NRC 2009-0057, Response to Request for Additional Information License Amendment Request 247 Spent Fuel Pool Storage Criticality Control

Response to Request for Additional Information License Amendment Request 247 Spent Fuel Pool Storage Criticality Control

ML091420436
Person / Time
Site:	Point Beach
Issue date:	05/22/2009
From:	Meyer L Nextera Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2009-0057
Download: ML091420436 (19)

Text

May 22,2009 NRC 2009-0057 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Point Beach Nuclear Plant, Units Iand 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Response to Request for Additional lnformation License Amendment Request 247 Spent Fuel Pool Storage Criticalitv Control References (1) FPL Energy Point Beach Letter to NRC, License Amendment Request 247, Transmittal of Changes to Technical Specifications re: Spent Fuel Pool Storage Criticality Control, dated July 24, 2008 (ML082240685)

(2) FPL Energy Point Beach Letter to NRC, Supplement to License Amendment Request Number 247, Spent Fuel Pool Storage Criticality Control, dated September 19, 2008 (ML082630114)

(3) FPL Energy Point Beach Letter to NRC, Response to Request for Additional Information, License Amendment Request 247, Spent Fuel Pool Storage Criticality Control, dated April 14, 2009 (ML091050499)

(4) NRC letter to FPL Energy Point Beach, Point Beach Nuclear Plant, Units 1 and 2 - Request for Additional lnformation from Reactor Systems Branch Related to License Amendment Request No. 247 Spent Fuel Pool Storage Criticality Control, dated April 22, 2009 (TAC Nos. MD9321 and MD9322) (ML09090067)

NextEra Energy Point Beach, LLC (formerly known as FPL Energy Point Beach, LLC) submitted a proposed license amendment request for Commission review and approval pursuant to 10 CFR 50.90 for the Point Beach Nuclear Plant (PBNP), Units Iand 2 (Reference I ) . The proposed amendment revises the licensing basis to reflect a revision to the spent fuel pool (SFP) criticality analysis methodology. The revised criticality analysis for the SFP storage racks credits burnup, integral fuel burnable absorber (IFBA), Plutonium-241 decay, and soluble boron, where applicable. FPL Energy provided a supplemental response (Reference 2) containing additional quantitative information to support the fidelity of key methodology aspects described in Reference ( I ).

On April 14, 2009, a teleconference was held between NRC and NextEra Energy Point Beach (NextEra) personnel to discuss additional information that was requested by the Commission to enable further review of the application. During the teleconference, NextEra stated that the response to the request for additional information would be submitted within 30 days of receipt.

On May 7,2009, a teleconference was held between NRC and NextEra personnel to discuss PBNP response to Question 1 of Reference (3) on the PBNP boraflex surveillance program.

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, W 54241

Document Control Desk Page 2 During the teleconference it was agreed that NextEra would clarify the response and include it in the Reference 4 response to request for additional information.

Enclosure Iof this letter provides the NextEra response to the request for additional information in Reference (4). provides the clarifying information of the PBNP boraflex surveillance program.

This supplement does not affect the no significant hazards conclusion, as defined in 10 CFR 50.92, provided in Reference (I).

This letter contains no new commitments and no revisions to existing commitments.

In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Wisconsin Official.

I declare under penalty of perjury that the foregoing and enclosed information is true and correct.

Executed on May 22,2009.

Very truly yours,

~ e x t ~ r a e ~I e aLLC c h ,

Site Vice President Enclosures cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE I NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS IAND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 247 SPENT FUEL POOL STORAGE CRITICALITY CONTROL The following information is provided by NextEra Energy Point Beach, LLC (NextEra) in response to the NRC staff's request for additional information dated April 14, 2009.

Question I: Code Validation Section 1.4.2 of Enclosure 6 of the letter dated July 24, 2008 (ADAMS Accession No. ML082240685), discusses the validation of the SCALE-PC code used in criticality calculations.

To allow the NRC staff to evaluate the adequacy of the validation, please provide the following additional information:

a. Discuss andjustify the method you used to select the benchmarks identified in Tables 1-1 and 1-2. For example, what parameters were considered to correlate the benchmarks to the systems being analyzed? What ranges were considered for those parameters?

b. Please provide additional details characterizing the benchmarks in terms of the parameters cited in Question l a above, or submit References 9 through 12 of WCAP-1654 1-P, Revision 2. Currently, the submittal lacks sufficient information to evaluate the applicability of the benchmarks to the systems being analyzed.

c. Document andjustify the area of applicability for the benchmarks.

d. Describe andjustify any statistical and trending analyses performed to support the determination of the bias and bias uncertainty.

e. How did you account for the measurement uncertainties for the benchmarks?

NextEra Response Point Beach Nuclear Plant (PBNP) uses low enriched uranium fuel in a water moderated system at relatively low temperatures, with boron as an absorber. The primary structural material of the PBNP spent fuel pool racks is SS-304. The critical benchmarks have similar fuel, absorber, moderator, and structural materials. As outlined in NUREGICR-6698, the important parameters for valid benchmarking are the fissile isotope, enrichment of the fissile isotope, fuel density, fuel chemical form, the types of neutron moderators and reflectors present, the ratio of moderator to fissile isotope, neutron absorbers, and physical configurations (i.e., geometry). Table 2.3 of NUREGICR-6698 summarizes the most important parameters and gives guidance as to how to define areas of applicability.

NUREGICR-6698 notes that, "perhaps the best source of critical benchmarks is found in the International Handbook of Evaluated Criticality Safety Benchmark Experiments for the Nuclear Energy Agency of the Organization for Economic Co-operation and Development (OECD-NEA).

The critical experiments described in this handbook have been found by the ANSIIANS-8 Subcommittee for NCS to be rigorously peer reviewed and should be accepted as refereed.. ."

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Document Control Desk Page 2 It is important to note that the 30 experiments used in the Benchmarking Suite presented in WCAP-16541-P, Revision 2, are included in the Handbook.

The benchmarks identified in Tables 1-1 and 1-2 of WCAP-I6541-P, Revision 2, have been the standard set of benchmarks considered for code validation by Westinghouse for spent fuel pool criticality safety analysis for many years.

A comparison of relevant parameters is shown in Table I.The first column gives a description of the parameter and the unit of measurement if applicable. The second column gives the range of each parameter for the benchmark experiments. The third column gives the range of each parameter as used in the PBNP analysis. The fourth column paraphrases the guidance given in Table 2.3 of NUREGICR-6698 relating to the area of applicability for each parameter.

The final column specifies if the parameter, as used in the PBNP analysis, is completely covered by the benchmark as defined by the area of applicability guidance. If the PBNP specific parameter is covered by the area of applicability as defined by Table 2.3 of NUREGICR-6698, the final column will list "Yes" if any part of the PBNP parameter is not covered; it is addressed in the discussion below the table.

Table I Comparison of Select Parameters between Point Beach Analysis and Benchmark Suite NUREGI Point Beach CR-6698 Parameter Benchmark Analysis Guidance Applicable Fissile lsotope 23.

5 11 23.5~

Same Yes Enrichment of Fissile 0-2wIo: 51% See Isotope (wt%) 2.5, and 4.31 I.33 - 5.0 2-5wlo: 51.5% Discussion Same or Fuel Chemical Form uo2 uo2 justified Yes See Temperature (K) 290 - 299 283 - 355 550 Discussion Neutron Moderators Same or and Reflectors Hz0 Hz0 justified Yes Same or Neutron Absorbers Boron, SS-304 Boron, SS-304 justified Yes Physical Thin fuel rods Thin fuel rods As similar as configurations in water in water possible Yes For the lowest fresh enrichment credited in the WCAP-16541-P, Revision 2, analysis (1.33 wt% 2 3 5 ~Table

), 2.3 of NUREGICR-6698 suggests that the benchmark should include experiments with enrichments between 0.33 - 2.33 wt% 2 3 5 ~ . However, the lowest enrichment in the benchmark suite is 2.5 wt% 2 3 5 ~ .

The upper end of the temperature range is 355 K. Table 2.3 of NUREGICR-6698 suggests that the benchmark should include experiments with temperatures between 305 - 405 K. The highest temperature in the benchmark suite is 299 K.

The basis for acceptability of the enrichment and temperature values being outside the range of applicability of the benchmarks is provided by the cross-section library being used. A key purpose of benchmarking is to validate the cross-section library for the proposed application.

As discussed in Section 1.4 of WCAP-16541-P, Revision 2, the 44-group Evaluated Nuclear Page 2 of 19

Document Control Desk Page 2 Data File Version 5 (ENDFIB-V) was used for the PBNP analysis. NUREGICR-6686 states that the ENDFIB-V library is particularly suited for light water reactor lattice applications; Section 7.1 notes that the library was developed for use in the analysis of fresh and spent light water reactor fuel.

The Shapiro-Wilks test for normality was applied to both the 44 and 238 group library results of the benchmark suite presented in Tables 1-1 and 1-2 of WCAP-16541-P, Revision 2. Both data sets pass the Shapiro-Wilks test and can thus be considered to have a normal distribution.

The calculated kernusing the 44-group cross-section library, of each of the thirty benchmark experiments was plotted versus the H/'~~u value of the experiment, and is shown in Figure I Figure 1 Calculated keR(44-groupCross-Section Library) vs. H / * ~ ~ U The data is well distributed and no trend is apparent as a function of H/'~~u.The results using the 238-group cross-section library showed similar distributions.

The measurement uncertainties in the benchmarks were not explicitly accounted for. This is justified by the fact that the experimental uncertainties are small and the methodology bias and uncertainty calculated with the benchmark data is comparably large.

Fission product critical data is not available. The effect of fission products is accounted for in the burnup uncertainty. In response to the acceptance review questions, cover letter Reference (2), the burnup uncertainty methodology was changed to the 5% decrement method.

This method is sufficiently conservative to account for the lack of fission product critical data.

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Document Control Desk Page 2 Question 2: TolerancelUncertaintv Calculations a) Why did you not include the fuel pellet diameter uncertainty in "All-Cell" and "I-out-of-4, 5.0 w/o Fresh with no IFBA [integral fuel burnable absorber]" when you included it in "I-out-of-4, 4.0 w/o Fresh with IFBA" case?

b) How do you determine what manufacturing tolerances to include in the uncertainty study?

c) You appear to assume that the sum of biases and uncertainties for a given configuration remains constant for the different combinations of enrichment, burnup, decay period, and number of IFBAs (for the "I-out-of-4, 4.0 w/o Fresh with IFBA"). Please substantiate this assumption quantitatively.

NextEra Response The fuel pellet diameter for Standard fuel was modeled as a bounding value, meaning the value modeled was greater than the nominal diameter plus the manufacturing tolerance. Because a bounding value was modeled, no uncertainty needed to be evaluated in configurations that only considered Standard fuel. The All-Cell configuration only considered Standard fuel as discussed in the response to Question 3. However, in response to this question, the pellet diameter uncertainty effect was explicitly calculated at both the positive and negative extremes of the manufacturing tolerance from nominal and is reported below.

Manufacturing tolerances that have a statistically significant effect on the calculated eigenvalue and a physical basis are included in the uncertainty rackup. This includes parameters associated with the individual fuel pin characteristics that can propagate across a given fuel assembly and even across an entire fuel batch. Such parameters include fuel enrichment, fuel pellet diameter, and cladding thickness and diameter among others. Neglecting spacer grids has been shown to be conservative.

Manufacturing tolerances on the spent fuel pool rack are also considered. This includes parameters associated with rack pitch, wall thickness, and rack cell inner diameter. If the rack contains neutron absorbing material, and the criticality analysis takes credit for that material, tolerances associated with the neutron poison must also be evaluated. The position of the assembly within the cell is also considered.

The fuel rod pitch tolerance has historically been neglected for a combination of reasons. The primary reason is that the real variability of the pin pitch is small and random. These small pitch changes would yield small reactivity effects. The effect of variability in pin pitch would introduce some slight disarray in the fuel assembly. This disarray would act to lower reactivity slightly as an ordered array is more reactive than a disordered array. The amount of pin pitch variability is also necessarily small given the established tolerances on grid strap parameters and overall assembly dimensions.

The sum of biases and uncertainties for a given configuration does not remain constant over the entire range of allowable enrichments, burnups, decay periods, and number of IFBA rods. Care was taken when determining the conditions at which the biases and uncertainties would be calculated to ensure that the resulting sum would be conservative for the range of conditions over which it is applied. Tables 2 and 3 demonstrate that the total biases and uncertainties documented in WCAP-I6541-P, Revision 2, are conservative.

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Document Control Desk Page 2 Table 3-4 of WCAP-16541-P, Revision 2, reports the biases and uncertainties calculated for the All-Cell configuration. With the exception of the temperature bias, these tolerances were calculated by modeling fresh Standard fuel at the highest allowable fresh fuel enrichment.

Standard fuel was modeled in this configuration because it is the design basis fuel assembly for low enrichment fresh fuel and for all burnt fuel, as discussed in the response to Question 3. The temperature bias was calculated with 5.0 wt% fuel depleted to 25,000 MWDIMTU burnup as discussed in cover letter Reference (2). To show that these calculations are conservative for the range of burnups and decay times over which they are applied, two additional bias and uncertainty rackups were created for the All-Cell configuration; one for 5.0 wt% fuel depleted to 35,000 MWdIMTU with zero decay time and one for 5.0 wt% fuel depleted to 35,000 MWDIMTU with 20-year decay time. SCALE 5.1 was used when creating the new rackups. SCALE 5.1 has gone through the same verification and validation as SCALE-PC. The methodology biases and uncertainties are properly accounted for between the two versions of SCALE allowing for a valid comparison of the total bias and uncertainty terms. The results are shown in Table 2 along with the WCAP-I 6541-P, Revision 2, results for convenience.

Table 2 Bias and Uncertainty Calculations, All-Cell Storage Configuration I WCAP-I 6541-P 1 35 GWdIMTU, 1 35 GWdIMTU, I I I Revision 2 1 0 vrs decav / 20 vrs decav 1 Case Description aeff aeff at#

Increased Fuel Enrichment 0.00692 0.00688 0.00688 Increased Pellet Diameter -- 0.00073 0.00013 Decreased Pellet Diameter -- 0.00025 -0.00001 Decreased Clad OD &

Thickness 0.009 69 0.00155 0.001 19 Decreased Cell Pitch 0.00171 0.001 50 0.00095 Decreased Rack Thickness 0.00309 0.00208 0.00222 Off-Center Assembly position 0.00708 0.00671 0.00621 Burnup Uncertainty 0.00781 0.00781 0.00781 Methodology Uncertainty 0.00639 0.00677 0.00677 Statistical Sum of Uncertainties 0.01467 0.01 445 0.01414 Methodology Bias 0.00310 0.00318 0.0031 8 0.01036 0.00932 0.00876 Sum of Uncertainties and Biases 0.02813 0.02695 0.02608 For the 35,000 MWDIMTU burnup 0 decay case, the result for the decreased pellet diameter is not statistically significant therefore, it is not included in the total. For the 35,000 MWDIMTU burnup 20-year decay case, neither increasing nor decreasing the pellet diameter gives statistically significant results so neither is included in the total. While there is both positive and negative variation in individual uncertainty terms, it can be seen that the total sum of uncertainties and biases as reported in WCAP-I 6541-P, Revision 2, bounds the range of burnups and decay times over which the uncertainties are applied.

The biases and uncertainties for the 1-out-of-4 4.0 wt% with IFBA configuration were calculated with 4.0 wt% fresh OFA fuel and no IFBA in one cell, and low enriched Standard fuel representing the burnt fuel in the other 3 cells. To show that the sum of uncertainties and Page 5 of 19

Document Control Desk Page 2 biases is conservative over the range of number of IFBA for which it is applied, two additional bias and uncertainty rackups were created for the I-out-of-4 4.0 wt% with IFBA configuration, one modeling 4.0 wt% fresh fuel and 16 1.5X IFBA rods, and one modeling 4.0 wt% fresh fuel and 32 1.5X IFBA rods. The results are shown in Table 3 along with the WCAP-16541-P, Revision 2, results for convenience.

Table 3 Bias and Uncertainty Calculations, I-out-of-4,4.0 wt% with IFBA Storage Configuration I WCAP-16541-P / 4.0 wt% 4.0 wt%

Revision 2 16 IFBA 32 IFBA Case Description k e f f meff k, Increased Fuel Enrichment 0.00549 0.00530 0.00524 Increased Pellet Diameter 0.00125 0.00074 0.00074 Decreased Pellet Diameter -- -0.00034 -0.0001 7 I Decreased Clad OD & I I I I Thickness 0.00198 0.001 10 0.00129 Decreased Cell Pitch 0.00146 0.00148 0.00177 Decreased Rack Thickness 0.00201 0.001 96 0.00230 Off-Center Assembly position 0.00420 0.00336 0.00358 Burnup Uncertainty 0.00589 0.00589 0.00589 Methodology Uncertainty 0.00644 0.00677 0.00677 I Statistical Sum of I I I I Uncertainties 0.01 165 0.01 130 0.01 147 Methodology Bias 0.00310 0.00318 0.00318 Pool Temperature Bias 0.00852 0.00769 0.00837 Sum of Uncertainties and I Biases 0.02327 0.0221 7 0.02302 Again, there is both positive and negative variation in individual uncertainty terms, however, it can be seen that the total sum of uncertainties and biases as reported in WCAP-16541-P, Revision 2, bounds the range of number of IFBA over which the uncertainties are applied.

Question 3: Bounding Fuel Desian a) In Section 1.5, you state that the Standard fuel design is bounding for spent fuel and OFA is bounding for fresh. Please quantitativelyjustify that this assumption is valid for all anticipated storage configurations and burnup/enrichment combinations at Point Beach.

b) In Section 3.2, you state, "Westinghouse standard fuel assembly design was modeled as the design basis fuel assembly to represent typical fresh and depleted fuel assemblies residing in all of the fuel assembly storage configurations." Does this contradict the statements in Section 1.5?

c) You also state "checkerboard storage configuration utilize the OFA fuel design." What do you mean by "checkerboard?" Are you referring to the Iout of 4 configuration?

Document Control Desk Page 2 NextEra Res~onse The optimized fuel assembly (OFA) is designed for improved in-reactor performance relative to the Standard assembly, and given the same assumed in-reactor conditions will be less reactive than a Standard assembly at realistic discharge burnups. Table 4 demonstrates the difference in calculated eigenvalue between the Standard and OFA fuel assemblies at 3.0 wt%, 4.0 wt%,

and 5.0 wt% enrichment in the All-Cell storage configuration.

Table 4 Ak [kSTD koFA]at 3.0 wt%, 4.0 wt%, and 5.0 wt% as a Function of Burnup Burnu0 I 3.0 wt% 4.0 wt% 5.0 wt%

For the 5 wt% case, the Standard assembly becomes more limiting at 15,000 MWDIMTU. In all the configurations with 5.0 wt% initial enrichment, the burnups used to determine the burnup limit are at or above 15,000 MWDIMTU. These results also confirm the conservatism of using Standard fuel as the design basis fuel assembly for spent fuel and OFA as the design basis fuel assembly for fresh fuel when modeling 5.0 wt% initial enrichment, as in the I-out-of-4, 5.0 wt%

no IFBA configuration.

For the 3.0 wt% and 4.0 wt% cases, Table 4 shows Standard fuel as more reactive than OFA at all times in life including fresh fuel. These results confirm the conservatism of using Standard fuel as the design basis fuel assembly for spent fuel when modeling 4.0 wt%, or less, initial enrichment, but calls into question the appropriateness of modeling 4.0 wt% fresh OFA in the I-out-of-4, 4.0 wt% Fresh with IFBA configuration. The four cases shown in Figure 2 were used to determine the design basis fresh fuel assembly for the I-out-of-4, 4.0 wt% case.

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Document Control Desk Page 2 1.6 UT Oib 1.6 UT 96 5 UT 96,STD, 5 I\* Yo,STD, Fresh STD Fresh STD 55,O00 hf'IVol%f T U 55,000hnVdlMTU 4.0 UT 9'0 5 U* 96,STD, 4.0 uft 9.b 1.6 U* % Fresh STD 55,000MLVdthfTU Fresh STI3 Fresh STD 1.6 UT 56 1-6ilitOib 5 U* 58, STD, 5 U% ?6,STD, Fresh STD Fresh STD 55.000 M%7d!hafTU 55,000M'Ii7d/MTU 4.0 x7.T 9'0 f.61r.tOib 4.0 wt Oib 5 wt %, STD, Fresh OFA Fresh S'hl) Fresh OFA 55,000M'IVdt'MTU Figure 2 Test Cases to Determine the Design Basis Fresh Fuel Assembly in the I-out-of-4,4.0 wt% Fresh Configuration.

The highest allowable fresh enrichment in 3 of the 4 cells is I.6 wt% according to Table 3-14 in WCAP-16541-PI Revision 2; and the highest enrichmentlburnup combination used was 5.0 wt%

at 55,000 MWDIMTU burnup so this covers the range of enrichmentlburnup combinations. The single 4.0 wt% assembly was modeled as both an OFA and a Standard assembly for each enrichmentlburnup combination; the results are shown in Table 5.

Table 5 - Ak Results for the I-out-of-4 4.0 wt% Configurations Shown in Figure 2 4.0 wt% Fresh 1 Description Ak [ ~ S T -D~ O F A ]

1.6 wt%, 0 MWdIMTU -0.00299 5.0 wt%. 55.000 MWdIMTU -0.00360 The results in Tables 4 and 5 confirm that the design basis fuel assemblies used in the WCAP-6541-PI Revision 2, analysis are appropriate and conservative. Furthermore, using Standard fuel as the design basis fuel assembly for spent fuel and OFA as the design basis fuel assembly for fresh fuel is consistent with the design basis fuel assemblies used in the previous, Page 8 of 19

Document Control Desk Page 2 approved, PBNP criticality analysis. It is also consistent with other approved criticality analysis for 14x14 fuel lattices (References Iand 2).

The statement referred to in part b of this question refers to the All-Cell storage configuration.

As discussed in the response to Question 2, the bias and uncertainty calculations reported in WCAP-16541-PI Revision 2, with the exception of the temperature bias, were done with fresh fuel. For the All-Cell configuration this was modeled as fresh Standard fuel because the uncertainties needed to be applied to spent fuel. Also, the results in Table 4 show that Standard fuel becomes more limiting relative to OFA as enrichment decreases.

In the two "I-out-of-& or "checkerboard" configurations the single fresh fuel assembly is modeled as OFA fuel and the 3 burnt assemblies are modeled as Standard fuel.

The term "checkerboard" as used in WCAP-I 6541-P, Revision 2, refers generically to the I-out-of-4 storage configurations.

Question 4: IFBA Depletion Effects a) Letter dated September 19, 2008 (ADAMS Accession No. ML082630114), provided a response to the staff acceptance review. You state in response to Question 4, that the "results demonstrate that including the residual 'OB provides sufficient reactivity margin to account for the spectral hardening caused by the presence of IFBA during the depletion. "

This statement conflicts with NUREG/CR-6760 which states that ",.. the ilk values become positive for fuel assembly designs containing IFBA rods but remain negative for gadolinia-bearing fuel assembly designs. " NUREG/CR-6760 further states that "... analyses show that there is a negative residual effect for gadolinia-bearing fuel but no such effect for fuel designs with IFBA rods." Please resolve the difference in conclusions between your analysis and that of NUREG/CR-6760.

b) What enrichment was used for the calculations in the table titled, "Results of Calculations with IFBA Present During Depletion?" Please justify that the results are based on the limiting enrichment and burnup combinations.

NextEra Response The analysis presented in the September 19, 2008 letter (cover letter Reference (2))) was performed using site specific information for the fuel assembly design, core operating parameters, and IFBA designs used at PBNP. This is in contrast with the generic analysis of 17x17 fuel presented in NUREGICR-6760. Not enough information is available in NUREGICR-6760 for the Licensee or Vendor to know exact differences between the two analyses. However, an attempt will be made to call out the differences that are apparent. The results presented in the cover letter Reference (2), and additional results below are applicable to the PBNP spent fuel pool criticality safety analysis. The results presented here are not necessarily representative of other fuel lattices, assembly designs, IFBA designs or core operating parameters outside of PBNP.

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Document Control Desk Page 2 Tables 6 and 7 contain modeling information from Section 3.3 of NUREGICR-6760. A column is added to each table to show the values used in the PBNP analysis.

Table 6 Summary of Parameters Used for the Depletion Calculations (Table I of section 3.3 of NUREGICR-6760)

Parameter NUREGICR-6760 Point Beach Analysis Moderator Temperature (K) 600 561.3 Fuel Temperature (K) I000 1079.3 Fuel Density (g/cm3) 10.44 (UOa) 10.69 (UOn)

Clad Temperature (K) 600 593.7 Clad Density (g/cm3) 5.78 (Zr) 6.56 (Zircaloy-4)

Power Density (MWIMTU) 60 37 Moderator Boron Concentration (ppm) 650 800 Most of the calculations in NUREGICR-6760 were performed assuming a uniform burnup profile. The PBNP analysis used the axial power distribution shown in Figure 3-9 of WCAP-16541-P, Revision 2; the fuel, moderator, and cladding temperatures varied as a function of power. The values shown in Table 6 correspond to the values at a relative power of I.O.

The footnote to Table Iof NUREGICR-6760 states that cases were also calculated using a power density of 30 MWIMTU and that the Ak results were not sensitive to variations in power density.

A significant difference is the moderator soluble boron concentration. By assuming a lower value, the results in NUREGICR-6760 will exaggerate the effect of the IFBA. Per NUREG-6665, using a high soluble boron concentration during depletion is conservative due to spectral hardening. The IFBA also provides spectral hardening, but the effect is localized. Using 800 ppm for a cycle average value is high for the PBNP units. The NUREGICR-6760 approach of using a lower soluble boron concentration will indicate a larger IFBA worth, but may underpredict the overall reactivity.

Standard fuel dimensions are presented in Table 7 because Standard fuel is the design basis assembly for burnt fuel.

Table 7 Fuel Assembly Specifications (Table 2 of Section 3.3 of NUREGICR-6760)

NUREGICR-6760 Point Beach Analysis Parameter Rod Pitch (cm) 1.260 1.412 Assembly Pitch (cm) 21.5 19.8 Cladding Outside Diameter (cm) 0.8898 1.0719 Cladding Inside Diameter (cm) 0.8001 0.9484 Pellet Outside Diameter (cm) 0.7840 0.9294 Guidellnstrurnent Tube Outside Diameter (cm) I.204 1.369 1 I.072 Guidellnstrument Tube Inside Diameter (cm) 1.124 1.2827 10.94996 Array Size 17x17 14x14 Number of Fuel Rods 264 179 Number of Guidellnstrument Tubes 25 17 Page 10 of 19

Document Control Desk Page 2 NUREGICR-6760 examines 17x17 assemblies containing 80,104, and 156 IFBA rods with boron loadings of 4.57 and 2.355 mg 10~/inch.The PBNP analysis uses 14x14 assemblies containing 120 IFBA rods with a 1.5X IFBA loading. The PBNP analysis IFBA pattern contains a significantly higher percentage of IFBA rods than those considered in NUREGICR-6760 (67% for PBNP versus 59% in NUREGICR-6760).

In the table titled "Results of Calculations with IFBA Present During Depletion," in the cover letter Reference (2), the calculations were performed with 5.0 wt% enriched fuel. The study reported in cover letter Reference (2) was performed with IFBA modeled over the entire axial length of the fuel.

Section 3.3.5.5 of NUREGICR-6760 states that modeling a shorter IFBA stack can result in larger differences in calculated eigenvalues between cases depleted with and without IFBA present. Tables 10 and I 1 of NUREGICR-6760 show the Ak effects when a non-uniform axial burnup profile is modeled with IFBA modeled over the entire axial length of the fuel and with IFBA modeled over 120 inches, centered axially in the fuel rod. The more realistic IFBA model, 120 inches centered axially, is more limiting.

Therefore, the PBNP specific study was re-performed modeling a 120 inch IFBA region centered axially in the fuel assembly. Results of this new study are presented and discussed below.

Some similar trends are seen between the two analyses. Figure 41 in Section 3 of NUREGICR-6760 shows Ak values versus burnup for the 156 IFBA pattern, with varying 2 3 5 ~

enrichments. Table 8 below shows similar trends for the 120 IFBA pattern, with varying 2 3 5 ~

enrichments in the All-Cell configuration of the PBNP analysis. The values shown in this table are plotted in Figures 3 and 4.

Table 8 Ak in the All-Cell Confinuration 4.0 $% 5.0 wt%

MWdlMTU Ak F(IFBA) - k(no IFBA)] Ak [~(IFBA)

- k(no IFBA)]

5,000 -0.07924 -0.08292 15.000 -0.01 332 -0.02247 The data was fit using a fifth order polynomial and the difference in eigenvalues was found for the All-Cell configuration burnup limits for 4.0 wt% and 5.0 wt%, reported in Table 4-1 of WCAP-16541-P, Revision 2. The difference in eigenvalues at the burnup limits is shown in Table 9. For this site specific analysis, at the burnup limits specified in WCAP-16541-P, Revision 2, the residual IFBA is enough to overcome the effect of spectral hardening. The results confirm that neglecting the presence of IFBA is conservative in the All-Cell storage configuration.

Table 9 Ak in the All-Cell Configuration at the Burnup Limit Limiting Burnup 4.0 wt% Limiting Burnup 5.0 wt%

MWdIMTU Ak [~(IFBA) - k(no IFBA)] MWdIMTU - k(no IFBA)]

Ak [~(IFBA) 18,475 -0.0051 2 27,349 -0.00006 Page 11 of 19

Document Control Desk Page 2 Figure 3 Ak in the All-Cell Configuration All Cell Configuration with 120 IFBA rods 20 30 Burnup (GWdlMTU)

Figure 4 Close-up of the Area of Interest in Figure 3 I

All Cell Configuration with 120 IFBA rods 20 25 30 Burnup (GWdlMTU)

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Document Control Desk Page 2 In the All-Cell configuration all four of the assemblies have been depleted with IFBA thereby maximizing the spectral hardening effect. However, since the All-Cell configuration also has the lowest burnup limits, the I-out-of-4 configurations were examined.

The All-Cell case and NUREGICR-6760 demonstrate that the effect at 4.0 wt% is stronger than at 5.0 wt% so the burnt assemblies modeled in the following I-out-of-4 cases were modeled as Standard assemblies at 4.0 wt% initial enrichment with 120 IFBA rods that contain 120 inches of IFBA centered axially. Comparison of IFBA versus non-IFBA eigenvalues were performed for burnup on the Standard fuel assemblies from 5,000 to 55,000 MWDIMTU.

For the I-out-of-4, 5.0 wt% Fresh no IFBA configuration, the three burnt assemblies were modeled as described in the preceding paragraph and the single fresh assembly was modeled as an OFA assembly at 5.0 wt% enrichment containing no IFBA rods.

Three additional cases were considered for the I-out-of-4, 4.0 wt% Fresh with IFBA configuration. The three burnt assemblies were modeled as previously described. The single fresh assembly was modeled in the first case as an OFA assembly at 4.0 wt% enrichment containing 16 IFBA rods. The second case was modeled as an OFA assembly at 4.0 wt%

enrichment containing 32 IFBA rods. The third case was modeled as an OFA assembly at 5.0 wt% enrichment containing 32 IFBA rods.

Because of the higher burnup limits in the I-out-of-4 configurations, there is no longer enough residual IFBA to offset the effect of increased plutonium production. Once again the Ak [k(IFBA) - k(no-lFBA)]

data was fit with a fifth order polynomial. Using the polynomial fits, the largest positive difference in eigenvalues was identified for each configuration, shown in Table 10.

Table 10 Largest Positive Difference in Eigenvalues for each Configuration I I-out-of-4, 5.0 wt% 1 I-out-of-4,4.0 w t I 1-out-of-4,4.0 wt% I I-out-of-4,5.0 wt% I I Fresh, No IFBA I Fresh, 16 IFBA I Fresh, 32 IFBA I Fresh, 32 IFBA I It can be seen from Table 10 that the largest positive difference in calculated eigenvalue for the I-out-of-4, 5.0 wt% Fresh, with no IFBA configuration is 0.00149. The response to Question 1 in the cover letter Reference (2) used 0.00302 of the 0.00500 Ak administrative margin in this configuration, which still leaves enough to cover the additional 0.00149 Ak.

It can be seen from Table 10 that the largest positive difference in calculated eigenvalue for the I-out-of-4, 4.0 wt% Fresh, with IFBA configuration is 0.00218. The response to Question 1 in the cover letter Reference (2) used 0.00203 of the 0.00500 Ak administrative margin in this configuration, which still leaves enough to cover the additional 0.00218 Ak. The remaining administrative margin for each configuration is shown in Table 11.

Table I 1 Remaining Administrative Margin for each Configuration Remainina Administrative Marain I All-Cell 0.00369 I-out-of-4, 5.0 wt% Fresh, No IFBA I 0.00049 I-out-of-4, 4.0 wt% Fresh, With IFBA I 0.00079 Page 13 of 19

Document Control Desk Page 2 This study showed trends similar to those reported in NUREGICR-6760, but was performed with PBNP specific conditions and concludes that analysis reported in WCAP-16541, Revision 2, remains conservative.

Question 5: Soluble Boron Credit Letter dated September 19,2008 (ADAMS Accession No. ML082630114), provided a response to the staff acceptance review. Response to Question 2 discussed the effect of "parallel" accounting method on the boron concentration required for accident conditions.

Please justify the effect of '~arallel"accounting method on the boron concentration required for nominal conditions.

NextEra Response WCAP-I 6541, Revision 2, reports 402 ppm as the soluble boron concentration, assuming 19.4 atom percent 'OB abundance, required to maintain keffless than or equal to 0.95 including all biases and uncertainties for nominal conditions.

lsotopics are not available at the burnup limits specified in WCAP-I 6541, Revision 2, so the available isotopics for burnups closest to the limits assuming 5.0 wt% initial enrichment were used. The limits associated with 5.0 wt% initial enrichment were used because this maximizes the required burnup. As burnup increases, soluble boron worth decreases due to the harder neutron spectrum. The isotopics at the burnup closest to, but less than the limit, are used because the effect of additional burnup on reactivity is expected to be larger than the reduction in boron worth due to the additional burnup. Therefore, if there is margin to the Upper Subcritical Limit, or target eigenvalue, for fuel that does not meet the burnup limit specified in WCAP-16541, Revision 2, it proves the required boron is conservatively high. Results using the isotopics at the burnup closest to, but over the burnup limit, are included to show that the reactivity effect due to additional burnup is larger than the reduction in boron worth, i.e., there is more margin to the target eigenvalue even though the soluble boron is worth slightly less.

The isotopics used in this analysis are summarized for the three configurations in Table 12.

Table 12 - Burnup Limits of Interest and lsotopics Used to Calculate Margin 1 Burnup Limit Specified in Configuration WCAP-1 6541, kevision 2 lsotopics Used All-Cell 27,349 MWdIMTU 25,000 I35,000 MWdlMTU I-out-of-4, 5.0 wt% Fresh, no IFBA 51,169 MWdIMTU 45,000 155,000 MWdIMTU I-out-of-4, 4.0 wt% Fresh, with IFBA 41,361 MWdlMTU 35,000 I45,000 MWdIMTU The three configurations were modeled as 2x2 infinite arrays as described in Section 3 of WCAP-16541, Revision 2. This is conservative by not accounting for leakage that would be present in the actual pool. WCAP-16541, Revision 2, reports 402 ppm as the required soluble boron concentration, assuming 19.4 atom percent 'OB abundance, so the moderator was modeled containing 400 ppm of soluble boron with 19.4 atom percent 'OB abundance which is close to, but conservatively less than, the required amount.

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Document Control Desk Page 2 The calculated eigenvalues were compared to the target eigenvalues documented in cover letter Reference (2) in response to Question 5. These target eigenvalues considered borated (648 ppm) biases and uncertainties and the 5% decrement method for burnup uncertainty documented in response to Question 1 in the same letter.

Margin is demonstrated by subtracting the calculated eigenvalue from the target eigenvalue; there is positive margin when the calculated value is less than the target. Margins to the target eigenvalue for each configuration at each burnup considered are shown in Table 13.

Table 13 Margin to Target Eigenvalue Burnup Margin to I I-out-of-4. 5.0 wt% Fresh. no IFBA I I AR nnn n ~IF~F~I I I

I-out-of-4, 4.0 wt% Fresh, with IFBA 35,000 0.01450 45.000 0.04023 Margin is shown to the target eigenvalue for every configuration, thereby justifying the "parallel" accounting method at nominal conditions.

The target eigenvalues are not expected to be significantly different at 400 pprn than at 648 ppm. The smallest amount of margin to the target eigenvalue is more than 20 times larger than the largest difference between the borated and unborated target eigenvalues shown in cover letter Reference (2).

As discussed in the response to Question 2 documented in cover letter Reference (2), this margin is the results of the conservatisms included in the soluble boron concentration equation shown below and in the method used to determine the soluble boron worth.

Where: SBCTotalis the total soluble boron concentration requirement (pprn)

SBC95195 is the soluble boron requirement for 95/95 keffless than or equal to 0.95 ( P P ~ )

SBCREis the soluble boron required to account for burnup and reactivity uncertainties (pprn)

SBCpAis the soluble boron required to offset accident conditions (pprn)

The reactivity uncertainty of the fuel assembly is accounted for in the burnup uncertainty included in the determination of the burnup limits. The soluble boron worth is also conservatively determined in the full pool model loaded with depleted fuel which is less sensitive to the addition of soluble boron. These two conservatisms, which are implicit in the Page 15 of 19

Document Control Desk Page 2 methodology used in the analysis presented in WCAP-16541-P, Revision 2, provide sufficient conservatism in the determination of the required boron concentration.

References:

I Letter to Dr Robert C. Mecredy (RG&E) from Guy S. Vissing (NRC), dated December 7, 2000 "R.E. Ginna Nuclear Power Plant - Amendment RE: Revision to the Storage Configurations Requirementswithin the Existing Storage Racks and Taking Credit for a Limited Amount of Soluble Boron" (ML003761578)

2. Letter to Thomas J. Palmisano (NMC) from Mahesh L. Chawla (NRC), dated February 5, 2006 "Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendments RE: 'Spent Fuel Pool Storage" (ML060250208)

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ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS IAND 2 PBNP BORAFLEX SURVEILLANCE PROGRAM As discussed in the cover letter, a teleconference was held between NRC and NextEra personnel to discuss PBNP response to Question Iof Reference (3). It was agreed that NextEra would formally respond to the staWs query on the boraflex surveillance program. That response follows.

Question I Reference 3, question I, where the licensee states "The Boraflex Surveillance Program described in the letter dated 10/23/96 was discontinued on 12/21/06. Commitments to implement a Boraflex Monitoring Program for the period of extended operation remain in place. "

Please clarify whether a Boraflex Surveillance Program is currently in effect (especially for the period of 12/21/06 to the beginning of the period of extended operation).

NextEra Response The PBNP Boraflex Surveillance Program includes blackness testing conducted every five (5) years. The blackness testing in 2006 was deferred. The next required test is areal density testing which must be performed prior to entering the period of extended operation in 2010.

Currently PBNP continues to:

I.Maintain a database that is updated each time fuel movements are performed

2. Monitor industry OE in accordance with PBNP Operating Experience program

3. Notify the NRC if the program is to be modified

4. Checkerboard fresh and spent fuel assemblies with burnup less than 38,400 MWDIMTU, and if significantly degraded Boraflex is found.

The current analysis bounds current and future gap formation. The trends of silica are monitored have not shown an upward acceleration, and remain steady around 19 ppm.

The following license renewal commitments were made and are documented in NUREG 1839:

I.Certain accelerated Boraflex panels will be areal density and blackness tested every two years during the period of extended operation.

2. The first Boraflex areal density testing of the Boraflex panels will be performed prior to the period of extended operation.

3. A new procedure to schedule and perform Boraflex areal desntiy and blackness testing will be created.

4. If silica sampling and trending indicates a boron areal density depletion trend to a value less than the acceptance criteria (i.e., maintaining the 5% subcriticality margin) prior to the next scheduled test, then an evaluation will be performed within the corrective action program and the frequency of blackness and areal density testing increased.

5. Corrective actions will be taken to ensure that the 5% subcriticality margin of the spent fuel racks in the SFP is maintained during the period of extended operation. Corrective actions will be initiated if the test results find that the 5% subcriticality margin cannot be maintained because of current or projected future degradation. Corrective actions may include, but are not necessarily limited to Reanalysis, Repair andlor Replacement.