.R1

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i analyses [ Reference 13]. The SFR accident analyses remains within the acceptance criteria when -

1 potential gaps in the Boraflex panels are modeled.

The Containment Temporary Storage Racks (CTSR's) rely on assembly spacing to control reactivity. The Base Assembly design was determined to be less reactive than the assembly evaluated in the Reference 13 analysis. Therefore, the previous criticality analysis remains bounding for the Base Assembly design when placed in the CTSR under both normal and accident conditions.

The analysis of the new fuel storage racks is not addressed in this report.

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1.0

SUMMARY

NEAD-SR-94/075.R1 6

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o 2,0 REFERENCES 1.

Letter: W3192-0396 A4.05 QA, "Waterford 3 SES, Docket No. 50-382, License No. NPF-38, Boraflex Surveillance Program", Burski to US Nuclear Regulatory Commission, December 23,1992.

2.

Report: "Boraflex Test Results and Evaluation ", EPRI TR-101986, Project 2813-04, February,1993.

3.

"CASMO-3 A Fuel Assembly Burnup Program, Version 4.7", STUDSVIK/NFA-89/3,Rev.2, March,1992.

4.

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

5.

Bierman, S. R., Clayton, E. D., Durst, B. M., " Critical Separation Between Subcritical Clusters of 4.29 w/o U-235 Enriched UO2 Rods in Water with Fixed Neutron Poisons",

NUREG/CR-0073, Batelle PNW Laboratories, May,1978.

6.

Marshall, W.,et al, " Criticality Safety Criteria", ANS Trans:35,278-279 (1980) 7.

Book: " Practical Non-Parametric Statistics", W. J. Conover, 1971, John Wiley and Sons 8.

" SCALE 4 - A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation", CCC-545, NUREG/CR-0200 REV.4 (ORN11NUREG/CSD-2/R4)

Vols.1, II, and III, ORNL, February,1990.

9.

Bierman, S. R., Clayton, E. D., Durst, B. M., " Critical Separation Between Suberitical Clusters of 2.35 w/o U-235 Enriched UO2 Rods in Water with Fixed Neutron Poisons," PNL-2438, Batelle PNW Laboratories, October,1977.

10. Tumer, S. E, Gurley, M. K., " Evaluation of AMPX-KENO Benchmark Calculations for High-Density Spent Fuel Storage Racks",(Southern Science Applications,Inc.), NSE:80,230-237 (1982).

I1. Grimes, Brian K., "U.S. NRC Letter to All Power Reactor Licensees", April 14,1978, Docket No. 50-289.

12. " Revised Boraflex Surveillance Program of the Spent Fuel Pool Racks, Waterford Steam Electric Station, Unit 3", TAC No. M86006, Docket No. 50-382, August 23,1993.

2.0 REFERENCES

NEAD-SR-94/075.R1

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13. "WSES-3 Fuel Storage Racks Upgrade for the Storage of 4.1 Weight Percent U-235 i

Assemblies", M. R. Eastburn, Middle South Senices,Inc., File:304-37, June,1986.

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2.0 REFERENCES

NEAD-SR-94/075.R1

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' s lt : 4 3.0 METHODOLOGY 3.1 Model Description ne Waterford 3 spent fuel rack was designed by Wachter Associates, Inc.. The rack is composed of cells containing stainless steel partitioas which provide a fuel assembly storage area and two poison insert areas per cell. The cells are oriented so that face adjacent assemblies are separated by a poison insert area. Poison inserts contain two Boraflex panels which are encapsulated in stainless steel cladding. The Boraflex panels are restricted from movement by indentations in the clad which apply continuous pressure over the entire length of the panel. The panels are arranged in a rectangular configuration which maintains a 1 inch flux trap between panels. This configuration is illustrated by Figure 3.1-3.

The panels were designed to be positioned approximately axially symmetric with the centerline of the active fuel column leaving a small portion of the fuel uncovered at the top and bottom of each panel. The fuel assembly design analyzed in this report has a slightly elevated fuel cc'- m because of the use of the debris resistant ABB CE Guardian Grid lower grid design. See Figure 3.1-1 for the relative axial positioning of the active fuel, shim, and Boraflex regions.

Waterford 3 uses an ABB CE lox 16 fuel assembly design with five large water holes for insertion of a control element assembly. The Base Assembly used in this analysis has an enrichment distribution containing 176 fuel pins at 4.50 w/o and 52 fuel pins at 4.10 w/o. The lower enrichment pins minimize power peakmg around the five water holes and in the corners of the assembly. The position of the eight B4C burnable poison rods is shown in Figure 3.1-2. Other configurations of poison rods considered in this evaluation are also shown in that figure.

The analysis assumes that all fresh fuel assemblies with a maximum enrichment above 4.1 w/o U-235 contain boron shims which displace fuel rods. Neutron absorption by U-234 is credited and a bounding fuel stack density is used. The Boraflex is modeled using the nummum design dimensions and the 95/95 lower limit of the assayed B-10 loading. The panel width is also assumed to shrink 4.1 %, consistent with Reference 2. The panel height is reduced by the gap size as described below. Table 3.1-1 shows the key geometric parameters used in the base modeling of the fuel assembly and SFR. The assembly is positioned radially synunetric in the SFR cells which was determined to be the most reactive. The spent fuel pool water is modeled at maxunum density and does not contain any soluble boron.

Formation of gaps occur when the stress in a restrained panel reaches the shearing strength of the Boraflex. A gap has three characteristics which effect the reactivity ofthe system; (1) gap size, 3.0 METHODOLOGY NEAD-SR-94/075.R1 9

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e (2) axial position of the gaps, and (3) configuration of the gaps. Larger gap sizes result in higher system reactivity because the surface area of the absorber is reduced. The reactivity effect of axial position is dependent on two factors: (1) the interaction of a gap and the end of panel region, and (2) coupling of gaps in the same flux trap. If both panels in a flux trap form gaps, but the gaps occur at different axial locations, then the reactivity impact will be much lower than if the gaps are formed at the same location. The criticality analysis summarized in this report conservatively assumes that all Boraflex panels contain 4-1/2 inch coplanar gaps at the top of all of the panels (the most reactive axial position). This gap size bounds the Waterford 3 measurements described in Reference 1 and is consistent with the maximum gap size reported in Reference 2. This approach is very conservative relative to more realistic analyses which would include the effect of the variations in gap size and location based on probability of occurrence distnbutions. Inclusion of these effects would be expected to reduce the upper limit of the rack k-effective by 1-2% delta k.

3.0 METHODOLOGY NEAD-SR-94/075.R1 10

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Figure 3.1-1.

Axial Geometry i

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Notes: The shim region of the poison rods is axially symmetric relative to the active fuel region.

t The Boraflex Panels (without gaps) are shifted approximately 1 inch down relative to the center of active fuel. The top of the panel is less than 1 inch above the top of the shim region of the poison rods.

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3.0 METHODOLOGY NEAD-SR-94/075.R1 11

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.o Figure 3.1-2.

Assembly Geometry 4 5HIM 8 5HIM 12 $HIM Ytv i iI l II I II MM

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Water Rod

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3.0 METHODOLOGY NEAD-SR-94/075.R1 12

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Figure 3.1-3.

Radial Geometry of Analyzed Cell

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Table 3.1-1.

WATERFORD 3 Dimensional Data f

Parameter Design Specification Analyzed Fuel Pellet diameter 0.325" 0.325"

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Fuel Clad 1.D.

0.332" 0.332" l

Fuel Clad 0.D.

0.382" 0.382" Fuel Stack Density 10.06 gm/cc 10.2 gm/cc j

Fuel Enrichments 4.05/3.6510.005 %

up to 4.90/4.50 w/o l

Rod Pitch 0.506" 0.506" l

Boraflex Panel Thickness 0.100+ 0.0-0.020" 0.080" l

Boraflex Panel Height 139.0 0.5" 138.32" Boraflex Panel Width 7.2+0.0-0.2" 6.713" (4.1% shrinkage)

Boraflex Areal Density 0.02299 gm B-10/cm2 0.02259 gm B-10/cm2 l

Insert Cladding 'Ihickness 0.031" 0.031" i

insert Width 7.31" 7.31"

{

Flux Trap Thickness 1.0+0.1-0.0" 1.0" Panel Centerline from SFR lower plate 79.7" 79.7" Poison Area Thickness 1.57+0.02-0.0" 1.57" l

Poison Area Width 8.5910.03" 8.59" l

SFR Stock Thickness 0.093 0.005" 0.093" SFR Cell 1.D, 10.164+0.060-0.0" 10.164" i

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3.2 Spent Fuel Rack The analytical methods to determine SFR reactivity use the computer codes SCALE 4

[ Reference 8] and CASMC 3 [ Reference 3]. Both codes have been widely used for criticality analysis throughout the nuclear industry. While the SCALE code has been extensively benchmarked by the industry, specific benchmark comparisons to critical experiments were performed to establish bias and uncertainty factors. The details and results of this benclunarking are described in Appendix A of this report.

3.2.1 Base Analyses As noted in the previous section of this report, the SCALE analyses conservatively assume that 4-1/2 inch gaps occur at the top of every panel. The SCALE calculations are nm for 210 generations with 3000 histories per generation. As the first ten generations are skipped in these analyses, a total of 600,000 histories are analyzed.

3.2.2 Tolerance Faetor Analyses Uncertainties or tolerance facters in the rack and fuel design parameters are evaluated by either (1) setting the parameter to its most conservative value, or (2) performing sensitivity studies to determine the reactivity impact of the tolerance factor. These sensitivity studies were performed using the CASMO model [ Reference 3].

The following tolerance factors were analyzed:

1, Higher fuel enrichments (nominal + 0.05 w/o U-235), calculated for increased fuel loading;

2. Higher fuel pellet density, calculated for increased fuel loading;

3. Higher form factor, calculated for increased fuel loading;

)

4. Larger fuel pellet diameter, calculated for increased fuel loading;

5. Larger clad I.D., calculated for reduced absorption by the fuel cladding;

6. Smaller clad O.D., calculated for increased neutron moderation in the fuel assembly;

7. Minimum guide tube thickness, calculated for increased neutron moderation in the fuel assembly;

8. Larger fuel pit area, calculated for increased neutron moderation around the fuel assembly; 3.0 METHODOLOGY NEAD-SR-94/075.R1 15

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9. Smaller shim lot loading, calculated for reduced neutron absorption in shims due to uncertainties in the average loading of a shim lot;

10. Smaller shim pellet loading, calculated for reduced neutron absorption in shims due uncertainties in the individual pellet loading relative to the average of a lot.

The base CASMO k-inf was subtracted from the calculated CASMO k-infs from cases I through i

10 above to obtain the delta-k values for each of the tolerances. The tolerance factors were then used to obtain the combined fabrication tolerance factors for each of the assembly designs. As these effects are independent, they were statistically combined by taking the square root of the sum of the squares ofeach contributing uncertainty due to the tolerances identified above.

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3.2.3 Resultant 95/95 Probability / Confidence Reactivities The SCALE calculated eigenvalue for each analyzed geometry was combined with the appropriate tolerance uncertainties and method uncertainties as follows:

keff-95/95 = kSCALE + Aktolerances + Akenrichment + Akmethod +K*& otal t

where:

Aktolerances

= overall tolerance uncertainty Akenrichment

= 1.0 -(1.001204 - 0.001894*w/o) from Appendix A Akmethod

= 0.00072 from Appendix A 2

2 6 total

=(USCALE + amethod ) %

f =[U 4/

4/

total (USCALE (210-10-1) + amethod (21-1)))-2 4/

Note: This formulation was taken from Reference 7.

Note: 210 generations (skipping 10 generations) and 21 critical experiments were analyzed.

K = 95/95 tolerance factor for f degrees of freedom I

1 3.0 METHODOLOGY NEAD-SR-94/075.R1 17 a

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3.2.4 Modeling Conservatisms The following modeling conservatisms have been identified:

1. The nominal Boraflex dimensions used are the muumum as-designed values with an additional

(

conservative 4.1 % width shnnkage and a 4-1/2 inch gap assumed at the top of each panel.

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2. Worst case geometry (symmetric) is used for the positioning of the fuel assembly in the SFR i

pit.

3. Water is set to the maximum density of I gm/cc.

)

4. The infmite lattice Dancoff factor is used.

j

5. No credit is'taken for fuel bumup.

6. No credit is taken for the soluble boron in the water (except for accidents).

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7. Much structural steel is not modeled in the problem.

8, No neutron leakage exists in the x-y directions.

3.0 METHODOLOGY NEAD-SR-94/075.R1 + 18 L

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3.3 Containment Temporary Storage Rack l

The containment temporary storage rack (CTSR) relics upon assembly spacing to control reactivity for normal operations and credits the presence of soluble boron for accident conditions. The rack was previously analyzed as documented in Reference 13. The continued applicability of that analysis was evaluated by comparing the reactivity of the assembly design assumed in that analysis to the Base Assembly design, using the SCALE code for both designs. Two assemblies, using the Base Asseinbly design, were modeled assuming pure maximum density water with separation i

distances ranging from 1.762 to 12 inches. This range includes state points consistent with nominal operating configurations and the nummum separation distance for a postulated fuel assembly handling accident. The same calculations were repeated using the Reference 13 design.

P 3.4 Fuel Handling Accidents The Reference 13 criticality analysis considered the effects of various fuel handling accident configurations. This analysis demonstrated significant margin to the 0.95 acceptance criteria when credit for boron was considered. The applicability of that analysis was evaluated by comparison of the reactivity of the Base Assembly design relative to that assumed in the Reference 13 analysis.

This evaluation considered the effects of variations in separation distance as described in i

Section 3.3 above and the effects due to the presence of the strong absorbers. It also considered the impact of Boraflex gaps on the spent fuel pool accident analysis. These evaluations were performed using the SCALE code.

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t 4.0 RESULTS t

4.1 Spent Fuel Rack Results 4.1.1 Reactivity vs. Axial Gap Location An evaluation was performed to determine the most reactive axial location to place the coplanar gaps in the Boraflex. The results of placing coplanar gaps at various positions for the Base Assembly (4.5/4.1 w/o zoned fuel with 8 shirns at 0.016 grams B-10/ inch) are shown in Figure 4.3-1. As can be seen, the most reactive placement was found to be at the top of the panel.

4.1.2 Tolerance Results As described in Section 3.2.2, the base CASMO reactivities were subtracted from the calculated CASMO reactivities for the off-nominal cases to obtain the delta-k values for each of the tolerances. The tolerance factors were then used to obtain the combined fabrication tolerance factors for each of the assembly designs. As the tolerances are independent, the statistically combined tolerance factors (Table 4.3-1) were obtained by taking the square root of the sum of the i

squares of each contributing uncertainty due to the tolerances identified above.

4.1.3 Final Reactivities i

The raw KENO eigenvalues and uncertainties were combined with the tolerance uncertainties and the methods uncertainties to obtain the final 95/95 values for k-eff. The components and the final 95/95 k-effs are given in Table 4.3-2. These results demonstrate that the k-effective is less than 0.950, including uncertainties and biases, at the 95/95 probability / confidence level. Therefore, the acceptance criteria of k-effective below 0.95 is met for the Base Assembly and other equivalent reactivity designs. A plot of the reactivity equivalent designs, i.e., those combinations of fuel enrichment and shim loadings which give final reactivities less than 0.95, is given in Figure 4.3 2.

4.0 RESULTS NEAD-SR-94/075.R1 20

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- 4.2 containmen: Temporary Storage Rack Restets The Reference 13 assembly design was found to bound the Base Assembly design for all configurations. Since the assessment considers tightly coupled configurations equivalent to accident conditions and loosely coupled configurations equivalent to normal operating conditions the Reference 13 results are bounding. Those results demonstrate that k-effective is less than 0.899 for normal operations and less than 0.90 for accident conditions including uncertainties and biases at the 05/95 probability / confidence level. Therefore, the acceptance criteria of k-effective below 0.95 is met for the Base Assembly and other equivalent reactivity designs.

4.3 Accident Results

'The reactivity of the Reference 13 assembly design was found to bound the reactivity of the Base Assembly design. The positive reactivity impact due to the presence of gaps was conservatively determined to be 0.05 delta-k. When this value added to the Reference 13 analysis results, the k-effective is less than 0.91, including uncertainties, when credit for soluble boron is inchded.

Therefore, the acceptance criteria of k effective below 0.95 is met for the Base Assembly and other equivalent reactivity designs.

i 4.0 RESULTS NEAD-SR-94/075.R1 21

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Table 4.3-1.

Tolerances FUEL STD STD STD STD STD STD STD STD STD STD STD INRfCH.

4.11 4.2 4.24 4.33 4.37 4.5 4.55 4.01 4.05 4.85 4.9

1. 3H'MS 4

4 4

4 8

0 0

0 0

12 10 0101N.

0.012 0.010 0.02 0.024 0.012 0.010 0.02 0.024 0.020 0.02 0.012 DELT4 1

0.00241 0.00234 0.00232 0.00225 0.00223 0.00215 0'4 0.00208 0.00204 0.00203 0.00202 2

0.00057 0.00057 0.00058 0.00058 0.00054 0.00059 b.Utal59 0.00050 0.00059 0.00064 0.00065 3

0.00153 0.00153 0.00155 0.00153 0.00152 0.00158 0.00150 0.00157 0.00157 0.00169 0.00172 4

0.00038 0.00038 0.00039 0.00039 0.00035 0.00039 0.00040 0.00040 0.00039 0.00043 0.00045 5

0.00001 0.00000 0.00002 0.00001 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 0

0.00238 0.00238 0.00247 0.00236 0.00234 0.00240 0.00242 0.00242 0.00243 0.00234 0.00232 7

0.00095 0.00090 0.00103 0.00090 0.00094 0.00005 0.00004 0.00099 0.00100 0.00108 0.00105 8

0.00100 0.00005 0.00104 0.00104 0.00084 0.00083 0.00082 0.00080 0.00079 0.00086 0.L0062 9

0.00185 0.00157 0.00163 0.00150 0.00200 0.00203 0.00202 0.00195 0.00194 0.00299 0.00361 10 0.00003 0.00002 0.00099 0.00000 0.00116 0.00120 0.00118 0.00113 0.00113 0.00175 0.00213 RMS 0.00454 0.00436 0.00449 0.00429 0.00448 0.00452 0.00449 0.00446 0.00444 0.00519 0.00587 e

Note: DELTA = delta k, tolerance - nominal Table 4.3-2.

Final Reactivities

1. 5HIMS 4

4 4

4 8

8 8

8 8

12 16 CM O 10!!N 0.012 0.016 0.02 0.024 0.012 0.016 0.02 0.024 0.028 0.02 0.012 i

ENRfCHMENT 4.11!3.71 4.20t3.80 4.24I3.84 4.33/3.93 4.37/3.97 4.50/4.10 4.55/4.15 4.6114.21 4.65/4.25 4.85i4.45 4.90/4.50 KENOKEFF 0.92904 0.92958 0.93141 0.93285 0.93004 0.53290 0.93033 0.93109 0.93065 0.93148 0.82731 OtASES METHOD 0.00072 0.00072 0.00072 0.00072 0.00072 0.00072 0.00072 0.00072 0.00072 0.00072 0.00072 ENRICH.

0.00658 0.00075 0.00683 0.00700 0.00707 0.00732 0.00741 0.00753 0.00760 0.00798 0.00808 UNCERT.

i KENO 0.00109 0.00120 0.00124 0.00130 0.00118 0.00120 0.00122 0.00116 0.00125 0.00123 0.00135 METHOD 0.00156 0.00156 0.00156 0.00150 0.00156 0.00150 0.00156 0.00156 0.00156 0.00156 0.00156 TOLER 0.00454 0.00436 0.00449 0.00429 0.00448 0.00452 0.00449 0.00446 0.00444 0.00519 0.00567 TOTAL 0.00190 0.00197 0.00100 0.00205 0.00196 0.00157 0.00198 0.00194 0.00200 0.00100 0.00206 f

40.980 47.133 48.917 54 864 46.250 47.133 48.022 44.408 49.893 48.979 55.569 k

2.11802 2.07978 2.07016 2.04027 2004734 2.07970 2.07492 2.00557 2.06514 2.08083 2.03691 k* TOTAL 0.00402 0.00410 0.00412 0.00418 0.00409 0.00410 0.00411 0.00407 0.00413 0.00412 0.00420 85l95 K EFF 0.94490 0.94551 0.94757 0.94904 0.94640 0.54950 0.94706 0.94787 0.94754 0.94949 0.94598 i

i 4.0 RESULTS NEAD-SR-94/075.R1 22 L

i.

6

. O.

Figure 4.3-1.

KENO K-eff for Eight Coplanar Gaps 0.93 q 0.92 -

i w

0.91 - - 4, 0.9 -

0.89 l

0 20 40 60 80 INCHES FROM TOP Figure 4.3-2.

Reactivity Equivalent Designs o 4.9 g

+4 Sgy3 iM

Mi. >

b 4.7 ~1 8 SHIMS F#@i

> YSs Z

--*- 12 SHIMS M

g" "n

-u-m 4'6 -4 E

--+-16 SHMS 7

. sj-:

..c 4.4 g

m

/

.e

u. 4.1,f

< :aa -

• y, 0.012 0.016 0.02 0.024 0.028 SHIM LOADING, G B-10/IN.

4.0 RESULTS NEAD SR-94/075.R1 = 23

(

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-l APPENDIX A: BENCHMARK OF SC ' E4 A benchmarking of the SCALE 4 system of codes against critical experiments is performed in i

order to qualify and validate its use for performing fuel storage rack criticality safety analyses at Entergy Operations, Inc. plants.

l i

'fhe benchmarked SCALE 4 codes includes the CSAS25 Control Sequence of the ORNL i

Criticality Safety Analysis Sequence No. 4 (CSAS4). The functional modules sequentially j

executed by the CSAS25 control module are the BONAMI-S code, the NITAWL-S code, and the l

KENO V.a code [ Reference 8].

{

i A total of twenty-one critical experiments are used to benchmark these codes. These criticals are i

selected from a list of seventy-fivr. critical experiments conducted by Babcock and Wilcox and i

Pacific Northwest Laboratory [ Reference 4, Reference 5, and Reference 9]. The twenty-one criticals are chosen because of their fuel characteristics, lattice geometry's water gap spacing ad materials are re.-

r. ably representative of those found in Entergy Operations, Inc. fuel stcrage rack _

arrays.

An evaluation of the benchmarking results identifies two significant biases in the 5 CALE i

calculated k-effs. The first is an observed trend toward over-prediction of reactivi:y and increasing Boron loading in the Boral plates for four of the Babcock and Wilcox e )res, as reported also in Reference 10. These four cases are corrected first for the Boron loading bias before fmther l

data reduction is performed. For conservatism, this credit is not taken in analyzing fuel storage j

rack.

The second observed bias is an under-prediction of reactivity with increasing fuel enrichment as follows:

Enrichment Bias = 1.0 -(1.001204 - 0.001894 * (Enr. in wt% U-235))

i Statistical analysis of the 21 calculated k-effs, using Criterion 2 [ Reference 6], gives an enrichment corrected mean k-eff of 0.99915 with a Monte Carlo uncertainty of10.00305, a mMod bias of 0.00072, and a method uncertainty ofio.00156. Table A.0-1 is a summary of the benchmark statistics.

APPENDIX A: BENCHMARK OF SCALE 4 NEAD-SR-94/075.R1 + 24 k-

f 1

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i Table A.0-1.

Summary of SCALE 4 Benchmark I

l l

~

No.of Neutrons Per Case No.

Case Nome Innehment wt%

SCALE Keff Stenderd Corrected Keff No. of gan U235 Deviation Gen's I

1 BM 2.46 0.00000 0.00230 0.00025 -

. 100 500 i

~

'2 BMI

- 2.46 0.00452 0.00205 -

0.00607 100 500 f

3 OWXI 2.48 0.00324 0.00245 0.00500 100

- 500 4

BM 2.40 1.01005 0.00335 1.00004

'100

' 500 5

OWXN 2.44 1.00550 0.00203 0.00035 100 500 0

Bml 2.48 0.00050 0.00200 1.00431 100 500-(

7 BWXIX 2.46 0.00005 0.00282 0.00010 100 500 0_

PNL5 2.35 0.00100 0.00200 0.00410 100 500 O

PNL26 2.35 1.00232 0.00207 1.00451 100 500 10 PNL20 2.35 0.00007 0.00203 1.00026 100 500 11 PNL32 2.35 0.00242 0.00325 0.00461 100 500 I

12 PNL33 2.35 1.00164 0.00337 1.00303 100

-500 i

13 PNL30 2.35 0.00032 0.00265 1.00051 100 500 14 PNL30 2.35 0.00364 0.00330 0.00503 100 500

{

15 PNL0 4.20 0.00003 0.00323 1.00340 100 500 16 PNLG 4.20 0.00640 0.00321 0.00326 100 500 -

17 PNL10R 4.20 0.00441 0.00330 1.00127 100 500' to PNL11 4.20 0.00400 0.00364 1.00108 100 500 10 PNL12 4.20 0.00321 0.00344 1.00007 100 500 20 PNL13 4.20 0.00030 0.00207 0.00725 100 500 21 PNL32 4.20 0.00587 0.00320 1.00273 100' 500 i

Asy. CorpusessfKedP = Amff Messe ferde Assareenery = AN7N2 NuebentDion = A8N721 Moe6stmusereenery= AW16N v

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APPENDIX A: BENCHMARK OF SCALE 4 NEAD-SR-94/075.R1 25 I

+

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D 4

9 NPF-38-163 ATTACHMENT IV e

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I ea' 'j UNITED STATES j

NUCLEAR REGULATORY COMMISSION i

j, t,,{*. c,wj' wAsMiworow, o c. nosas. coo, August 23, 1993

~

Docket No. 50-382 Mr. Ross P. Barkhurst AllC 27 W Vice President Operations ILN:Ab-ClEO

$1fe]o K111ona, Louisiana 70066 Deer Mr.- Barkhurst:

SUBJECT:

REVISED BORAFLEX SURVEILLANCE PROGRAM 0F THE SPENT FUEL POOL RACKS WATERFORD STEAM ELECTRIC STATION, UNIT 3 (TAC N0. M86006)

By letter dated December 31, 1992, the staff of Entergy, Inc. (the current Itcensee, formerly Louisiana Power and Light ;LPL)), submitted a revised surveillance program for monitoring the Boraf ex panels contained in the spent fuel racks at the Waterford Steam Electric Station, Unit No. 3 (Waterford 3).

It had been determined earlier that use of coupons might not be adequate to monitor the extent of gamma radiation induced gaps in spent fuel pool 8eraflex panels. Consequently, in December 1987, LPR proposed an alternative surveillance program for the Boralflex panels in lieu of the previous Boraflex coupon surveillance program. The new surveillance proriram, as documented in Section D of the attachment to the December 31, 1992, etter, includes the following steps:

Gamma exposure tracking of the spent fuel pool Boraflex panels.

Periodic nondestructive testing, at a maximum 4-year interval time, of selected Soraflex panels. The testing uses neutron attenuation (blackness testing) as a means of detecting gaps (discontinuities) along the Boraflex panels. The Itcensee's nondestructive testing data of selected Boralflex panels in November 1992 will be used to provide a baseline for the trending of gap forsation in the panels.

Sixteen _ spent fuel pool exposure cells with freshly disi:ha ed fuel in them to serve as samples for trending the effects of hig pamma i

radiation exposurs on Boraflex panels. These 16 exposure cel s will be tested as described above and are expected to lead the spent fuel storage racks in gap formation.

Periodic destructive testing on selected panels if engineering assessment deterstnes it is necessary.

Monitoring and comparisons to industry developments to determine the latest methods for testing and monitoring Boraflex performance.

P-A. f A f l d~

I" QV( /.s i

i

Mr. Ross P. Barkhurst,

We have reviewed the alternative Boralflex surveillance program and find that the program provides an acceptable means of monitoring the integrity of the Boraflex panels used in the construction of the Waterford 3 spent fuel pool storage racks. The baseline surveillance results in the submittal indicate that the Boraflex panels in the spent fuel pool storage racks are currently capable of performing their intended safety function.

The Entergy submittal of December 31, 1992, indicates that you have met your prior comitments to (1) terminate the Boraflex coupon surveillance program, i

(2) establish and maintain a database of the calculated Boraflex panel gama exposures in the spent fuel storage racks, (3) submit nondestructive test results on several Boraflex panels, and (4) establish a surveillance program as discussed above. We agree that you have completed the first three commitments and that you will continue to have a surveillance program as necessary to meet the stated objectives.

If you have any questions on this matter, please let me know. This completes the action under TAC No. M86006.

Sincerely, o

n 1

-w David L.

igginton, Senior Project Manager Project Cirectorate IV-1 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation cc:

See next page l

gn---e

-nm

,me-

Mr. Ross P. Barkhurst Entergy Operations, Inc.

Waterford 3 Cc:

Mr. Hall Bohlinger, Administrator Regional Administrator, Region IV Radiation Protection Division U.S. Nuclear Regulatory Commission Office of Air Quality and Nuclear Energy 611 Ryan Plaza Drive, Suite 1000 Post Office Box 82135 Arlington, Texas 76011 Baton Rouge, Louisiana 70884-2135 Resident Inspector /Waterford NPS Mr. John R. McGaha Post Office Sox 822 Vice President, Operations Killona, Louisiana 70066 Support Entergy Operations, Inc.

Parish President Council P. O. Box 31995 St. Charl'as Parish Jackson, Mississippi 39286 P. O. Box 302 l

Hahnville, Louisiana 70057 i

William A. Cross Bethesda Licensing Office Mr. Harry W. Keiser, Executive Vice-3 Metro Center President and Chief Operating Officer Suite 610 Entergy Operations, Inc.

Bethesda, Maryland 20814 P. 0. Box 31995 Jacksod, Mississippi 39286-1995 Mr. Robert B. McGehee Wise, Carter, Child & Caraway Chainsan P.O. Box 651 Louisiana Public Service Commission Jackson, Mississippi 39205 One American Place, Suite 1630 Baton Rouge, Loutstana 70825-1697

$r. D. F. Packer General Manager Plant Operations Mr. R. F. Burski, Director Entergy Operations, Inc.

Nuclear Safety P. O. Box B Entergy Operations, Inc.

Killona, Louisiana 70066 P. O. Box B K111ona, Louisiana 70066 Mr. L. W. Laughlin, Licensing Manager Entergy Operations, Inc.

P. O. Box B Killona, Louisiana 70066 Winston & Strawn Attn:

N. S. Reynolds 1400 L Street, N.W.

Washington, DC 20005-3502

l 4s e

a i

i NPF-38-163 ATTACHMENT V l

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W3F192-0396 A4.05 QA December 31, 1992 U.S. Nuclear Regulatory Comission Attn: Document Control Desk Washington, DC 20555 i

Subject:

Waterford 3 SES Docket No. 50-382 License No. NPF-38 Boraflex Surveillance Program Gentlemen:

i Waterford 3 comitted in December 1987 to propose to the NRC. by January 1, 1993 an appropriate surveillance program for surveillance of Boraflex in the spsr.t fuel storage racks.

The purpose of this letter is to provide b the NRC the information which satisfies this comitment including a discussion of the detailed commitments and the description of the proposed surveillance program.

i This information is provided in the attachment to this letter.

Please contact me or Robert J. Murillo should there be any questions regarding this letter.

Very truly yours, s

/

g, ir or, uclear Safety RFB/RJN/dc Attachment cc:

J.L. Milhoan, NRC Region IV D.L. Wigginton, NRC-NRR R.B. McGehee N.S. Reynolds NRC Resident Inspectors Office l

.) O f

/CU O

)

.p.

I Attachment to letter W3F192-0396 A.

Backoround The pertinent information regarding the Boraflex surveillance program was documented in the following documents:

Louisiana Power and Light (LP&L) letter W3P87-2055 dated 9/16/87, LP&L Letter W3P87-2527 dated 12/15/87, and NRC Safety Evaluation Report (SER) dated 12/21/87. These documents established NRC and Waterford 3 agreement for the following commitments:

1.

The current Boraflex coupon surveillance commitment will not be performed.

2.

Waterford 3 will develop a log to track the gamma dose buildup in the spent fuel storage racks (SFSR).

3.

Waterford 3 will provide the NRC by January 1,1993 with actual data, from nondestructive techniques, on several Boraflex panels to verify that the poison material has not been unacceptably degraded by the formation of gaps.

4.

Waterford 3 will provide the NRC by January 1, 1993.with a surveillance program to verify the effectiveness of the Boraflex poison material in the Waterford 3 spent fuel storage racks. This surveillance program will be based on the latest industry developments on Boraflex surveillance methods and techniques as l

well as studies on Boraflex degradation mechanisms from radiation exposure.

4 B.

Commitment Resolution The resolution for each of the foregoing commitments is the following:

1 1.

The Boraflex coupon surveillance program was terminated and is not active.

2.

Waterford 3 maintains a database of the calculated Boraflex panel gamma exposures in the Waterford 3 SFSR.

3.

Testing of Boraflex panels was performea in November,1992.

A review of the test results indicates that the Boraflex continues to perform its intended neutron attenuation function. Additional information is provided in section C of this attachment.

4.

The proposed surveillance program has been developed, and it is described in section D of this attachment.

]

l 1

1, Attachment To letter W3F192-0396 l

C.

Testino of Boraflex panels

~

Testing of a representative sample of Boraflex panels was performed in November, 1992.

A review of the test results indicates that the Boraflex continues to perform its intended neutron attenuation function.

r Testing was conducted to test the highest exposed panels for gaps. The test results are therefore bounding on expected gap formation in the r

Boraflex.

The following is a summary of the test results:

Number of Panels Tested 697

=

Total Number Of Gaps 185

=

Number of Panels With No Gap 538

=

Average Exposure 8.03E+09 rads

=

Maximum Exposure 1.66E+10 rads

=

136 Number of Panels With 1 Gap

=

Average Exposure 8.93E+C9 rads i

=

Maximus Fxposure-1.66E+10 rads

=

Average Gap Size (A&B) 2.17 inches

=

3.57 inches j

Maximum Gap Size (A&B)

=

Number of Panels With 2 Gaps 20 Average Exposure 9.00E+09 rads

=

Maximus Exposure 1.62E+10 rads

=

Average Gap Size (A&B) 2.04 inches l

=

Maximum Gap Size (A&B) 2.74 inches

=

i Number of Panels With 3 Gaps 3

=

Average Exposure 1.01E+10 rads

=

Maximus Exposure

-1.31E+10 rads Average Gap Size (A&B) 1.05 inches

=

Maximum Gap Size (A48) 1.24 inches

=

The distribution of gaps size indicates that the Boraflex is performing as expected based on EPRI/ industry data. EPRI data indicates that after approximately 7x10E+09 rads, shrinkage, and gap formation, of the panels reaches a maximum value.

We expect that the locations with fuel currently stored will have a much lower number of gaps, due to being in lower exposed areas.

2

(. ' ^

• Attachment To letter W3F192-0396 D.

Descriotion of Boraflex Surveillance Procram Obiective The objectives of the Boraflex surveillance program are:

1.

To verify the effectiveness of the Boraflex poison material in the SFSR.

2.

To ensure the* requirements of Technical Specification 5.6.1 are met. These requirements are:

The spent fuel storage racks are designed and shall be maintained with:

a.

A k.,, equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance for uncertainties.

b.

A nominal 10.38 center-to-center distance between fuel assemblies placed in the spent fuel storage racks.

3.

To ascertain the rate of change (gap formation) in the Boraflex and determine the interval between surveillance >.

Methodoloov

~

The Boraflex surveillance program at Waterford 3 will consist of the following:

1.

Tracking of the gamma exposure of the Boraflex panels.

2.

Periodic nondestructive testing (" blackness" testing) of selected Boraflex panels to determine the extent of gap formation.

The proposed schedule for the blackness testing is four (4) year intervals but no later than 12/31/94 for the next testing with the exact schedule being determined by the rate of gap formation.

The results of the nondestructive testing will be compared to industry data and EPRI sponsored research. The Waterford 3 SFSR criticality analysis will be reviewed by April 15, 1993 based on EPRI data for maximum exnected gap formation. The review will be performed by modeling the current status.of the Boraflex panels and projecting the expected gap formation due to calculated exposure through 12/31/94. Data collected in November,1992 will provide a baseline for the trending of gap formation.

3

,. v. ~ -

Attachment To letter W3F192-0396 i

Waterford 3 will have lead exposure cells in the SFSR to ensure that degradation due to gamma exposure will be identified early.

These 16 cells will have freshly discharged fuel loaded into them following refueling outages, and these cells will be tested during each nondestructive test. The configuration of these cells in the SFSR is shown in Figure One (1).

FIGURE ONE l

i l

D Peripheral boundary no Boraflex l

k.

l C

- ""='l l

B

- Normal Boraflex panel i

l l

l l

A 29 30 31 32 Since these locations contain some of the leading exposure panels, fuel will not be placed into these cells until the review of the criticality analysis is complete. These lead exposure cells will provide data for the performance of Boraflex under gamma exposure and early indication of unacceptable trends.

These cells are expected to lead the SFSR in gap formation.

3.

Periodic destructive testing of selected panels if engineering assessment determines a need for destructive testing.

The need for destructive testing will be determined by a review of nondestructive test'results, trend data, and industry experience.

4.

Monitoring of industry developments to determine the latest methods for testing Boraflex and to determine Boraflex performance at other sites.

I 4

i

,,- ~

Attachment To letter W3F192-0396 Summary The Boraflex surveillance program will provide sufficient data to verify the effectiveness of the Boraflex poison material and to review the SFSR criticality analysis. Additionally, it will provide trending data for changes in the Botaflex with gamma exposure and thus provide a technical basis for taking early corrective action if problems arise.

The-surveillance program satisfies the program objectives and provides assurance that the margin of safety in the SFSR will be maintained.

/

5

'.*(**

W3F192-0396 i

bec:

D.C. Hintz R.P. Barkhurst F.J. Drummond t

0.F. Packer R.G. Azzarello R.S. Starkey T.R. Leonard L.W. Laughlin-T.J. Gaudet J.R. McGaha W.A. Cross C.L. Alday P.M. Melancon S.S. Didohn T.M. Manzella Waterford 3 Records Center e

'