ML20062K644
| ML20062K644 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/30/1993 |
| From: | Fecteau M, Newmyer W, Savage C WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20062K631 | List: |
| References | |
| NUDOCS 9312230115 | |
| Download: ML20062K644 (84) | |
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V. C. SUMMER SPENT FUEL RACK CRITICALITY ANALYSIS CONSIDERING BORAFLEX SHRINKAGE & GAPS lI lI Apr',19 3 Authored:
L M. W. Fe teau Core De ign A Verified: ft/ d awfpb W. D. Newr$yer /
Criticality Services Leader Approved:
M,C I-C. Savage, Mardger Core Design B 1
lI 1
h ll l
9312230115 931213 gDR ADOCK0500g5
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1 TABLE OF CONTENTS j
1.0 introduction 1
1.1 Design Description 2
1.2 Design Criteria 3
lI 2.0 Analytical Methods 4
f 2.1 Criticality Calculation Methodology 4
2.2 Reactivity Equivalencing for Burnup and IFBA Credit 5
2.3 Boraflex Shrinkage and Gap Methodology 6
3.0 Region 1 Fuel Storage Racks 9
3.1 Reactivity Calculations 9
3.2 IFBA Credit Reactivity Equivalencing 13 3.2.1 IFBA Requirement Determination 14 3.2.2 Infinite Multiplication Factor 16 g
L 3.3 Sensitivity Analysis 17 p
4.0 Region 2 Fuel Storage Racks 18 L
4.1 Reactivity Calculations 18 4.2 Burnup Credit Reactivity Equivalencing 22 g
4.3 Sensitivity Analysis 23 L
5.0 Region 3 Fuel Storage Racks 24 y
5.1 Reactivity Calculations 24 L
5.2 Burnup Credit Reactivity Equivalencing 27 5.3 Sensitivity Analysis 27
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lL 6.0 Soluble Boron Worth and Relaxed Limits 29 6.1 Soluble Boron Worth 29 6.2 Relaxed Limits with 300 PPM Soluble Boron 29 e
7.0 Discussion of Postulated Accidents............................ 31 F
L 8.0 Summary of Criticality Results 33
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Bibliography 61 Table of Contents i
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lI LIST OF TABLES
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Table
- 1. Fuel Parameters Employed in the Criticality Analysis 35 Table
- 2. Benchmark Critical Experiments 36 Table
- 3. Comparison of PHOENIX lsotopics Predictions to Yankee Core 5 I
37 Measurements Table
- 4. Benchmark Critical Experiments PHOENIX Comparison 38 Table
- 5. Data for U Metal and U02 Critical Experiments 39 Table
- 6. Spent Fuel Rack Region 1 K tv Summary 41 Table
- 7. Spent Fuel Rack Region 2 K.ve Summary 42 Table
- 8. Spent Fuel Rack Region 3 Kee, Summary 43 Table
- 9. Minimum Absorber & Burnup Requirements - No Soluble Boron 44 Table 10. Minimum Absorber & Burnup Requirements - 300 PPM Soluble Boron 45 iL I
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1 1
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L List of Tables ii t..
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LIST OF ILLUSTRATIONS Figure
- 1. Region 1 Spent Fuel Storage Cell Diagram 46 Figure
- 2. Region 2 Spent Fuel Storage Cell Diagram 47 Figure
- 3. Region 3 Spent Fuel Storage Cell Diagram 48 I
Figure
- 4. Region 1 Spent Fuel Storage Cell Axial Dimensions 49 Figure
- 5. Region 2 Spent Fuel Storage Cell Axial Dimensions 50 Figure
- 6. Region 3 Spent Fuel Storage Cell Axial Dimensions 51 52 Figure
- 7. Spent Fuel Storage Rack Layout Figure
- 8. Region 1 Minimum IFBA Requirements 53 Figure
- 9. Region 1 Reactivity Sensitivities 54 Figure 10. Region 2 Minimum Burnup Requirements 55 Figure 11. Region 2 Reactivity Sensitivities 56 Figure 12. Region 3 Minimum Burnup Requirements 57 Figure 13. Region 3 Reactivity Sensitivities 58 Figure 14. Spent Fuel Rack Soluble Boron Worth 59 E
Figure 15. Regions 2 & 3 Mmimum Burnup Requirements With 300 PPM Boron 60 L
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List of Illustrations iii
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1.0 INTRODUCTION
l This report presents the results of the criticality re-analysis of the V. C. Summer Spent Fuel Storage Racks with consideration of Boraflex shrinkage and gaps.
The spent rack designs considered herein are existing arrays of poisoned and unpoisoned racks, previously qualified f or storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 5.0 w/o U *.
Occurrences of absorber panel shrinkage and gaps were not considered in the previous criticality analyses for these racks.
I In this analysis, each of the three unique storage configurations in the V. C.
Summer Spent Fuel Rack will be re-analyzed with consideration of Boraflex l
shrinkage and gap development. To provide for future fuel management flexi-bility, storage limits will be developed for enrichments up to and including 5.0 I
w/o by employing credit for integral Fuel Burnable Absorbers (IFBA) and accu-(
mulated fuel assembly burnup.
The following storage configurations and enrichment limits are considered in this analysis:
Region 1 Storage of fresh fuel assemblies with nominal enrichments up to 4.0 w/o U*
utilizing all available storage cells.
Fresh fuel assemt> lies with higher initial enrichments can also be stored in this region provided a minimum number of IFBAs are present in each fuel assembly.
IFBAs consist of neutron absorbing material applied as a thin ZrB2 coating on the outside of the UU2 fuel pellet.
As a result, the neutron absorbing material is a non-removable or integral part of the fuel assembly once it is manuf actured.
I Region 2 Storage of fresh fuel assemblies with nomine!
enrichments up to 2.5 w/o U*
utilizing all available storage cells. Burned fuel assemblies with higher initial enrichments can also be stored in this region provided the minimum requirements for enrichment /burnup are satisfied.
Introduction 1
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Region 3 Storage of fresh fuel assemblies with nominal enrichments up to 1.4 w/o U""
utilizing all available storage cells. Burned fuel assemblies with higher initial enrichments can also be stored in this region provided the minimum requirements for enrichment /burnup are satisfied.
In addition to the analyses described above, relaxed limits will also be devel-oped for each storage region assuming the presence of a minimum soluble boron concentration of 300 ppm. Since the spent fuel pool is nominally maintained at a boron concentration of 2000 ppm, the considerable reactivity conservatism
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present in the spent fuel storage pool is not compromised.
The V. C. Summer Spent Fuel Rack criticality analysis described in this report is based on maintaining K et 5 0.95 for storage of all Westinghouse 17x17 fuel assembly products, including STANDARD, OFA, VANTAGE 5,
VANTAGE SH, VANTAGE + and PERFORMANCE +. For each region, the most reactive or limiting j
fuel assembly type is analyzed to establish the reference K.vf and confirm that the 0.95 limit is not exceeded.
The V. C. Summer Fresh Fuel Racks have been previously analyzed" for storage of Westinghouse 17x17 fuel assemblies with enrichments up to 5.0 w/o U"'.
Since the previous analysis remains valid and applicable, the Fresh Fuel Racks I
are not re-analyzed in this evaluation.
1.1 DESIGN DESCRIPTION The V. C. Summer spent fuel rack Regions 1, 2, and 3 storage cell designs are depicted schematically in Figure 1 on page 46 through Figure 3 on page 48, re-spectively.
Nominal dimensions are provided on each figure.
Axial feature drawings for each region are provided in Figure 4 on page 49 through Figure 6
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on page 51. The layout of the racks within the V. C. Summer spent fuel storage pool is shown in Figure 7 on page 52.
The fuel parameters relevant to this analysis are given in Table 1 on page 35.
With the simplifying assumptions employed in this analysis (no grids, sleeves, axial blankets, etc.), the various types of Westinghouse 17x17 fuel assemblies can be categorized into two basic designs: 17x17 STANDARD (STD). which uti-lizes a 0.374 inch OD fuel rod, and 17x17 OFA, which utilizes a 0.360 inch OD fuel rod. The advanced features of the improved fuel assembly designs (V5, V5H, V+, and P+) are beneficial in terms of extending burnup capability and im-proving fuel reliability, but do not contribute to any meaningful increase in the basic assembly reactivity. Therefore, for this analysis, only the Westinghouse 17x17 STD and OFA fuel assembly types are analyzed since these designs will yield equivalent or conservative reactivity results.
Introduction 2
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1.2 DESIGN 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 assemblies and inserting neutron poison between 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 f actor, Ken, of the fuel assembly array
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will be less than 0.95 as recommended by ANSI 57.2-1983 and NRC guidance"'.
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Introduction 3
4 g.
E 2.0 ANALYTICAL METHODS 2.1 CRITICALITY CALCULATION METHODOLOGY The criticality calculation method and cross-section values are verified by comparison with critical experiment data for fuel assemblies similar to those
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for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty 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 uses the AMPX"" system of codes for cross-section gener-ation and KENO Va for reactivity determination.
The 227 energy group cross-section library that is the common starting point for all cross-sections used for the benchmarks and the stcrage rack is generated from ENDF/B-V* data.
The NITAWL program includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry.
The Nordheim Integral Treatment is used.
Energy and spatial weighting of cross-sections is performed by the XSDRNPM program which is a one-dimensional Sn transport theory code. These multigroup cross-section sets are then used as input to KENO Va which is a three dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 44 critical experiments 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 ge-ometries, 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 (B4C, aluminum, steel, water, etc) that simulate LWR fuel shipping and (l
storage conditions to dry, harder spectrum, uranium metal cylinder arrays at high enrichments (93.2 w/o) with various interspersed materials (Plexiglas and air).
Comparison with these experiments demonstrates the wide range of applicability of the method. Details of the experiments are provided in References 6 through
- 10. Table 2 on page 36 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 benchmarks 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 ap-propriate and more conservative bias.
Analytical Methods 4
--________--._m._m_
The 32 low enriched, water moderated experiments result in an average KENO Va K.tv of 0.9933. Comparison with the average mer.sured experimental K.es of
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1.0007 results in a method bias of 0.0074. The standard deviation of the bias value is 0.0013 AK. 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 AK.
This KENO Va bias and uncertainty are consistent with the previous Westinghouse bias and uncertainty calculated for KENO IV"".
2.2 REACTIVITY EQUIVALENCING FOR BURNUP AND IFBA CREDIT Storage of spent fuel assemblies with initial enrichments higher than that al-lowed by the methodology described in Section 2.1 is achievable by means of the concept of reactivity equivalencing. Reactivity equivalencing is predicated upon the reactivity decrease associated with fuel depletion or the addition of IFBA fuel rods. A series of reactivity calculations is performed to generate a
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set of enrichment-burnup or enrichment-IFBA ordered pairs which all yield an equivalent Keft when the fuel is stored in the V. C. Summer spent fuel racks.
The data points on the reactivity equivalence curve are generated with a trans-
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port theory computer code, PHOENIX"".
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""
library is used with PHOENIX.
l A study was done to examine fuel reactivity as a function of time following discharge from the reactor.
Fission product decay was accounted for using
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CINDER". CINDER is a point-depletion computer code used to determine fission product activities. The fission products were permitted to decay for 30 years af ter discharge. The fuel reactivity was found to reach a maximum at approxi-mately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter discharge. At this time, the major fission product poi-son, Xe"', has nearly completely decayed away. Furthermore, the fuel reactivity
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was found to decrease continuously from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 30 years following dis-charge. Therefore, the most reactive time for a fuel assembly af ter discharge from the reactor can be conservatively approximated by removing the Xe"'.
The PHOENIX code has been validated by comparisons with experiments where the isotopic fuel composition has been examined following discharge from a
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reactor. In addition, an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Comparisons between measured and predicted uranium and plutonium isotopic fuel compositions are shown in Table 3 on page 37.
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The measurements were made on fuel discharged from Yankee Core 5".
The data in Table 3 on page 37 shows that the agreement between PHOENIX pred-ictions and measured isotopic compositions is good.
rt Analyt: cal Methods 5
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The agreement between reactivities computed with PHOENIX and the results of 81 critical benchmark experiments is summarized in Table 4 on page 38.
Key parameters describing each of the 81 experiments are given in Table 5 on page
- 39. These reactivity comparisons again show good agreement between exper-I iment and PHOENIX calculations.
Uncertainties associated with the burnup and IFBA dependent reactivities com-puted with PHOENIX are accounted for in the development of the individual re-activity equivalence limits. For burnup credit, an uncertainty is applied to the PHOENIX calculational results which starts at zero for zero burnup and increases linearly with burnup, passing through 0.01 AK a'i 30,000 MWD /MTU. This bias is considered to be very conservative and is based on consideration of the good agreement between PHOENIX predictions and measurements and on conservative estimates of fuel assembly reactivity variances with depletion history. 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.
Ad-ditional information concerning the specific uncertainties included in each of the V. C. Summer burnup credit and IFBA credit limits is provided in the individual sections of this report.
2.3 BORAFLEX SHRINKAGE AND GAP METHODOLOGY As a result of blackness testing measurements performed at V. C. Summer, the I
presence of shrinkage and gaps in some of the Boraflex absorber panels has been noted. The effects of Boraflex shrinkage and gaps wil be considered in the spent fuel rack criticality evaluations performed for this report.
Previous generic studies of Botaflex shrinkage and gap reactivity effects have been perf ormed
- for storage rack geometries which resemble the V. C. Summer spent fuel racks. The results of these studies (and experience gained in per-forming similar studies for other rack geometries) indicate that:
When absorber panel shrinkage occurs evenly and uniformly (equal pull-back a
is experienced at both ends and the panel remains axially centered and in-tact), meaningful increases in rack reactivity will not occur until more than 7.0 inches of total active fuel length is exposed (3.5 inches on each end).
I Assuming a conservative 4% shrinkage scenario, combined top and bottom fuel exposure will reach 10.56 inches given the initial V. C. Summer Boraflex panel length of 139 inches. For this level of uniform top and bottom ex-I posure, generic study data indicates that reactivity will increase by less than 0.015 AK.
When absorber panel shrinkage occurs all at one end, experience has shown s
that the reactivity impact will remain approximately constant even when an identical length of exposure is added to the opposite end. For the one-end
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scenario, generic data indicates that reactivity will increase by well over 0.06 AK when 4% uniform, one-end shrinkage is assumed in the V. C. Sum-mer racks.
6 Analytical Methods a
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When absorber panel shrinkage is assumed to result in the formation of a s
single large gap in every panel, and all panel gaps are conservatively po-sitioned at the vertical centerline of the active fuel, generic study data in-dicates that reactivity will increase dramatically once a gap size of 1 inch has been exceeded. For an assumed 4% shrinkage at V. C. Summer, the data indicates that reactivity will increase by more than 0.06 AK if all shrinkage is modeled as a single, large (5.56 inch) gap at the centerline.
These generic study results indicate that Boraflex shrinkage and gap formation will result in extremely large reactivity impacts for the conservative scenarios f
of single-end exposure and mid plane gap development.
Accommodating this L.
level of impact in the V. C. Summer rack limits would cause an unreasonable and unacceptable loss of enrichment storage capability. Therefore, a conserva-tive, but more realistic treatment of shrinkage and gap formation will be con-sidered in this criticality evaluation.
To conservatively bound the current measured data and f uture oevelopment of p(
additional shrinkage and gaps, the following assumptions will be employed in the criticality evaluations performed for each of the V. C. Summer storage re-gions which utilize Boraflex absorbers:
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1 1.
All absorber panels will be modeled with 4% width shrinkage.
F 2.
All absorber panels will be modeled with 4% length shrinkage (5.56 inches)
L which wiii be assumed to occur either uniformly (where the panel remains intact over its entire length) or non-uniformly (where a conservative, single p
4 inch gap develops somewhere along the panel length).
2.
Fcr those pancis which are modeled with a gap, the remainder cf the 4%
length shrinkage not accounted for by the single 4 inch gap will be L
conservativeiv "PPiied as bottom end 5h'inka9e-4.
Gaps will be distributed randomly with respect to axial position for the absorber panels which are modeled with gaps.
5.
Shrinkage will be conservatively applied to the bottom end for those
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absorber panels which are modeled with uniform shrinkage (the bottom end L
results in more active fuel exposure than the top end).
Applying all shrinkage to the bottom is very conservative since it ignores the realistic p
possibilities of uniform top / bottom shrinksge or random distribution of top L
or bottom shrinkage among the many absorber panels.
6.
Determination of which panels experience shrinkege and which experience p
L gaps wiii be based on random selection.
Severai scenarios will be con-
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sidered to cover the complete spectrum of shrinkage and gap combinations:
100% of the panels experience non-uniform shrinkage (random gaps).
e 75% of the panels experience non-uniform shrinkage (random gaps) and e
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the remaining 25% of panels experience uniform shrinkage (pull-back) from the bottom-end.
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Analytical Methods 7
I 50% of the panels experience non-uniform shrinkage (random gaps) and a
l the remaining 50% of panels experience uniform shrinkage (pull-back) from the bottom-end.
25% of the panels experience non-uniform shrinkage (random gaps) and m
the remaining 75% of panels experience uniform shrinkage (pull-back) f rom the bottom-end.
100% of the panels experience uniform shrinkage (pull-back) from the e
bott o m-end.
7.
A criticality model which simulates 16 storage cells and 64 individual absorber panels will be employed to provide suf ficient problem size and f: exibility for considering gaps and shrinkage on a random basis.
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A..
absorber material Which is lost to shrinkage or gaps will be conservatively removed from the model. In reality, the absorber material is not lost -- it is simply repositioned by shrinkage to the remaining intact areas of the panel.
The above assumptions are conservative and bounding with respect to the actual shrinkage / gaps measurements taken at V. C. Summer. The use of 4% shrinkage and maximum 4 inch gap bounds the measured shrinkage / gaps at V. C. Summer and is consistent with the upper bound values recommended by EPRI.
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L r
L F
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r Analytical Methods 8
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3.0 REGION 1 FUEL STORAGE RACKS
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This section develops anu describes the analytical techniques and models em-played to perform the criticality analysis and reactivity equivalencing evaluations for the V. C. Summer Region 1 spent fuel racks.
Section 3.1 describes the analyses performed to show that storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 4.0 w/o U* is acceptable in all cell locations.
Section 3.2 describes the reactivity equivalencing analysis which establishes the minimum Integral Fuel Burnable Absorber (lFBA) requirements for assemblies with nominal enrichments above 4.0 w/o. Finally, Section 3.3 presents the results of calculations performed to show the reactivity sensitivity caused by variations in enrichment, c ent er-t o-center spacing, and absorber poison loading.
3.1 REACTIVITY CALCULATIONS
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To show that Region 1 storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 4.0 w/n satisfies the 0.95 K.n criticality acceptance criteria, KENO is used to establish a nominal reference reactivity and PHOENIX is used to assess the effects of material and construction tolerance variations.
A final 95/95 Ken is developed by statistically combining the individual tolerance impacts with the calculational and met.hodology uncertainties and summing this term with the nominal KENO reference reactivity.
The following assumptions are used to develop the nominal case KENO model for the Region 1 fuel storage rack evalue. tion:
1.
The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse l'7x17 OFA design (see Table 1 on page 35 for fuel
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parameters). At the enrichment level being considered for this application, and with the simplified assembly modeling assumptions (no grids, sleeves, axial blankets, etc.), the 17x17 OFA design yields equivalent or bounding reactivity results relative to the other Westinghouse 17x17 fuel types.
2.
All fuel rods contain uranium dioxide at a nominal enrichment of 4.0 w/o over the entire length of each rod.
3.
The fuel pellets are modeled assuming nominal values for theoretical den-sity and dishing fraction.
4.
No credit is taken for any natural enrichment axial blankets.
5.
No credit is taken for any U* or U* in the fuel, nor is any credit taken
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for the build up of fission product poison material.
Region 1 Fuel Storage Racks 9
5
6.
No credit is taken for any spacer grids or spacer sleeves.
L 7.
No credit is taken for any burnabie absorber in the fuei rods.
8.
The moderator is pure water (no boron) at a temperature of 68"F and a p
L density of 1.0 gm/cm 10 r
9.
A nominal Boraflex poison material loading of 0.0264 grams B per square L
centimeter is used throughout the array, based on asbuilt measurements provided by the Boraflex material vendor.
- 10. The array is infinite in lateral (x and y) extent and finite in axial (vertical) cxtent. This allows neutron leakage from only the axial direction.
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- 11. All available storage cells are loaded with fresh fuel assemblies.
L, With the above assumptions, the KENO calculation f or the nominal case (without F
absorber panel shrinkage / gap ef fects) results in a K.v, of 0.9072 with a 95 percent probability /95 percent confidence level uncertainty of 0.0051. The nominal case result can be compared to the results from the shrinkagelgap calculations to p
determine the relative impact of the Boraflex assumptions. The nominal case L
is also used as the center point f or the sensitivity analyses.
To quantify the benefit of axial leakage, a two-dimensional KENO calculation identical to the nominal three-dimensional calculation is performed, except that
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axial leakage is eliminated. The 2D KENO calculation results in a K.sv of 0.9104 with a 95 percent probability /95 percent confidence level uncertainty of t o.0024.
Comparison with the reference 3D result indicates that axial leakage is worth about 0.0032 10.0051 AK.
a To conservatively evaluate the effects of Boraflex shrinkage and gap develop-ment, the methodology described in Section 2.3 is employed. Five shrinkagelgap scenarios are examined to cover the spectrum of shrinkage-to-gap ratios from I
100% gaps an. 0% shrinkage through 0% gaps and 100% shrinkage. Assignment of which panels have gaps or shrinkage, and the axial location of the gap is based on random selection.
L The results of the KENO calculations for the various shrinkage / gap scenarios are provided below:
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Shrinkage / Gap Scenario K.ef 95/95 Uncertainty 100% Random Gaps 0.9170 0.0024 75% Random Gaps /25% Bottom Shrinkage 0 9273 10.0029 50% Random Gaps /50% Bottom Shrinkage 0 9377 0.0024
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Region 1 Fuel Storage Racks 10
L 25% Random Gaps /75% Bottom Shrinkage 0 95B0 10.0024 100% Bottom Shrinkage 0 9754 0.0024 l
L Examination of the trend in reactivity with fraction of gaps / shrinkage indicates that reactivity increases as more of the total absorber shrinkage is positioned at the panel bottom. Positioning all of the absorber shrinkage at the bottom is extremely conservative since it ignores the realistic possibilities of uniform e
top / bottom shrinkage or random occurrences and positioning of the top / bottom I
shrinkage. Previous studies performed for racks similar to V. C. Summer indi-cate that assuming a 50/50 mix of top and bottom shrinkage with random po-sitioning results in a K.ev which is significantly reduced from the one-end only I
scenario.
L The results of blackness testing performed in the Region 1 racks has revealed the presence of gaps in 75% of the inspected panels. Based on this data, the
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Kev from the 75% gap scenario will be used as the reference reactivity for the Region 1 K.ve summation. This scenario most closely represents the measured
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absorber panel shrinkage / gap development experienced to date and will bound future accumulation of even more gaps. Even though a realistic (random) dis-tribution of gaps has been assumed, this calculation is still very conservative due to the assumptions of 4% total width and length shrinkage in every absorber panel; the use of a single, maximum 4 inch gap in every panel with gaps (with the placement of the remaining 1.56 inches of shrinkage at the bottom end); and the placement of the entire 4% of length shrinkage (5.56") at the bottom for those panels without gaps.
Calculational and methodology biases must be considered in the final K ve sum-I mation prior to comparing against the 0.95 Ke t Dmit. The f ollowing biases are a
included:
I Methodology-As discussed in Section 2 of this report, benchmarking of the L
Westinghouse KENO Va methodology resulted in a method bias of 0.0074 AK.
r l
L B10 Self Shielding-To correct for the modeling assumption that individual B" atoms are homogeneously distribt ted within the absorber material (ver-sus clustered about each B4C particler, a bias of 0.0010 AK is applied.
Water Temperature:
To account for the normal range of spent fuel pool water temperatures (40* to 180"F), a reactivity bias of 0.0016 AK is applied.
To evaluate the reactivity eff ects of possible variations in material character-istics and mechanical / construction dimensions, PHOENIX perturbation calculations
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are perf ormed. For the V. C. Summer Region 1 spent fuel storage rack, UO2 and Botafiex material tolerances are considered along with construction tolerances related to the cell I.D., cell center-to-center spacing, and stainless steel thick-f ness. Uncertainties associated with calculational and methodology accuracy are also considered in the statistical summation of uncertainty components.
I k
11 Region 1 Fuel Storage Racks
s L
The following tolerance and uncertainty components are considered in the total uncertainty statistical summation:
U" Enrichment: The standard DOE enrichment tolerance of 10.05 w/o U'"
about the nominal 4.00 w/o U'" reference enrichment was evaluated with PHOENIX and resulted in a reactivity increase of 0.0023 AK.
1
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UO2 Density-A 12.0% variation about the nominal 95% reference theoretical L
density was evaluated with PHOENIX and resulted in a reactivity increase of 0.0025 AK.
Ci L
Fuel Pellet Dishing: A variation in fuel pellet dishing fraction from 0.0% to 2.0% (about the nominal 1.2110% reference value) was evaluated with PHOENIX and resulted in a reactivity increase of 0.0015 AK.
r L
Storage Cell I.D.:
The 10.032 inch tolerance about the nominal 8.85 inch ref erence cell 1.D. was evaluated with PHOENIX and resulted in a reactivity r
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increase of 0.0008 AK.
Center-to-Center Spacing:
The 10.0625 inch tolerance about the nominal g
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10.4025 inch reference cell center-to-center was evaluated with PHOENIX and resulted in a reactivity increase of 0.0052 AK.
Stainless Steel Thickness: The 0.003/0.004 inch tolerances about the nominal O.049/0.065 inch reference stainless steel thicknesses were evaluated with PHOENIX and resulted in a reactivity increase of 0.0011 AK.
Boraflex Absorber Width: The 10.0625 inch tolerance about the nominal 8.45 inch Botaflex panel width was evaluated with PHOENIX and resulted in a f
reactivity increase of 0.0003 AK.
L Boraflex Absorber Thickness: The 10.007 inch tolerance about the nominal
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0.082 inch Boraflex panel thickness was evaluated with PHOENIX and re-L suited in a reactivity increase of 0.0028 AK.
r Boraflex Absorber Length-An assumed 10.25 inch tolerance about the L
nominal 139 inch Boraflex panel length cannot be assessed with PHOENIX due to the 3D nature of the effect. Instead, the impact can be assessed using the results of Westinghouse generic evaluations on this subject'.
The generic data indicates that removal of 0.25 inches of material from the nominal length will not result in any reactivity increase since the threshold f
for 3.5" of active fuel exposure on one end is not exceeded. However, for the reference 75% gap /25% shrinkage scenario where bottom fuel exposure does exceed the 3.5" threshold, a 0.25" reduction applied to the bottom end of all absorber panels is conservatively estimrted from the generic data to
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cause a 0.0075 AK increase in rack reactivity. This w.lue will be considered in the statistical summation of uncertainty terms.
Region 1 Fuel Storage Racks 12
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1 Boraflex B" Loading: The measured 95/95 minimum B" loading of 0.0255 grams per square centimeter was evaluated with PHOENIX and resulted in a reactivity increase of 0.0011 AK.
Assembly Position: The KENO reference reactivity calculation assumes fuel assemblies are symmetrically positioned within the storage cells since ex-I perienr.e has shown that centered fuel assemblies yield equal or more conservative results in rack K.u than non-centered (asymmetric) positioning.
There1 ore, no reactivity uncertainty needs to be applied for this tolerance I
since the most reactive configuration is considered in the calculation of the i
reference K.vv.
Calculation Uncertainty: The KENO calculation for the nominal reference re-activity resulted in a Ken with a 95 percent probability!95 percent confidence level uncertainty of 10.0029 AK.
Methodology Uncertainty: As discussed M Section 2 of this report, compar-ison against benchmark experiments showed that the 95 percent I
probability /95 percent confidence uncertainty in reactivity, due to method, is not greater than 0.0029 AK.
- l The maximum K.e for the V. C. Summer Region 1 spent fuel storage rack is developed by adding the calculational and methodology biases and the statistical sum of independent uncertainties to the nominal KENO reference reactivity. The summation is shown in Table 6 on page 41 and results in a maximum K.,v of i
I 0.9485.
Since Kerr is less than 0.95 including uncertainties at a
95/95 probability / confidence level, the acceptance criteria for " criticality is met for the V. C. Summer Region 1 spent fuel rack for storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 4.0 w/o U*.
I 3.2 IFBA CREDIT REACTIVITY EQUIVALENCING g
Storage of fuel assemblies with nominal enrichments greater than 4.0 w/o U*
in the V. C. Summer Region 1 spent fuel storage racks is achievable by means I
of the concept of reactivity equivalencing.
The concept of reactivity equiv-atencing is predicated upon the reactivity decrease associated with the addition of integral Fuel Burnable Absorbers (IFBA)". IFBAs consist of neutron absorbing material applied as a thin ZrB2 coating on the outside of the UO2 fuel pellet.
As a result, the neutron absorbing material is a non-removable or integral part
+
of the fuel assembly once it is manuf actured.
Two analytical techniques are used to establish the criticality criteria for the i
storage of IFBA fuel in the fuel storage rack. The first method uses reactivity I
equivalencing to establish the poison material loading required to meet the i
criticality limits. The poison material considered in this analysis is a zirconium diboride (ZrB2 ) coating manuf actured by Westinghouse. The second method uses Region 1 Fuel Storage Racks 13
O e
the fuel assembly :nfinite multiplication f actor to establish a reference reactiv-it y.
The reference reactivity point is compared to the fuel assembly peak re-activity to determine its acceptability for storage in the fuel rack.
3.2.1 IFBA REQUIREMENT DETERMINATION I
A series of reactivity calculations are performed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the equivalent Ken when the fuel is stored in the V. C. Summer Region 1 spent fuel rack. The fuel burnup used in the reactivity calculation is that burnup which yields the highest equiv-alent K.n when the fuel is stored in the rack. Fuel assembly depletions per-I formed in PHOENIX show that for the number of IFBA rods per assembly considered in this analysis, the maximum reactivity for rack geometry occurs at zero burnup. Although the boron content in the IFBA rods decreases with fuel
{
depletion, the fuel assembly reactivity decreases more rapidly, resulting ire a 5
maximum fuel rack reactivity at zero burnup.
l 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 Westinghouse 17x17 OFA design (see Table 1 on page 35 for fuel parameters). At the enrichment level being considered for this application, and with the simplified assembly modeling assumptions (no grids, sleeves, axial blankets, etc.), the 17x17 OFA design yields equivalent or bounding j
reactivity results relative to the other Westinghouse 17x17 fuel types, t
2.
The feel assembly is modeled at its most reactive' point in life.
l 3.
The fuel pellets are modeled assuming nominal values for theoretical den-sity and dishing fraction.
4.
No credit is taken for any natural enrichment axial blankets.
5.
No credit is taken for any U* or U* in the fuel, nor is any credit taken j
for the build up of fission product poison material.
i 6.
No credit is taken for any spacer grids or spacer sleeves.
j 7.
The IFBA absorber material is a zirconium diboride (ZrB2 ) coating on the fuel pellet. Each IFBA rod has a nominal poison material loading of 1.50 I
milligrams B" per inch, which is the minimum standard loading offered by Westinghouse for 17x17 OFA fuel assemblies. This rod and IFBA loading assumption is equivalent to or bounding with respect to the 17x17 STD and other advanced Westinghouse 17x17 fuel assembly designs.
8.
The IFBA B" loading is reduced by 5 percent to conservatively account for manuf acturing tolerances and then by an additional 25% to conservatively model a minimum poison length of 108 inches.
Region 1 Fuel Storage Racks 14
I 9.
The moderator is pure water (no boron) at a temperature of 68*F with a density of 1.0 gm/cm'.
- 10. A nominal Boraflex poison material loading of 0.0264 grams B" per square centimeter is used throughout the array, based on asbuilt measurements provided by the Boraflex material vendor.
i
- 11. The array is infinite in lateral (x and y) and axial (vertical) extent.
This precludes any neutron leakage from the array.
- 12. All available storage cells are loaded with fuel assemblies.
Figure 8 on page 53 shows the constant K n contour generated for the V. C.
Summer Region 1 spent fuel storage rack. Note the endpoint at 0 IFBA rods where the nominal enrichment is 4.0 w/o and at 80 IFBA rods where the nominal enrichment is 5.0 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 5.0 w/o U* with each containing 80 IFBA rods is equivalent to the reactivity of the rack when filled with fuel assemblies enriched to a nominal 4.0 w/o and containing no IFBAs.
The data in Figure 8 on page 53 is also provided on Table 9 on page 44.
9 it is important to recognize that the curve in Figure 8 on page 53 is based on reactivity equivalence calculations for the specific enrichment and IFBA combi-nations in actual rack geometry (and not just on simple comparisons of indi-vidual fuel assembly infinite multiplication factors).
In this way, the environment of the storage rack and its influence on assembly reactivity is implicitly considered.
+
The IFBA requirements of Figure 8 on page 53 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 I
to local variations in thermal flux), studies were performed to evaluate this ef-fect 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 intermediate enrichments where standard IFBA
~
patterns may not be available.
In addition, to conservatively account for calculational uncertainties, tne IFBA requirements of Figure 8 on page 53 also include a conservatism of approximately 10% on the total number of IFBA rods at the 5.0 w/o end (i.e., about 8 extra IFBA rods for a 5.0 w/o fuel assembly).
Additional IFBA credit calculations were performed to examine the reactivity ef fects of higher IFBA linear B* loadings (1.5X and 2.0X).
These calculations L
confirm that assembly reactivity remains corstant provided the net B* material per assembly is preserved.
Therefore, with higher IFBA B* loadings, the re-quired 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 5.0 w/o fuel assemblies allows a reduction in the IFBA rod requirement from 80 IFBA rods per assembly to 40 IFBA rods per assembly (80 divided by the the ratio 2.0X/1.0X).
Region 1 Fuel Storage Racks 15
~
w-The equivalent K.et for the storage of fuel in the V. C. Summer Region 1 spent
{-
fuel rack 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 8 on page 53. The KENO Va computer code was used to calculate the storage rack multiplication factor with an equivalent nominal enrichment of 4.0
~~
w/o and no IFBAs.
The reference KENO calculation for this case, including conservative consideration of Boraflex gaps and shrinkage, resulted in a nominal K vf of 0.9273 with a 95 percent probabilityl95 percent confidence level uncer-
[-
tainty of 10.0029. Af ter summation with appropriate biases and uncertainties, the final K.tv f or the V. C. Summer Region 1 spent fuel rack was shown to sat-isfy the 0.95 K.et limit.
l 3.2.2 INFINITE MULTIPLICATION FACTOR The infinite multiplication f actor, Km, is used as a reference criticality reactivity 1
point, and offers an alternative method for determining the acceptability of fuel
)
assembly storage in the V. C. Summer Region 1 spent fuel storage rack. The j
reference K= is determined for a nominal fresh 4.0 w/o fuel assembly.
i The fuel assembly K= calculations are performed using the Westinghouse li-censed core design code PHOENIX-P". The following assumptions were used j
to develoo the infinite multiplication factor model:
1.
The Westinghouse 17x17 OFA fuel assembly was analyzed (see Table 1 on page 35 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 4.0 w/o U* over the entire length of each rod (this is the maximum nominal enrichment that can be stored in the Region 1 rack without IFBA rods).
3.
The fuel array model is based on a unit assembly configuration (infinite in the lateral and axial extent) in V. C. Summer reactor geometry (no rack).
I 4.
The moderator is pure water (no boron) at a temperature of 68* F with a density of 1.0 gm/cm'.
Calculation of the infinite multiplication factor for the Westinghouse 17x17 OFA fuel assembly in V. C. Summer core geometry resulted in a reference K= of 1.460.
This includes a 1% AK reactivity bias to conservatively account for calculational uncertainties. This bias is consistent with the standard conserva-tism included in the V. C. Summer core design refueling shutdown margin cal-culations.
For IFBA credit, all 17x17 fuel assemblies placed in the V. C. Summer Region 1 spent fuel storage rack must comply with the enrichment-IFBA requirements of Figure 8 on page 53 or have a reference Km less than or equal to 1.460.
By Region 1 Fuel Storage Racks 16
I meeting either of these conditions, the maximum rack reactivity will then be less
(
than 0.95, as shown in Section 3.1.
I rL 3.3 SENSITIVITY ANALYSIS To show the dependence of K tv on fuel and storage cells parameters as re-quested by the NRC, the variation of the K se with respect to the f ollowing parameters was developed using the PHOENIX computer code:
235 L
1.
Fuel enrichment, with a 0.50 w/o U delta about the nominal case enrichment.
2.
Center-to-center spacing of storage cells, with a half inch delta about the nominal case center-to-center spacing, j
3.
Poison loading. with a 0.01 gm-B /cm delta about the nominal case poison loading.
~
Results of the sensitivity analysis are shown in Figure 9 on page 54.
L r
i L
e i
k r
I(
Region 1 Fuel Storage Racks 17 s
F L
4.0 REGION 2 FUEL STORAGE RACKS This section develops and describes the analytical techniques and models em-p ployed to perform the criticality analysis and reactivity equivalencing evaluations k
for the V. C. Summer Region 2 spent fuel racks.
Section 4.1 describes the analyses perfcrmed to show that storage of
(
Westinghouse 17x17 fuel assemblies with nominal enrichments up to 2.5 w/o U'"
is acceptable in all cell locations.
Section 4.2 describes the reactivity equivalencing analysis which establishes the minimum burnup requirements for assemblies with nominal enrichments above 2.5 w/o. Finally, Section 4.3 pre-sents the results of calculations performed to show the reactivity sensitivity caused by variations in enrichment, center-to-center spacing, and absorber poi-son loading.
4.1 REACTIVITY CALCULATIONS To show that Region 2 storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 2.5 w/o satisfies the 0.95 K.fr criticality acceptance I
criteria, KENO is used to establish a nominal reference reactivity and PHOENIX is used to assess the effects of material and construction tolerance variations.
A final 95/95 K.tv is developed by statistically combining the individual tolerance impacts with the calculational and methodology uncertainties and summing this term with the nominal KENO reference reactivity.
The following assumptions are used to develop the nominal case KENO model for the Region 2 fuel storage rack evaluation:
1.
The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 17x17 OFA design (see Table 1 ori page 35 for fuel parameters). At the enrichment level being considered for this application, and with the simplified assembly modeling assumptions (no grids, sleeves, axial blankets, etc.), the 17x17 OFA design yields equivalent or bounding reactivity results relative to the other Westinghouse 17x17 fuel types.
2.
All fuel rods contain uranium dioxide at a nominal enrichment of 2.5 w/o over the entire length of each rod.
3.
The fuel pellets are modeled assuming nominal values for theoretical den-sity and c'ishing fraction.
4.
No credit is taken for any natural enrichment axial blankets.
(
5.
No credit is taken for any U* or U'" in the fuel, nor is any credit taken L
for the build up of fission product poison material.
Region 2 Fuel Storage Racks 18_
{
6.
No credit is taken for any spacer grids or spacer sleeves.
7.
No credit is taken for any burnable absorber in the fuel rods.
8.
The moderator is pure water (no boron) at a temperature of 68*F and a density of 1.0 gm/cm'.
A nominal Boraflex poison material loading of 0.0037 grams B" per square 9.
centimeter is used throughout the orray, based on asbuilt measurements provided by the Boraflex material vendor.
- 10. The array is infinite in lateral (x and y) extent and finite in axial (vertical) extent. This allows neutron leakage from only the axial direction.
- 11. All available storage cells are loaded with fresh fuel assemblies.
With the above assumptions, the KENO calculation for the nominal case (without absorber panel shrinkage / gap effects) results in a Kees of 0.8845 with a 95 percent I
probability /95 percent confidence leve, uncertainty of i0.0039. The nominal case result can be compared to the results from the shrinkage / gap calculations to determine the relative impact of the Botaflex assumptions. The nominal case is also used as the center point for the sensitivity analyses.
To quantify the benefit of axial leakage, a two-dimensional KENO calculation identical to the nominal three-dimensional calculation is performed, except that axial leakage is eliminated. The 2D KENO calculation results in a Kw, of 0.8878 with a 95 percent probability /95 percent confidence level uncertainty of 10.0037. Comparison with the reference 3D result indicates that axial leakage is worth about 0.0033 10.0039 AK.
{
To conservatively evaluate the effects of Botaflex shrinkage and gap develop-ment, the methodology described in Section 2.3 is employed. Five shrinkage / gap scenarios are examined to cover the spectrum of shrinkage-to-gap ratios from 100% gaps and 0% shrinkage through 0% gap 2 and 100% shrinkage. Assignment of which panels have gaps or shrinkage, and the axial location of the gap is based on random selection.
The results of the KENO calculations for the various shrinkage / gap scenarios are provided below:
Shrinkage / Gap Scenario K.ve 195/95 Uncertainty 100% Random Gaps 0.8932 10.0022 75% Random Gaps /25% Bottom Shrinkage 0.8944 0.0021 50% Random Gaps /50% Bottom Shrinkage 0.8987 10.0021 Region 2 Fuel Storage Racks 19
______________m
l 25% Random caps /75% Bottom shrinkage 0 9040 0.0021 1
100% Bottom Shrinkage 0 9126 0.0046
.I Examination of the trend in reactivity with fraction of gaps / shrinkage indicates I
that reactivity increases as more of the total absorber shrinkage is positioned at the panel bottom. Positioning all of the absorber shrinkage at the bottom is extremely conservative since it ignores the realistic possibilities of uniform top / bottom shrinkage or random occurrences and positioning of the top / bottom j
shrinkage. Previous studies performed for racks similar to V. C. Summer indi-cate that assuming a 50/50 mix of top and bottom shrir*.oge with random po-sitioning results in a Keve which is significantly reduced from the one-end only scenario.
,L Blackness testing performed in the Region 2 racks has failed to find any oc-currence of gaps in any of the inspected panels. Based on this data, the K.ve from the 0% gap /100% bottom shrinkage scenario is used as the reference re-activity for the Region 2 K.tv summation. This scenario will bound future accu-mutation of gaps since the 100% bottom shrinkage scenario results in the most l
conservath e X.4 v.
It should be noted that this calculation is very conservative i
u due to the assumptions of 4% total width and length shrinkage in every absorber panel and the placement of the entire 4% of length shrinkage (5.56") at the bottom of every absorber panel.
Calculational and methodology biases must be considered in the final K.ve sum-F mation prior to comparing against the 0.95 K.<v limit. The following biases are included:
Methodology-As discussed in Section 2 of this report, benchmarking of the Westinghouse KENO Va methodology resulted in a method bias of 0.0074 AK.
B10 Self Shielding: To correct for the modeling assumption that individual B" atoms are homogeneously distributed within the absorber material (ver-sus clustered about each B4C particle), a bias of 0.0053 AK is applied.
Water Temperature:
To account for the normal range of spent fuel pool l
water temperatures (40 to 180 F), a reactivity bias of 0.0016 AK is applied.
l l
To evaluate the reactivity effects of possible variations in material character-istics and mechanical / construction dimensions, PHOENIX perturbation calculations l
[
are perf ormed. For the V. C. Summer Region 2 spent fuel storage rack, UO2 and
)
Boraflex material tolerances are considered along with construction tolerances I
related to the cell I.D., cell center-to-center spacing, and stainless steel thick-ness. Uncertainties associated with calculational and methodology accuracy are also considered in the statistical summation of uncertainty components.
s The following tolerance and uncertainty components are considered in the total uncertainty statistical summation:
e f"
Region 2 Fuel Storage Racks 20
U* Enrichrnent: The standard DOE enrichment tolerance of 10.05 w/o U" about the nominal 2.50 w/o U* reference enrichment was evaluated with PHOENIX and resulted in a reactivity increase of 0.0046 AK.
UO2 Density: A 2.0% variation about the nominal 95% reference theoretical density was evaluated with PHOENIX and resulted in a reactivity increase of 0.0028 AK.
Fuel Pellet Dishing: A variation in fuel pellet dishing fraction from 0.0% to 2.0% (about the nominal 1.2110% reference value) was evaluated with
[
PHOENIX and resulted in a reactivity increase of 0.0017 AK.
L
)
Storage Cell 1.D.:
The 10.032 inch tolerance about the nominal 8.85 inch f
reference cell 1.D. was evaluated with PHOENIX and resulted in a reactivity L
increase of 0.0019 AK.
f Center-to-Center Spacing:
The 10.0625 inch tolerance about the nominal L
10.4025/10.1875 inch reference ceii center-to-center spacing was evaiuated with PHOENIX and resulted in a reactivity increase of 0.0061 AK.
r-L Stainless Steel Thickness: The 10.003/0.004 inch tolerances about the nominal 0.049/0.065 inch reference stainless steel thicknesses were evaluated with PHOENIX and resulted in a reactivity increase of 0.0010 AK.
s Boraflex Absorber Width: The 0.0625 inch tolerance about the nominal 8.45 inch Boraflex panel width was evaluated with PHOENIX and resulted in a reactivity increase of 0.0003 AK.
Boraflex Absorber Thickness: The 10.007 inch tolerance about the nominal I
0.032 inch Boraflex panel thickness was eva!aated with PHOENIX and re-h suited in a reactivity increase of 0.0118 AVo Boraflex Absorber Length-An assumed 10.25 inch tolerance about the nominal 139 inch Boraflex panel length cannot be assessed with PHOENIX due to the 3D nature of the ef fect. Instead, the impact can be assessed using the results of Westinghouse generic evaluations on this subject".
The generic data indicates that removal of 0.25 inches of material from the nominal length will not result in any reactivity increase since the threshold for 3.5" of active fuel exposure on one end is not exceeded. However, for the reference 0% gap /100% shrinkage scenario where bottom fuel exposure does exceed the 3.5" threshold, a 0.25" reduction applied to the bottom end f
of all absorber panels is conservatively estimated to cause a 0.0031 AK increase in rack reactivity. This value will be considered in the statistical L
summation of uncertainty terms.
Boraflex B" Loading: The measured 95/95 minimum B" loading of 0.0033 grams per square centimeter was evaluated with PHOENIX and resulted in a reactivity increase of 0.0069 AV.
Region 2 Fuel Storage Racks 21
Assembly Position: The KENO reference reactivi!v caiculation assumes fuel assemblies are symmetrically positioned within the storage cells since ex-perience has shown that centered fuel assemblies yield equal or more conservative results in rack K.vf than non-centered (asymmetric) positioning.
Therefore, no reactivity uncertainty needs to be applied for this tolerance since the most reactive configuration is considered in the calculation of the ref erence K.es.
Calculation Uncertainty: The KENO calculation for the nominal reference re-activity resulted in a K.vi with a 95 percent probabilityl95 percent confidence level uncertainty of 0.0046 AK.
I Methodology Uncertainty: As discussed in Section 2 of this report, compar-ison against benchmark experiments showed that the 95 percent probability /95 percent confidence uncertainty in reactivity, due to method, is not greater than 0.0029 AK.
The maximum K.tv for the V. C. Summer Region 2 spent fuel storage rack is developed by adding the calculational and methodology biases and the statistical I
sum of independent uncertainties to the nominal KENO reference reactivity. The summation is shown in Table 7 on page 42 and results in a maximum K.ev of 0.9442.
Since K.fi is less than 0.95 including uncertainties at a
95/95 probability / confidence level, the acceptance criteria for criticality is met for the V. C. Summer Region 2 spent fuel rack for storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 2.5 w/o U"'.
I 4.2 BURNUP CREDIT REACTIVITY EQUIVALENCING Storage of burned fuel assemblies in the V. C. Summer Region 2 spent fuel storage rack area is achievable by means of the concept. of reactivity equiv-
.g alencing. The concept of reactivity equivalencing is predicated upon the reac-g tivity decrease associated with fuel depletion. For burnup credit, a series of reactivity calculations are performed to generate a set of enrichment-fuel as-sembly discharge burnup ordered pairs which all yield an equivalent K.ef when I
stored in the spent fuel storage racks.
Figure 10 on page 55 shows the constant K.vf contour generated for close packed storage in the V. C. Summer Region 2 spent fuel racks. This curve represents combinations of fuel enrichment and discharge burnup which yield the same rack multiplication f actor (Kevi) as the rack loaded with fresh fuel at 2.5 w/o U"'.
Note in Figure 10 on page 55 the endpoints at 0 MWD /MTU where the enrichment is 2.5 w/o and at 21600 MWD /MTU where the enrichment is 5.0 w/o. The in-I terpretation of this endpoint data is as follows: the reactivity of the spent fuel rack containing 5.0 w/o U"' fuel at 21600 MWD /MTU burnup is equivalent to the reactivity of the rack containing fresh fuel having an initial nominal enrichment of 2.5 w/o. The burnup credit curve shown in Figure 10 on page 55 includes a Region 2 Fuel Storage Racks 22
I..
reactivity uncertainty of 0.0072 AK, consistent with the minimum burnup re-quirement of 21600 MWD /MTU at 5.0 w/o.
It is important to recognize that the curve in Figure 10 on page 55 is based on calculations of constant rack reactivity.
In this way, the environment of the storage rack and its influence on assembly reactivity is implicitly considered.
For convenience, the data from Figure 10 on page 55 is also provided on Table 9 on page 44.
The tabulated values have been conservatively reported to allow the use of linear interpolation between the data points.
The effect of axial burnup distribution on assembly reactivity has been consid-ered in the development of the V. C. Summer Region 2 burnup credit limit.
I Previc,us evaluations have been performed to quantify axial burnup reactivity effects and to confirm that the reactivity equivalencing methodology described in Section 2.2 results in calculations of conservative burnup credit limits". The I
evaluations show that axial burnup effects can cause assembly reactivity to in-crease, but the burnup-enrichment combinations required to cause this are well beyond those required by the V.
C.
Summer Region 2 burnup credit limit.
I Therefore, additional accounting of axial burnup distribution effects in the V. C.
Summer red on 2 burnup credit limit is not necessary.
i
!I 4.3 SENSITIVITY ANALYSIS e o show the dependence of K.<<
on fuel and storage cells parameters as re-f""
quested by the NRC, the variation of the K.ef with respect to the following I
parameters was developed using the PHOENIX computer code:
am 1.
Fuel enrichment, with a 0.50 w/o U*
delta about the nominal case enrichment.
2.
Center-to-center spacing of storage cells, with a half inch delta about the nominal case center-to-center spacing.
3.
Poison loading, with a 0.0037 gm-B"/cm' delta about the nominal case poison loading.
Results of the sensitivity analysis are shown in Figure 11 on page 56.
lI l
I h
i 1
Region 2 Fuel Storage Racks 23 I
I
?
I I
5.0 REGION 3 FUEL STORAGE RACKS This section develops and describes the analytical techniques and models em-ployed to perform the criticality analysis and reactivity equivalencing evaluations for the V. C. Summer Region 3 spent fuel racks.
Section 5.1 describes the analyses performed to show that storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 1.4 w/o I
U'"
is acceptable in all cell locations.
Section 5.2 describes the reactivity equivalencing analysis which establishes the minimum burnup requirements for assemblies with nominal enrichments above 1.4 w/o. Finally, Section 5.3 pre-I sents the results of calculations performed to show the reactivity sensitivity caused by variations in enrichment, center-to-center spacing, and stainless steel structural material.
I 4
5.1 REACTIVITY CALCULATIONS To show that Region 3 storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 1.4 w/o satisfies the 0.95 K.ve criticality acceptance 4
criteria, KENO is used to establish a nominal reference reactivity and PHOENIX is used to assess the effects of material and construction tolerance variations.
A final 95/95 K.ve is developed by statistically combining the individual tolerance impacts with the calculational and methodology uncertainties and summing this term with the nominal KENO reference reactivity.
!I The following assumptions are used to develop the nominal case KENO model for the Region 3 fuel storage rack evaluation:
1.
The fuel assembly parameters relevant to the criticality analysis are based I
on the Westinghouse 17x17 STD design (see Table 1 on page 35 for fuel parameters). At the enrichment level being considered for this application, and with the simplified assembly modeling assumptions (no grids, sleeves, I
axial blankets, etc.), the 17x17 STD design yields equivalent or bounding reactivity results relative to the other Westinghouse 17x17 fuel types.
2.
All fuel rods contain uranium dioxide at a nominal enrichment of 1.4 w/o over the entire length of each rod.
I 3.
The fuel pellets are modeled assuming nominal values for theoretical den-sity and dishing fraction.
4.
No credit is taken for any natural enrichment axial blankets.
1 5.
No credit is taken for any U* or U'" in the fuel, nor is any credit taken for the build up of fission product poison material.
I l
l Region 3 Fuel Storage Racks 24
t g
'E..
6.
No credit is taken for any spacer grids or spacer sleeves.
7.
No credit is taken for any burnable absorber in the fuel rods.
'I 8.
The moderator is pure water (no boron) at a temperature of 68*F and a density of 1.0 gm/cm*.
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Figure 9 Region 1 Reactivity Sensitivities 54
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Figure 11 Region 2 Reactivity Sensitivities 56
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Figure 14 Spent Fuel Rack Soluble Boron Worth 59
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Figure 15 Regions 2 & 3 Minimum Burnup Requirements With 300 PPM 4
Boron 60 J
b b
r-H BIBLIOGRAPHY L
1.
Boyd, W.
A., etal., Criticality Analysis of V. C. Summer Fuel Racks., October 1987.
r L
2.
Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. Yu Grimes OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978.
3.
W.
E.
Ford ill, CSRL-V:
Processed ENDFIB-V 227-Neutron-Group and C
Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding L
Studies, ORNL/CSDITM-160 June 1982.
4.
N.
M.
Greene, AMPX: A Modular Code System for Generating Coupled
~
u Multigroup Neutron-Gamma Libraries from ENDFIB, ORNLITM-3706, March 1976.
I 5.
L.
M.
Petrie and N.
F.
Landers, KENO Va--An Improved Monte Car /o
~
Criticality Program With Supergrouping, NUREGICR-0200, December 1984.
7 6.
M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage L
of Power Reactor Fuel, BAW-1484-7, July 1979.
r 7.
S. R. Bierman and E. D. Clayton, Criticality Separation Between Subcritical U
Clusters of 2.35 wt% 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2438, October 1977.
F L
8.
S. R. Bierman and E. D. Clayton, Criticality Separation Between Subcritical Clusters of 4.29 wt% 235U Enriched UO2 Rods in Water with fixed Neutron Poisons, PNL-2615, August 1979.
I" 9.
S. R. Bierman and E. D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 wt% and 4.31 wt% 235U Enriched UO2 Rods in Water at a Water-to-Fuel Volume Ratio of 1.6, PNL-3314 July 1980.
10.
J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) Metal Cylinders, Nuclear Science and Engineering, Volume 52, pages 350-359,1973.
c 11.
D.
E.
- Mueller, W.
A.
- Boyd, and M.
W.
Fecteau (Westinghouse NFD), Qualification of KENO Calculations with ENDFIB-V Cross Sections, i
American Nuclear Society Trrnsactions, Volume 56, pages 321-323, June 1988.
12.
A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors, WCAP-10106, June 1982.
~
Bibliography 61
- 13. Askew, J. R., Fayers, F. J., and Kemshell, P. B., A General Description of the Lattice Code W/MS, Journal of British Nuclear Energy Society, 5,
pp.
564-584, 1966.
- 14. England, T.
R., C/NDER - A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.
- 15. Melehan, J.
B., yankee Core Evaluation Program Final
- Report, WC AP-3017-6094, January 1971.
16.
W. A. Boyd and D. E. Mueller (Westinghouse NFD), E//ects of Poison Panel Shrinkage and Gaps on Fuel Storage Rack Reactivity, American Nuclear Society Transactions, Volume 56, pages 323-324, June 1988.
I
- 17. Davidson, S.L.,
et al., VANTAGE 5 Fuel Assembly Reference Core Report, Addendum t, WCAP-10444-P-A, March 1986.
- 18. Nquyen, T.
O., et al., Qualification of the PHOENIX-PIANC Nuclear Design System for Pressurized Water Reactor Cores, WC AP-11597-A, November 1987.
- 19. W. A. Boyd and M. W. Fecteau (Westinghouse NFD), E//ect of Axial Burnup on Fuel Storage Rack Burnup Credit Reactivity, American Nuclear Society I
Transactions, Volume 62, pages 328-329, November 1990.
I I
i I
j i
Bibliography 62
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[-
ATTACHfdE N / ~f 5.0 CRITICALITY ANALYSIS OF FRESH FUEL RACKS This section cescribes the analytical techniques and models employed to per-form the criticality analysis for storage of fresh fuel in the V. C. Summer fresh fuel racks.
Since the fresn fuel racks are maintained in a dry condition. the criticality analysis will show that the rack Ken is less than 0.95 for the full density and low censity octimum moceration conditionc. The low censity optimum moder-ation scenario is an accicent situation in wnich no creait can be taken for soluble boron. The criticality method anc cross-section library are the same as those ciscussed in Section 2 of this report.
The f ollowing assumptions wece used to develop the nominal case KENO model for the storage of fresh fuel in the fresh fuel racks uncer full density and low density optimum moderation conditions:
The fuel assembly contains the highest enrichment authorized. is at its most 1.
reactive point in life, and no credit is taken for any burnable poison in the fuel rocs.
2.
All fuel rocs contain uranium dioxide at an enrienment o f 5.0 w/o U * *
- over the infinite length of eacn roc.,
3.
No credit is taken for any U* *
- or U8** in the fuel, nor is any credit taken for the buildup of fission product poison material.
4 No credit is taken for any spacer grics or spacer sleeves.
Calculations for these racks have shown that the W 17x17 OFA fuel assembly yields a larger K.o than does the W 17x17 Standard fuel assembly wnen both fuel assemblies have the same U* *
- enrichment in full density water. Thus, i
only the W 17x17 OFA fuel assembly was analyzed (See Table 2 for fuel pa-rameters) in full density water.
Criticality Analysis of Fresn Fuel Racks 13 l
I
i i
i i
l 5.1 FULL DENSITY MODERATION ANALYSIS in the nominal case KENO mocel for the full density moceration analysis, the moderator is pure water at a temperature of 68'F.
A conservative value of 1.0 gmiem* is used for the density of water. The fuel array is infinite in lateral and axial extent which precludes any neutron leakage from the array.
The KENO calculation for the nominal case resulted in a K.,e of 0.9235 witn a 95 percent procacility/95 percent confidence level uncertainty of 20.0082.
The maximum K.o under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing process in acdition to asymmetric positioning of fuel assemblies within the storage cells. Stuoies of asymmetric positioning of fuel assemolies within trie Storage cells has snown that symmetrically placea fuel assemolies yield con-servative results in racx K.,,. The manuf acturing tolerances are stacked in sucn a manner to minimi:e tne assemoly center-to-center spacing and the total vol-ume of steel thereov causing an increase in reactivity. The sneet metal toler-ancec are considerec c,iong with construction tolerances related to the cell !.O.
and cell center-to-center spacing. For the fresh foal storage racxs, the assemoly center-to-center spacmg is recuced from a nominal va!ue of 21" to a minimum of 20.94".
Thus, the most conservative, or " worst case", KENO mocel of the fresh fuel storage racks contains a minimum water gap of 11.72" with sym-metrically placed fuel assemblies.
Based on the analysis desr:ribed above, the following ecuation is used to de-velop the maximum K.o for the V. C. Summer fresh fuel storage racks:
K.n =
K..,u - Sm.mee * (((ks) 2..,n + (k s)
- m.m.a ]
where:
K..
worst case KEND K.o that incluces material
=
tolerances. and mechanical tolerances wnich can result in spacings between assemolies less than nominal Bm.mos method bias determined from benchmark a
critical comparisons 1
Criticality Analysis of Fresh Fuel Racks i
14
)
1 I
s 1
~~"
{5/S5 uncertainty in the worst case KENO
=
95/95 uncertainty in the method bias
=
Substitutmg calculated values in the order listed above, tne result is:
K.v. = 0.9235 - 0.0083 + (((0.0082)* + (0.0018)8 ] = 0.9402 Since K.e is less than 0.95 including uncertainties at a 95/95 probability confi-dance level, the acceptance criteria for criticality is met.
5.2 LOW DENSITY OPTIMUM MODERATION ANALYSIS in the low density optimum moderation analysis, the fuel array is infinite in only the axial extent which prectuoes any neutron leakage from the toD or bottom of the array.
Calculations have shown that the W 17x17 STD fuel assembly yields a larger K.se tnan oces the W 17x17 OFA fuel assemoly when both assemblies have the same U***
enrienment at low water densities. Thus, the W 17x17 STD as-sembly was used in the optimum moderation analysis.
Analysir, of the V. C. Summer racks has shown that the maximum rack K.e under low density moderation conditions occurs at 0.04 gm/cm
- water density. The KENO calculation of the V. C. Summer fresh racks at 0.04 gm/cm' water density resulted in a peak K.e. of 0.8959 with a 95 percent probability and 95 percent confidence level uncertainty of 0.0079. Figure 19 shows the fresh fuel rack reactivity as a function of the water density.
The minimum cell center-to-center spacing, rack module spacing ano material tolerances have been includeo in the base case model and result in a storage cell separation distance of 11.86*' and a rack module separation distance of 20.94 inches.
Studies of asymmetric positioning of fuel assemblies within the storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack K.ve.
Baseo on the analysis described above, the following ecuatien is used to de--
velop the maximum K.e for the V. C. Summer fresh fuel storage rects under low density optimum moderation conditions:
K.ee = Ko...
+ Bm.,*
+ /[(ks)*b... + (ks)
- mna
]
where:
Criticality Analysis of Fresh Fuel Racks 15 l
Q~.
~
~.
s s
base case KENO K.e that includes nominal
=
mechanical and material dimension method bias determined from benchmark
=
critical comparisons ksen.
95/95 uncertainty in the base case KENO K.e, 95/95 uncertainty in the method bias
=
Substituting calculated values in the order listed above, the result is:
K.ee = 0.8959 + 0.0083 + /[(0.0079)8 + (0.0018)8 ] = 0.9123 Since K..,
is less than 0.95 including uncertainties at a
95/95 probability / confidence level, the acceptance criteria for criticality is met.
Criticality Analysis of Fresh Fuel Racks 16
6.0 ACCEPTANCE CRITERION FOR CRITICALITY The neutron multiplication factor in spent fuel pool and fresh fuel vault shall be less than or equal to 0.95, including all uncertainties, under all conditions.
The analytical methods employed herein conform with ANSI N18.2-1973, "Nu-clear Safay Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System; ANSI 57.21983, " Design Objectives f or LWR Spent Fuel Storage Facilities at Nuclear Power Stations." Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Methoos for Nuclear Criticality Safety," NRC Stancard Review Plan, Section 9.1.2. " Spent Fuel Storage"; and the NRC guidance. "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," ANSI 57.3-1983, " Design Reouirements for New Fuel Storage Facilities at Light Water Reactor Plants."
~
I Acceptance Criterion For Criticality 17
r--
.s i
Table 2.
Fuel Parameters Employed in Criticality Analysis r
Parameter W 17x17 OFA W 17x11 STANDARD
[
Number of Fuel Reds per Assemoly 26fe 264 Rod Zire-4 Clad 0.D.
(inch) 0 360 0 374 i
Clad Thickness (inch) 0.0225 0.0225
\\
Fuel Pe!let 0.D. (Inch) 0 3088 0 3225 Fuel Pellet Dansity
(% of Theoretical) 96 96 1
i Fuel Pellet Dishing Factor 0.0 0.0
-Rod Piteh (Inch) 0.496 0.496 1
Number of Zirc-4 Guide Tubes 24 24' I
Guide Tube 0.D.
(1nch) 0.474 0.4841 Guide -Tube Thickness (inch) 0.016 0.0182 i
Number of instrument Tubes
.1 1
i Instrument Tube 0.D.
(i nch) 0.474 0.4842' l
1
~
l Instrument Tube Thickness (inch) 0.016 0.0185' i
i
{
f 2
m.
is
..,.=..
in. v e.. o.o..n.in.c
.i.. o.as:..e o.cieinen..
ei mey.
in.
==
- a.........
.ev.e, in........
..n..
e in,.
.v..
4 19 -
d
-av
r s
r
~
e o
i
~
I s
11.85" j
s W
9.00" v
s r
9 -
CELL CENTER TO CENTER ( 21.0"
)
Figure 6.
SCE&G Fresh Fuel Storage Call Nominal Dimensione, 30
~~.
.m.
g 1
4
\\\\%%\\W%\\\\\\\\\\%%%_%_\\\\WW. %\\T.O.'.O..'.?'/_.:0..:0. :
N ////< l'///////////// /// /////////////,:....
E
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'_.0.:.00.:.00.'. '/.t..d :0._:0:0._207
\\
N
.--e _.l*.r REFLECTOR
- Z N/
bismsun// / '/ ////
Z I
xj
,,. /
I d._._} _ ////////////%.....,,.,.._,,.. b.i i i !_j
/
h
/,,,~....
l_ _ [! I l i _j l_j l:-
7._..-..
xN
/
T': :-, _.,,__,,._., r--
i p L. J i... ; L..J L._.! L ].; J L._Ji.1 - >.
- i N
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{
x\\
- 3 II
- +::
a
_. e t:
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~
..c
- t. :.i.t..
.,.__.,..... r..
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.._I t.
,.. }
2.,.8
.i i..
t.. a.
.,r.
,{ a1 ul.:
.L
...; t.
- g...!.
t.J*_{l..dtw.
)
- .:4 L
..8 is i!
5
. _.,j
[.
lt 1;..
L.:
[*.
- ..,_,i t...s: :
..s.
.3 t
gl.:
i
... a l s.
- l t
1
.,-l g.
.=
4.:
- .'t r.';*. '.' *.': *.'.. *:.' ' :.: ' ~ ':
- c ~, *: =.~- '
1.=,~.~.*:~,.*:.T - : ': ' ~.~ *.~~ *..s:: - ~.'.-=- ~.- *. *.~ : *.~:'.*. =.
m..
- REFLECTOR ASSUMED TO BE FULL DCNSITY WATER IN ANALYTICAL MODEL i
Figure 7.
SCE&G Fresh Fuel Rack Layout 31
_s
~
5
.90 a
I g
g g
_____J___..____I_____
l I
i l
I *'
I I
I
.85 I
i i
i 1
i i
i
.____r.____
____.7____
_...__y____..
i l
I i
.86 I
i I
i I
i l
I r-~~-
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e 1
I OTA
.34 4
i i
STD I
i a
u_
i i
VORST CPSE POINT L.
I I
I l
l L.I.E2 I
I I
l M
,____L-___
____.L____
_____!____..____L___
1 I
I I
o I
I I
1
.80 i
I I
I
.____p____
____4____
____a____..____w.____
l i
I I
I I
I I
,73 I
I I
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i 1
1
_7____
_____g____..____r___
I I
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.76 I
1 l
1 I
I I
i l
i l
i I
1 I
l
- d.00
.05
.10
.15
.20 WATERDENSITY(G/CC)
Figure 19.
SensitMty of K.n to Water Density in the SCE&G Fresh Fuel Storage Racks F
43 i
q
~
i I
BIBLIOGRAPHY i
1.
Nuclear Regulatory Commission.
Letter to All Power.
Reactor Licensees., from B. K. Grimes OT Position for Review and Acceptance of Spent Fuel Storage and Hand /Ing Applications.,, April 14 1978.
l 2.
W.
E.
Ford \\ \\ \\, CSRL-V:
Processed ENDFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety Reactor and Shielding Studies. ORNL/CSDlTM-130. June 1982.
3.
N. M. Greene. AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDFIB, ORNLITM-3706. March 1976.
4 L M. Petrie anc N. F. Cross. KENO IV--An improved Monte Carlo Criticality Program. DRNL-4938. November 1975.
5.
M. N. Baldwin Critical Experiments Supporting Close Proximity Water Storage of Power Reactor fuel, BAW-1484-7, July 1979.
6.
J.
T.
Thomas. Critical Three-Dimensional Arrays of U 193.21 Metal Cylinoers. Nuclear Science and Engineermg, Volume 52. pages 350-363.
1973.
7.
A. J. Harris. A Description of the Nuclear Design and Analysis Programs for Solling Water Reactors, WCAP-10106, June 1982.
I 8.
Askew. J. R., Fayers. F. J., and Kemshell, P. B., A General Description of the Latt/ce Code W/MS, Journal of British Nuclear Energy Society, 5. pp.
564-584, 1966.
)
9.
England, T.
R CINDER i
A One-Point Depletion and Fission Product Program, WAPD-TM-334, August 1962.
?
- 10. Meishan.
J.
B Yankee Core Evaluation Program Final
- Report, WCAP-3017-6094, January 1971.
Bibliograpny 44 i
4
l~
ATTsch>oXX Westinghouse NuclearManufacturing sa ns Electric Corporation Divisions nneuemsmam e23nns s
93CG*-G 0041 Aptil 30,1993 CU-27202 Mr. B. L Johnson, Supervisor Core Engineering South Carolina Electric and Gas Company
" G._.' g, Mail Code 563
~"
V. C. Summer Nuclear Station P. O. Box 88 Jenkinsville, SC 29065
Dear Mr. Johnson:
SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION FINAL CRITICALITY ANALYSIS REPORT Enclosed is the final report, entitled "V. C. Summer Spent Fuel Rack Criticality Analysis Considering Boraflex Shrinkage and Gaps." The report shows that Westinghouse 17x17 fuel assemblies with nominal enrichments up to 5.0 w/o can be safely stored in all three regions of the V. C. Summer Spent Fuel Storage Racks. Credit is taken for burnup and IFBAs. The criticality report also includes analysis which takes credit for 300 ppm of soluble baron in the spent fuel pool.
1 Also attached is the " Procedure to Calculate the infinite Multiplication Factor for the V. C.
Summer Region 1 Spent Fuel Racks." This attachment describes how to use PHOENIX-P to calculate an equivalent fuel assembly reactivity for IFBA credit in the V. C. Summer Region 1 Spent Fuel Racks.
The attached report has been verified. SCE&G comments from the draft review have also been i
incorporated.
l Per SCE&G specifications, ten bound and one unbound copy have been provided.
4 l ~
150lv
,.(
r.-
i Mr. B. L Johnson 93CG'-G-0041 h
l I
If you should require further assistance, please call either Bill Newmeyer at 412/374-6534 or me l
at 412/374-2373.
.i Sincerely, c JM
- /
v t
Kevin C. Hoskins Project Engineer Domestic Sales & Customer Projects cc:
M. N. Browne i
L Cartin i
B. Christiansen W. Haltiwanger B. Jolley W i
i i
i f
f f
i i
l T
i 150lv
'I 1
1
Westinghouse Proprictuy Class 2 CDB-93-088 1
l 1
i Procedure to Calculate the Infinite Multiplication Factor for the V. C. Summer Region 1 Spent Fuel Storage Racks In addition to the supplied IFBA credit curve for storing fuel assemblies with nominal U2" enrichments greater than 4.0 w/o in the V. C. Summer Region I spent fuel racks, an alternate method can be used to establish the criticality criteria for storage ofIFBA fuel in the spent fuel storage racks. This method uses the fuel assembly infinite multiplictn factor, k., to establish a reference reactivity. The reference reactivity point i' compared to the fuel assembly peak reactivity to determine its acceptability for storage h1 the fuel rack.
The established fuel assembly reactivity, k., as determined for the V. C. Summer Region 1 spent fuel racks is 1.460. This method is useful when the fuel assembly type being l
considered for storage does not quite ratisfy the IFBA credit curve. The procedure to calculate the infinite multiplication factor for the V. C. Summer Region I spent fuel rack is discussed below.
~
j The fuel assembly k. calculation is performed using the Westinghouse licensed core design code PHOENIX-P. The following assumptions are used to develop the infinite multiplication factor model:
- 1. The fuel assembly is modeled at its most reactivity point in life.
- 2. The fuel pellets are modeled assuming nominal value for theoretical density and dishing fraction.
- 3. No credit is taken for any natural emichment axial blankets.
- 4. No credit is taken for any U2" or U236 in the fuel, nor is any credit taken for the build up of fission product poison material.
- 5. The moderator is pure water (no boren) at a temperature of 68 F with a density of 1.0 gm/cm'.
- 6. Burnable' absorber loading are as-built or nominal less a 5 % manufacturing tolerance.
- 7. Burnable absorber locations are modeled exactly.
- 8. Pan-length burnable absorbers are modeled with a reduced B loading based on the 5
ratio of the absorber length to the fuel rod length. For example, the BS loading for a 108 inch IFBA rod would be reduced by 25% (108 inches /144 inches).
1 of 4
Wectinghouse Proprietary Class 2 CDB-93-088 r
l Based on standard core design methodology, ALPHA can be used to run a Hot Full Power l
Unit Assembly as shown in Figure 1. The results should then be restarted at Cold Zero Power conditions as shown in Figure 2. These example decks were used to develop the infinite multiplication factor, k,, of 1.460 which is the limit for acceptable storage in the Region 1 racks at 4.0 w/o U2".
The example input decks can be modified to determine the reactivity of fuel assembly types used at V. C. Summer. If the result is less than the k. limit of 1.460, the fuel assembly type is acceptable for storage in the Region I racks.
P
.)
W. D. New lyer
/
Criticality Product Line Leader Verified:
2N~D
-i M.
. Fecreau Core Design A Date: <//2g/r3 i
e r
~
f c
i
)
k i
2 of 4 1
k L
-l
.1 Westinghouse Proprietary Class 2 CDB-93-088 Figure 1 l
Sample HFP Input Deck (ALPHA Loader)
/
TITLE =SCE&G PHNX UA 170FA 4.00 W/O HFP CONDITIONS
/
i CALC (31)= 01/ UNIT (31)= 1/ FILEID(31)= 170FA40H /
/
I PUNCH = FALSE / CORE =3 LOOP / CATEGORY =2 / 3-LOOP 17X17 PLANT
/
POWER 2775.0 / FROM WCAP-12564 CY6 NDR
=
THZP 557.0 /
=
TIN 554.8 /
=
LOADING = 66411/ 157*0.423 OFA LOADING e
/
ENRICHMENT (1)= 4.0 / TYPEFUEL(1)= 2 /170FA FUEL ASSEMBLY
/
FRACDENS(l)=_0.950 / UTOPICS(2,1)=.1.0E-20,1.0E-20 /
/
DISH (1)= 1.2110 / GASPRES(l)= 275.0 / IFBADENS(l)= 0 /
/
i ASSEMGEOM(1,1)= 16,1,1,1,6*212,217 /170FA FUEL ASSEMBLY
/
ASSEMBU(1,1)= "O / PPM ="O /
l
/
/ ** OFF-NOMINAL RESTART INPUTS "
I
/
READFILE(31,1) 170FA40H / READUNIT(31,1) 1/ READSTEP(1,31) 1/
f
- q 4
l RPRESSURE 14.7 / RRELPOW 0 / TCZP 68.0 /
i
/
STOP
_j
~
tj 3 of 4
i
~ '
htinghouse Proprinny Clus 2 CDB-93-088 r
Figure 2 Sample CZP Input Deck for Final k. Calculation t
k (ALPHA Loader) t
/
TITLE =SCE&G PHNX UA 170FA 4.00 W/O CZP CONDITIONS
/
CALC (31)= 03 / UNIT (31)= 1/ FILEID(31)= 170FA40C /
I i
/
PUNCH = FALSE / CORE =3 LOOP / CATEGORY =2 / 3-LOOP 17X17 PLANT
/
POWER = 2775.0 / FROM WCAP-12564 CY6 NDR THZP
= 557.0 /
= 554.8 /
~
LOADING = 66411/ 157*0.423 OFA LOADING
/
ENRICHMENT (1)= 4.0 / TYPEFUEL(1)= 2 /170FA FUEL ASSEMBLY
/
FRACDENS(l)= 0.950 / UTOPICS(2,1)= 1.0E-20,1.0E-20 /
/
DISH (1)= 1.2110 / GASPRES(l)= 275.0 / IFBADENS(l)= 0 /
/
ASSEMGEOM(1,1)= 16,1,1,1,6*212,217 /170FA FUEL ASSEMBLY I
/
ASSEMBU(1,1)= **0 / PPM =**0 /
/
/ " OFF-NOMINAL RESTART INPUTS **
/
.{
READFILE(31,1)= 170FA40H / READUNIT(31,1)= 1/ READSTEP(1,31)= 1/
RPRESSURE= 14.7 / RRELPOW= 0 / TCZP= 68.0 /
/
i STOP l
i I
4 of 4 i