ML20141G841
| ML20141G841 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/14/1996 |
| From: | Fecteau M, Lam H, Srinita S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20141G839 | List: |
| References | |
| CAA-96-146, NUDOCS 9705230027 | |
| Download: ML20141G841 (25) | |
Text
I CAA-96-146 f
Criticality Analysis of the Point Beach Nuclear Plant Spent Fuel Storage Racks Considering Boraflex Gaps and Shrinkage with Credit for
} Integral Fuel Burnable Absorbers S. Srinilta Prepared : 5l-dW) L E!l9/46 S. Srinilta Core Analysis B Verified: 4)q(
lYQTLam Core Analysis C Approved- ,A o V!94 M. W. Fecteau O Core Analysis A
{ Westinghouse Commerical Nuclear Fuel Division r
DR DO K 05 00 66 P PDR
l Table of Contents 1.0 I n t rod u ct ion ... .. .. .... .. .. .. ...... ........ .. .... .. ...... ...... .. .I 1.I Desi gn Description ..... .. .... . ..... .. ...... .. .. .. . .. .. .. . . . . .. .. .. .. . . . . ........ .. .. .. .. .. .. .. ...... .. . . ...... I 1.2 De s i g n Cri teri a . . .. .. . .. .. .. . . . . . . . . .. . . . . . . . . .. .. .. . . . . . . . . . . . . .. .. . . . . . . . . .. .. .. .. . . . . . . . . . .. . . .. . . . . . . . . . . . .. . . .. . . 2 f
2.0 Analytical Met hods .. .. ...... .. .. ...... .......... .. .......... .. 3 2.1 Criticality Calculation M ethod ology ...................... .......... ...................................... 3 2.2 Reactivity Equivalencing for IFB A Credit .................... .......................................... 3 2.3 Boraflex S hrinkage And Gap Methadology ......... ............................................. . ... 3 3.0 Criticality Analysis of Spen 1 Fuel Racks .... .. ........ .. .. 6 3.1 Reactivity Calc ulati o n s .. .. .. .. .... ...... . .......... .... .. ............ .. .. .. .. . . .. .. .. .. ...... .. .................. . 6 3.2 IFB A Credit Reactivity Equivalencing......................................................... ........... 8 3.2.1 IFB A Req uirement Determination................... ........................................... 9 3.2.2 Infinite M ultiplica tion Factor.................................................................... . 10 3.3 S ol uble B oron Worth . .. .. .. .. ........ .. .. ...... .. .. .... .. .. ...... .... .... .. .. .. .. .... .. ............. ... . ....
!l 4.0 Discussion of Postulated Accidents .... ........ .. ........ 12 i
5.0 Summary of Criticality Results ... ...... .. .. .. .. .. .... 14 B i bliog ra p hy .. .. ........ ...... . ........ .. .. ........ .. .. .. .... .. 21 f
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f Point Beach Spent Fuel Racks i
List of Tables h Table 1. Fuel Parameters Employed in the Criticality Analysis.................................15 Table 2. Point Beach Spent Fuel Rack Ke g Summary................................................16 Table 3. Point Beach Spent Fuel Rack IFBA Requirement........................................17
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List of Figures Figure 1. Point Beach Spent Fuel Rack Cross-Section View of Array of Storage Cells......18 Figure 2. Point Beach Spent Fuel Rack IFBA Requirement............................. ..................19 -
' Figure 3. Point Beach Spent Fuel Rack Soluble Boron Worth ................................. .... .... 20 1
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Point Beach Spent Fuel Racks iii
1.0 Introduction This report presents the results of a criticality analysis of the Point Beach spent fuel storage racks with consideration of Boraflex shrinkage and gap development.
The spent fuel storage rack designs considered herein are existing arrays of fuel racks, previously qualified for storage of various 14x14 fuel assembly types with maximum enrichments up to and including 4.75 w/o 23SUW . To provide for future fuel management flexibility, storage limits will be developed to allow storage of Westinghouse 14x14 OFA or 14x14 STD fuel with nominal enrichments up to and including 5.0 w/o 235 U by employing credit for Integral Fuel Bumable Absorbers (IFBA).
The following storage configurations and enrichment limits are considered in this analysis:
Westinghouse Storage of Westinghouse 14x14 OFA and 14x14 STD fuel assemblies 14x14 OFA and with nominal enrichments up to and including 4.6 w/o 235U utilizing 14x14 STD Fuel all available storage cells is allowed. Fresh fuel assemblies with higher Assemblies initial nominal enrichments up to and including 5.0 w/o 235 U can also be stored in these racks 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 UO2 fuel pellet. As a result, the neutron absorbing material is a non-removable or integral part of the fuel assembly once it is manufactured.
The Point Beach spent fuel rack analyses are based on maintaining Ke g s 0.95 for storage of Westinghouse 14x14 OFA and 14x14 STD fuel assemblies under full water density conditions.
1.1 Design Description The cross-section view of array of storage cells of the Point Beach spent fuel racks is shown in Figure 1 on page 18, with nominal dimensions provided on the figure.
The fuel parameters relevant to this analysis are given in Table 1 on page 15. With the simplifying assumptions employed in this analysis (no grids, sleeves, axial blankets, etc.), the various advanced products (V5, V+, and P+) of Westinghouse 14x14 OFA fuel and the various advanced products (V5H and .422 0.D. P+) of Westinghouse 14x14 STD fuel are beneficial in terms of extending burnup capability and improving fuel reliability, but do not contribute to any meaningful increase in the basic assembly reactivity. Therefore, future fuel assembly upgrades do not require a criticality analysis if the fuel parameters specified in Table 1 continue to remain bounding.
Point Beach Spent Fuel Racks I
1.2 Design Criteria Criticality of fuel assernblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between fuel assemblies and inserting neutron poison between them.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor, K eg, of the fuel assembly array will be less than 0.95 as recommended by ANSI 57.2-1983 and NRC guidanceW .
t Point Beach Spent Fuel Racks 2
2.0 Analytical Methods 2.1 Criticality Calculation Methodology The criticality calculation method and cross-section values are verified by comparison with critical experirnent data for fuel assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps, low moderator densities and spent fuel pool soluble boron.
The design method which insures the criticality safety of fuel assemblies in the fuel storage rack is described in detail in the Westinghouse Spent Fuel Rack Criticality Analysis Methodology topical reportW . This report describes the computer codes, benchmarking, and methodology which are used to calculate the criticality safety limits presented in this report.
As determined in the benchmarking in the topical report, the method bias using the described methodology of NITAWL-II, XSDRNPM-S and KENO-Va is 0.00770 AK with a 95 percent probability at a 95 percent confidence level on the bias of 0.00300 AK. These values will be used throughout this report as needed.
2.2 Reactivity Equivalencing for IFBA Credit Storage of fuel assemblies with higher initial enrichments than 4.6 w/o 235 U is achievable by means of the concept of reactivity equivalencing. Reactivity equivalencing is predicated upon the reactivity decrease associated with the addition of IFBA fuel rods. A series of reactivity calculations is performed to generate a set of enrichment-IFBA ordered pairs which all yield an equivalent K err when the fuel is stored in the Point Beach spent fuel racks. The data pSnts on the reactivity equivalence curve are generated with the transport theory computer code, PHOENIX-P.
PHOENIX-P is a depletable, two-dimensional, multigroup, discrete ordinates, transport theory code which utilizes a 42 energy group nuclear data library.
Uncertainties associated with the IFBA dependent reactivity computed with PHOENIX-P are accounted for in the development of the individual reactivity equivalence limits. An uncertainty of approximately 10% of the total number of IFBA rods is accounted for in the development of the IFBA requirements. Additional information concerning the specific uncertainties included in the Point Beach IFBA credit limit is provided in Section 3.2 of this report.
2.3 Boraflex Shrinkage And Gap Methodology As a result of blackness testing measurements performed at other storage rack facilities, the presence of shrinkage and gaps in some of the Boraflex absorber panels has been noted. The effects of Boraflex shrinkage and gaps were considered in the spent fuel rack criticality evaluations performed for this report.
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Previous generic studies of Boraflex shrmkage and reactivity effects have been performedW for storage rack geometries which resemble the Point Beach spent fuel racks. The results of these studies (and experience gained in performing similar studies for other rack geometries) indicate that:
- When absorber panel shrinkage occurs evenly and uniformly (equal pullback is experienced at both ends and the panel remains axially centered and intact), 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). Assuming a conservative 4% shrinkage scenario, combined top and bottom fuel exposure will reach 3.84 inches given the initial Point Beach Boraflex panel length of 146 inches. For this level of uniform top and bottom exposure, gen'eric study data indicates that reactivity will not increase.
+When absorber panel shrinkage occurs all at one end, experience has shown 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 scenario, the largest fuel exposure of 5.56 inches occurs at the bottom end for the Point Beach spent fuel racks. Generic data for the 11.12 inch total fuel exposure indicates that reactivity will increase by almost .02 AK.
- When absorber panel shrinkage is assumed to result in the formation of a single large gap in every panel, and all panel gaps are conservatively positioned at the vertical centerline of the active fuel, generic study data indicates that reactivity will increase dramatically once a gap size of 1 inch has been exceeded. For an assumed 4% shrinkage in the Point Beach spent fuel racks, the data indicates that reactivity will increase by more than 0.07 AK if all shrinkage is modeled as a single, large (5.84 inches) 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 of mid-plane gap development.
Accommodating this level of impact in the Pdt Beach spent fuel rack limits would cause an unreasonable and unacceptable loss of enrichment storage capability. Therefore, a conservative, but more realistic treatment of shrinkage and gap formation will be considered in this criticality evaluation.
To bound the current and future development of shrinkage and gaps, the following assumptions will be employed in the criticality evaluations performed for each of the Point Beach storage regions which utilize Boraflex absorbers:
. l. All absorber panels will be modeled with 4% width shrinkage.
- 2. - All absorber panels will be modeled with 4% length shrinkage (5.84 inches) which will be assumed to occur either uniformly (where the panel remains intact over its entire length) or non-uniformly (where a conservative, single 4 inch gap develops somewhere along the panel length).
- 3. For those panels which are modeled with a gap, the remainder of the 4% length shrinkage not accounted for by the single 4 inch gap will be conservatively applied as bottom or top end shrinkage.
- 4. Gaps will be distributed randomly with respect to axial position for the absorber panels which are modeled with gaps.
- 5. Determination of which panels experience shrinkage and which experience gaps will be based Point Beach Spent Fuel Racks 4
on random selection. Several scenarios will be considered to cover the complete spectrum of shrinkage and gap combinations:
h 100% of the panels experience nonuniform shrinkage (random gaps).'
50% of the panels experience nonuniform shrinkage (random gaps) and the remaining 50%
of panels experience uniform shrinkage (pullback) from the bottom-end.
50% of the panels experience nonuniform shrinkage (random gaps) and the remaining 50% of panels experience uniform shrinkage (pullback) from the top-end.
100% of the panels experience uniform shrinkage (pullback) from the bottom-end.
100% of the panels experience uniform shrinkage (pullback) from the top-end.
- 6. . A criticality model which simulates 16 storage cells and 64 individual absorber panels will be employed to prcride sufficient problem size and flexibility for considering gaps and shrinkage on a random basis.
- 7. All 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 bounding with respect to the upper bound values for shrinkage and gaps recommended by EPRIW . These assumptions have been used in previous NRC-approved criticality reports relating to boraflex gaps and shrinkage.
L Point Beach Spent Fuel Racks 5 r
3.0 Criticality Analysis of Spent Fuel Racks This section describes the analytical techniques and models employed to perform the criticality analysis and reactivity equivalencing evaluations for the Point Beach spent fuel storage racks.
Section 3.1 describes the reactivity calculations performed with the nominal enrichment up to and j including 4.6 w/o 235 e of assemblies '
U. Section 3.2 describes the analysis which allows for2 storag5 with nominal enrichments above 4.6 w/o 235 U and up to and including 5.0 w/o U by taking credit for Integral Fuel Burnable Absorbers (IFBAs). Section 3.3 presents the results of l calculations performed to show the reactivity sensitivity of soluble boron.
3.1 Reactivity Calculations To show that storage of burned and fresh Westinghouse 14x14 OFA or 14x14 STD fuel assemblies in the Point Beach spent fuel racks satisfies the 0.95 Ke rrcriticality acceptance criteria, KENO is used to establish a nominal reference reactivity and PHOENIX-P is used to assess the I effects of material and construction tolerance variations. The nominal temperature range of 50*F !
to 180*F is considered in the analysis. A final 95/95 Kerr 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 storage of fuel assemblies in the Point Beach spent fuel rack:
- 1. The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 14x14 OFA design which is more reactive than the Westinghouse 14x14 STD design (see Table 1 on page 15 for fuel parameters).
- 2. All fuel assemblies conthin uranium dioxide at a nominal enrichment of 4.6 w/o over the entire length of each rod.
- 3. The fuel pellets are modeled assuming nominal values for theoretical density and dishing fraction.
- 4. No credit is taken for any natural or reduced enrichment axial blankets.
- 5. No credit is taken for any 234 U or 236 U in the fuel, nor is any credit taken for the buildup of fission product poison material.
- 6. No credit is taken for any spacer prids or spacer sleeves.
- 7. No credit is taken for any bumable absorber in the fuel rods.
- 8. The moderator is pure water (no boron) at a temperature of 68'F. A limiting value of 1.0 gm/cm3is used for the density of water to conservatively bound the range of normal (50*F to 180*F) spent fuel pool water temperatures.
- 9. The array is infinite in lateral (x and y) extent and finite in axial (vertical) extent.
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- 10. All available storage cells are loaded with fuel assemblies.
I1. Nominal Boraflex poison plate dimensions for width, thickness and length are assumed.
- 12. Minimum Boraflex loading of 0.02 gm/cm 2is assumed.
Point Beach Spent Fuel Racks 6
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I With the above assumptions, the KENO calculation for the nominal case results in a Ke g of 0.92921 with a 95 percent probability /95 percent confidence level uncertainty of 0.00146 AK.
This K eg is the nominal reactivity assuming no Boraflex gaps or shrinkage.
To conservatively evaluate the effects of Boraflex shrinkage and gap development, 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 and 100% shrinkage. Assignment of which panels have gaps or shrinkage, and the axial location
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of the gap is based on random selection.
Of the five KENO cases mentioned under item 5 near the end of Section 2.3, the 50% gaps and 50% shrinkage case results in the highest Ke g. The KENO Ke g for this worst case of Boraflex
[ gaps and shrinkage is 0.93176 with a 95 percent probability /95 percent confidence level
) uncertainty of 0.00146 AK. This K ge will be used as the reference reactivity for the Point Beach spent fuel rack storage configuration.
Calculational and methodology biases must be considered in the final Ke g summation prior to comparing against the 0.95 Ke g 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 0.00770 AK.
3'B Self Shielding: To correct for the modeling assumption that individual10B atoms are f homogeneously distributed within the absorber material (versus clustered about each B4C particle), a bias of 0.00140 AK is applied.
Water Temperature: To account for the effect of the normal range of spent fuel pool water temperatures (50*F to 180*F) on water cross section properties, a reactivity bias of 0.00081 AK is applied.
To evaluate the reactivity effects of possible variations in material characteristics and mechanical /
construction dimensions, PHOENIX-P perturbation calculations are performed. For the Point Beach spent fuel rack configuration, UO2material tolerances are considered along with construction tolerances related to the cell I.D., cell pitch, and Boraflex poison panels.
Uncertainties associated with calculation and methodology accuracy are also considered in the statistical summation of uncertainty components.
The following tolerance and uncertainty components are considered in the total uncertainty statistical summation:
235 U Enrichment: The standard DOE enrichment tolerance ofi0.05 w/o 235 U about the nominal 4.6 w/o 235 U reference enrichment was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00202 AK.
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UO 2Density: A 12.0% variation about the nominal 95% reference theoretical density was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00271 AK.
Fuel Pellet Dishing: A reduction of the fuel pellet dishing fraction to 0.0% from the nominal 1.1926% reference value was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00141 AK.
Storage Cell I.D.: The +0.083/-0.0 inch tolerance about the nominal 8.25 inch reference cell f
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{ I.D.was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00427 AK.
Storage Cell Pitch: The +0.093/-0.01 inch tolerance about the nominal 9.938 inch reference i cell pitch was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00097 AK.
Horaflex Absorber Width: The +0.0/-0.3 inch tolerance about the nominal 8.0 inch Boraflex panel width was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00079 AK.
Horaflex Absorber Thickness: The +0.01/-0.02 inch tolerance about the nominal 0.10 inch Boraflex panel thickness was evaluated with PHOENIX-P and resulted in a reactivity increase of 0.00152 AK.
Assembly Position: The KENO reference reactivity calculation assumes fuel assemblies are symmetrically positioned within the storage cells since experience has shown that centered fuel assemblies yield equal or more conservative results in rack Keg 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 reference Keg. -
Calculation Uncertainty: The KENO calculation for the nominal reference reactivity resulted in a K eg with a 95 percent probability /95 percent confidence level uncertainty of 0.00241 AK.
Methodology Uncertainty: As discussed in Section 2 of this report, comparison against benchmark experiments showed that the 95 percent probability /95 percent confidence uncertainty in reactivity, due to method, is not greater than 0.00300 AK.
The maximum K eg for the Point Beach spent fuel rack storage configuration is developed by f adding the calculational and methodology biases and the statistical sum ofindependent uncertainties to the KENO reference reactivity. The summation is shown in Table 2 on page 16 and results in a maximum Keg of 0.94876.
Since K eg is less than 0.95 including uncertainties.at a 95/95 probability / confidence level, the acceptance criteria for criticality is met for storage of Westinghouse 14x14 OFA or 14x14 STD fuel assemblies with nominal enrichment up to and including 4.6 w/o 235 U in the Point Beach j spent fuel racks. i{
3.2 IFBA Credit Reactivity Equivalencing Storage of fuel assemblies with nominal enrichments greater than 4.6 w/o 235 U in the Point Beach spent fuel storage racks is achievable by means of the concept of reactivity equivalencing. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with the addition of Integral Fuel Bumable Absorbers (IFBA). IFBAs consist of neutron absorbing material applied as a thin zirconium diboride (ZrB 2) 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 manufactured.
Two analytical techniques are used to establish the criticality criteria for the storage of IFBA fuel in the fuel storage rack. The first method uses reactivity equivalencing to establish the poison
{ material loading required to meet the criticality limits. The poison material considered in this analysis is a ZrB 2coating manufactured by Westinghouse. The second method uses the fuel Point Beach Spent Fuel Racks 8
assembly infinite multiplication factor to establish a reference reactivity. The reference reactivity point is compared to the fuel assembly peak reactivity to determine its acceptability for storage in the fuel racks.
3.2.1 IFBA' Requirement Determination A series of reactivity calculations are performed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the equivalent K e rrwhen the fuelis stored in the spent fuel racks. The following assumptions were used for the IFBA rod assemblies in the PHOENIX-P
' models:
- 1. The fuel assembly parameters relevant to the criticality analysis are based on the Westinghouse 14x14 OFA design which is more reactive than the Westinghouse 14x14 STD design (see Table 1 on page 15 for fuel parameters).
- 2. The fuel assembly is modeled at its most reactive point in life.
- 3. The fuel pellets are modeled assuming nominal values for theoretical density and dishing fraction.
- 4. No credit is taken for any natural enrichment or reduced enrichment axial blankets.
- 5. No credit is taken for any 234 U or 236 U in the fuel, nor is any credit taken for the buildup of fission product poison material.
- 6. No credit is taken for any spacer grids or spacer sleeves.
- 7. The IFBA absorber materialis a ZrB 2coatin on the fuel pellet. Each IFBA rod has a nominal poison materialloading of 1.67 milligrams I per inch (1.0X loading), which is the minimum standard loading offered by Westinghouse for 14x14 OFA fuel assemblies.
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- 8. The IFBA B loading is reduced by 5% to conservatively account for manufacturing tolerances and then by an additional 33% to conservatively model a minimum absorber length of 96 inches.
- 9. The moderator is pure water (no boron) at a temperature of 68'F with a density of 1.0 gm/cm3 .
- 10. The array is infinite in lateral (x and y) and axial (vertical) extent. This precludes any neutron leakage from the array.
Figure 2 on page 19 shows the constant Ke rrcontour generated for the Point Beach spent fuel racks. Note the endpoint at 0 IFBA rods where the nominal enrichment is 4.6 w/o 235U and at 16(1.0X) IFBA rods where the nominal enrichment is 5.0 w/o 235 U. The interpretation of the endpoint data is as follows: th'e reactivity of the fuel rack array when filled with fuel assemblies 235 enriched to a nominal 5.0 w/o U with each containing 16 (1.0X) IFBA rods is equivalent to the reactivity of the rack when filled with fuel assemblies enriched to a nominal 4.6 w/o 235 U and containing no IFBAs. The data in Figure 2 on page 19 is also provided on Table 3 on page 17.
It is important to recognize that the curve in Figure 2 on page 19 is based on reactivity equivalence calculations for the specific enrichment and IFBA combinations in actual rack geometry (and not just on simple comparisons of individual fuel assembly infinite multiplication factors). In this way, the environment of the storage rack and its influence on assembly reactivity is implicitly considered.
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The IFBA requirements of Figure 2 on page 19 were developed based on the standard IFBA patterns used by Westinghouse. However, since the worth of individual IFBA rods can change depending on position within the assembly (due to local variations in thermal ficx), studies were performed to evaluate this effect and a conservative reactivity margin was included in the development of the IFBA requirement to account for this effect. This assures that the IFBA requirement remains valid at intermediate enrichments where standard IFBA patterns may not be available. In addition, to conservati vely account for calculational uncertainties, the IFBA requirements of Figure 2 on page 19 also include a conservatism of approximately 10% on the total number of IFBA rods at the 5.0 w/o 235 U end (i.e., about 2 extra IFBA rods for a 5.0 w/o 235 U fuel assembly).
Additional IFBA credit calculations were performed to examine the reactivity effects of higher 10 IFBA linear B loadings (1.5X and 2.0X). These calculations confirm that assembly reactivity IU remains constant pWided the net B material per assembly is preserved. Therefore, with higher 10 IFBA B loadings, the required number of IFBA rods per assembly can be reduced by the ratio of the higher loading to the nominal 1.0X loading. For example, using 2.0X IFBA in 5.0 w/o 235 U fuel assemblies allows a reduction in the IFBA rod requirement from 16 IFBA rods per assembly to 8 IFBA rods per assembly (16 divided by the ratio 2.0X/1.0X).
3.2.2 Infinite Multiplication Factor The infinite multiplication factor, Km, is used as a reference criticality reactivity point, and offers an alternative method for determining the acceptability of fuel assembly storage in the spent fuel racks. The reference K,is determined for a fresh Westinghouse 14x14 OFA fucl assembly at 4.6 235 w/o U nominal enrichment.
The fuel assembly K, calculations are performed using the Westinghouse licensed core design code PHOENIX-P. The following assumptions were used to develop the infinite multiplication factor model:
- 1. The Westinghouse 14x14 OFA fuel assembly was analyzed (see Table 1 on page 15 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 a nominal enrichment of 4.6 w/o 235 U over the entire length of each rod.
- 3. The fuel array modelis based on a unit assembly configuration (infinite in the lateral and axial extent) in Point Beach reactor geometry (no rack).
- 4. The moderator is pure water (no boron) at a temperature of 68* F with a density of 1.0 gm/cm3.
Calculation of the infinite multiplication factor for the Westinghouse 14x14 OFA fuel assembly in the Point Beach core geometry at cold condition resulted in a ref:rence K, of 1.49364. This -
includes a 1% AK reactivity bias to conservatively account for calculational uncertainties. This bias is consistent with the standard conservatism included in the Point Beach core design refueling shutdown margiu calculations.
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For IFBA credit, all Westinghouse 14x14 fuel assemblies placed in the Point Beach spent fuel racks must comply with the enrichment-IFBA requirements of Figure 2 on page 19 or have a reference K. less than or equal to 1.49364. By meeting either of these conditions, the maximum rack reactivity will then be less than u.95, as shown in Section 3.1.
I 3.3 Soluble Boron Worth PHOENIX P calculations were also performed to evaluate the reactivity benefits of soluble boron for the Point Beach spent fuel storage configuration. Results of these calculations are provided in Figure 3 on page 20. As the curve shows, the presence of soluble boron in the Point Beach spent fuel pool provides substantial reactivity margin.
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Point Beach Spent Fuel Racks 11
4.0 Discussion of Postulated Accidents Most accident conditions will not result in an increase in K eg of the rack. Examples are:
Fuel assembly drop The rack structure pertinent for criticality is not excessively deformed on top of rack and the dropped assembly which comes to rest horizontally on top of the rack has sufficient water separating it from the active fuel height of stored assemblies to preclude neutronic interaction.
Fuel assembly drop Design of the spent fuel racks is such that it precludes the insertion of a between rack fuel assembly between rack modules.
modules Loss of:ooling Reactivity decreases since loss of cooling causes an increase in systems temperature, which causes a decrease in water density, which results in decreased reactivity.
However, two accidents can be postulated which would increase reactivity beyond the analyzed condition. One such postulated accident would be placement of a fresh fuel assembly of the highest permissible enrichment outside and adjacent to a storage rack module. This abnormal location of a fresh fuel assembly could result in an increased reactivity. To very conservatively estimate the reactivity impacts of such an occurrence in the spent fuel racks, the impact of loading a fresh assembly at 4.6 w/o 235 U enrichment adjacent to an outside face, which has no boraflex, of a 4x5 array of the Point Beach spent fuel rack cells loaded with fresh assemblies at 4.6 w/o 235 U enrichment was determined. The reactivity increase associated with this ubnormal placement of the fresh assembly is less than 0.024 AK.
A second accident which could result in increased reactivity would be a "cooldown" event during which the pool temperature would drop below 50*F. Calculations show that if:he Point Beach spent fuel pool water temperature was to decrease from 50*F to 32*F, reactivity could increase by about 0.00081 AK.
For occurrences of any of the above postulated accidents, the double contingency principle of ANSI /ANS 8.1-1983 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The worth of soluble boron in the Point Beach spent fuel pool has been calculated with PHOENIX-P and is shown in Figure 3 on page 20. As the curves show, the presence of soluble boron in the pool water reduces rack reactivity significantly and is more than sufficient to offset the positive reactivity impacts of any of the postulated accidents. To bound the 0.024 AK reactivity increase from the most limiting accident in the spent fuel racks, it is estimated that 250 ppm of soluble boron is required.
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Point Beach Spent Fuel Racks 12
(' Therefore should a postulated accident occur which causes a reactivity increase in the Point Beach spent fuel racks, Keg will be maintained less than or equal to 0.95 'due to the presence of at least 250 ppm of soluble boron in the spent fuel pool water.
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Point Beach Spent Fuel Racks 13
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5.0 Summary of Criticality Results For the storage of fuel assemblies in the spent fuel storage racks, the acceptance criteria for criticality requires the effective neutron multiplication factor, K y, et to be less than or equal to 0.95, including uncertainties, under all conditions. Maintaining at least 250 ppm of soluble boron in the spent fuel pool water is required to mitigate the consequences of a postulated accidents as described in Section 4.0 of this report.
[ This report shows that the acceptance criteria for criticality is met for the Point Beach spent fuel storage racks for the storage of Westinghouse 14X14 OFA and 14x14 STD fuel assemblies with the following configurations and enrichment limits:
Westinghouse Storage of Westinghouse 14x14 OFA and 14x14 STD fuel assemblies p 14x14 OFA and with nominal enrichments up to and including 4.6 w/o 235 U utilizing L 14x14 STD Fuel all available storage cells is allowed. Fresh fuel assemblies with higher Assemblies initial nominal enrichments up to and including 5.0 w/o 235 U can also
[ be stored in these racks provided a minimum number of IFBAs are l present in each fuel assembly. 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 materialis a non-removable
[ or integral part of the fuel assembly once it is manufactured.
The analytical methods employed herein conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7 Fuel
[ Handling System; ANSI 57.2-1983, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations," Section 6.4.2; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety"; and the NRC Standard Review Plan, Section 9.1.2, " Spent Fuel
( Storage".
[
L Point Beach Spent Fuel Racks 14
l l
[ Table 1. Fuel Parameters Employed in the Criticality Analysis
( Parameter W 14x14 OFA W 14x14 STD Number of Fuel Rods per Assembly 179 179
{
Rod Zirc-4 Cladding O.D. (inch) 0.400 0.422 Cladding Thickness (inch) 0.0243 0.0243
(
Fuel Pellet O.D.(inch) 0.3444 0.3659
( Fuel Pellet Density (% of Theoretical) 95 95 Fuel Pellet Dishing Factor (%) 1.1926 1.187
[ Rod Pitch (inch) 0.556 0.556 Number of Zirc-4 Guide Tubes 16 16 Guide Tube O.D. (inch) 'O.526 0.539 Guide Tube Thickness (inch) 0.017 0.017
[ Number ofInstrument Tubes 1 1 Instrument Tube O.D. (inch) 0.399 0.422 Instrument Tube Thickness (inch) 0.0235 0.0240
{
(
(
(
[
[
[
[
t Point Beach Spent Fuel Racks 15
Table 2. Point Beach Spent Fuel Rack K,g Summary AK K,g Nominal KENO Reference Reactivity: 0.93176
{
Calculational & Methodologv Biases:
Methodology (Benchmark) Bias +0.00770
( 10 B Particle Self-Shielding Bias +0.00140
[ Pool Temperature Bias (50*F- 180*F) +0.00081 TOTAL Bias +0.00991
[- Best-Estimate Nominal K,g: 0.94167 Tolerances & Uncertainties:
UO2EnrichmentTolerance +0.00202 UO2Density Tolerance +0.00271 Fuel Pellet Dishing Variation +0.00141 Cell Inner Diameter Tolerance +0.00427 Cell Pitch Tolerance +0.00097 Boraflex Thickness Tolerance +0.00152
{
Boraflex Width Tolerance +0.00079 Calculational Uncertainty (95/95)
( +0.00241 Methodology Bias Uncertainty (95/95) +0.00300
(' TOTAL Uncertainty (statistical) +0.00709 I9
( [ ((tolerance;.. .or... uncertainty;)2 )
ki-1
( Final K en Including Uncertainties & Tolerances: 0.94876
[
[
m Point Beach Spent Fuel Racks 16
I
( Table 3. Point Beach Spent Fuel Rack IFBA Requirement 23 5 U l.0X IFBA 1.5X IFBA 2.0X IFHA
( Enrichment Rods In Rods In Rods In (w/o) Assembly Assembly Assembly 4.60 0 0 0 5.00 16 12 8
(
(
(
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f Point Beach Spent Fuel Racks 17 I
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- ,. . . .. m s ;
- D ::D '
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, BORAFLEX m m
- m, 9.938" -
8.000" WIDE 0.100" THICK WITH 0.021" SS BOTH SIDES
(
Figure 1. Point Beach Spm Fuel Rack Cross-Section View of Array of Storage Cells Point Beach Spent Fuel Racks 18
l 16 1.0X IFBA Loading 1.5X IFBA Loading 2.0X IFBA Loading 12
/
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$ /
i# ACCE PTABL E
/
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- ~
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[ 4.60 4.70 4.80 4.90 5.00 235 U Enrichrnent (w/o) f Figure 2. Poirt Beach Spent Fuel Rack IFBA Requirement Point Beach Spent Fuel Racks 19
l
{
{-
0.00 f
\
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x
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r t
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0 500 1000 1500 2000 Soluble Boron Concentration (ppm) f Figure 3. Point Beach Spent Fuel Rack Soluble Horon Worth Point Beach Spent Fuel Racks ,
20
Bibliography
- 1. Pickard, Lowe and Garrick, Inc., Analysesfor Storage ofHigher Enrichment Westinghouse 11:14 Optimized Fuel Assemblies With or Without Natural Uranium Ends in the Point Beach New Fuel and Spent Fuel Storage Racks, April 1988.
- 2. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees from B. K. Grimes, OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14,1978.
- 3. Newmyer, W.D., Westinghouse Spent FuelRack Criticality Analysis Methodology, WCAP-14417, June 1995.
- 4. W.A.Boyd and D.E.Mueller (Westinghouse NFD), Efects ofPoison PanelShrinkage and Gaps on Fuel Storage Rack Reactivity, American Nuclear Society Transactions, Volume 56, pages 323-324, June 1988.
- 5. Electric Power Research Institute (EPRI), Boraflex Test Results andEvaluation, EPRI TR-101986, Interim Repon, February,1993.
Point Beach Spent Fuel Racks Bibl-1 -
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