L-08-100, Responses to Request for Additional Information in Support of License Amendment Request No. 204, Revision 1

Responses to Request for Additional Information in Support of License Amendment Request No. 204, Revision 1

ML080730245
Person / Time
Site:	Beaver Valley
Issue date:	03/11/2008
From:	Hubley E FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-08-100, TAC MD2378
Download: ML080730245 (23)

Text

FENOC FirstEnergyNuclear OperatingCompany Edward H. Hubley 724-682-4862 Director,Maintenance March 11,2008 L-08-1 00 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Responses to a Request for Additional Information in Support of License Amendment Request No. 204, Revision 1 (TAC No. MD2378)

By letter dated March 6, 2008, the U.S. Nuclear Regulatory Commission (NRC) issued a request for additional information (RAI) pertaining to License Amendment Request (LAR) No. 204, Revision 1. This LAR was submitted by FirstEnergy Nuclear Operating Company (FENOC) on December 21, 2007 by letter L-07-517 (Reference 1). The LAR proposes Technical Specification changes that incorporate the results of a new spent fuel pool criticality analysis that will permit utilization of vacant storage locations in the Beaver Valley Power Station Unit No. 2 spent fuel storage pool. The new criticality analysis was submitted by FENOC on July 26, 2007 by letter L-07-103 (Reference 2).

Attachment 1 contains the FENOC responses to the March 6, 2008 RAI. The regulatory commitments contained in this letter are listed in Attachment 2. Approval of the proposed amendment is requested by March 2008 to support the Unit No. 2 refueling outage scheduled for the spring of 2008. Once approved, the amendment shall be implemented within 30 days.

If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager- FENOC Fleet Licensing, at 330-761-6071.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March J/, 2008.

Sincerely,

Beaver Valley Power Station, Unit No. 2 Responses to a Request for Additional Information in Support of License Amendment Request No. 204, Revision 1 (TAC No. MD2378)

L-08-1 00 Page 2 Attachments:

1. Response to RAI Items

2. Regulatory Commitment List

Reference:

1. FENOC Letter L-07-517, License Amendment Request Number 204, Revision 1 (TAC Nos. MD2377 and MD2378), dated December 21, 2007

2. FENOC Letter L-07-103, Supplemental Information for License Amendment Request Nos. 333 and 204, (Revision 2 of WCAP-16518) (TAC Nos. MD2377 and MD2378), dated July 26, 2007 cc: Mr. S. J. Collins, NRC Region I Administrator Mr. D. L. Werkheiser, NRC Senior Resident Inspector Ms. N. S. Morgan, NRR Project Manager Mr. D. J. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

ATTACHMENT 1 L-08-100 Responses to RAI Items By letter dated June 14, 2006, as supplemented by letters dated July 20, July 26, and December 21, 2007, FirstEnergy Nuclear Operating Company (FENOC, licensee) requested an amendment to the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2) Technical Specifications (TSs). The proposed changes to the TSs would incorporate the results of topical report, (WCAP)-16518, which will permit utilization of vacant storage locations dictated by the existing TS storage configurations in the BVPS-2 spent fuel storage pool. By letter dated December 21, 2007, the licensee withdrew their license amendment request (LAR) for BVPS-1. Based on previous RAI questions (References 1 and 2) and information obtained from the audit, the Nuclear Regulatory Commission (NRC) staff request further information on the following items:

RAI Item 1 On January 25, January 28, February 1, February 4, February 11, and February 12, 2008, the NRC staff conducted an audit of Westinghouse calculations associated with the BVPS-2 spent fuel pool criticality analysis. Those calculations were documented as calculations notes, CN-CRIT-224 and CN-CRIT-244. Please provide Section 7 of CN-CRIT-244 to support the NRC staff's safety evaluation (SE) for the BVPS-2 SFP criticality LAR. Also, identify any portions that are replaced due to response to questions listed below.

RAI Item 1 Response The following response consists of a condensed non-proprietary version of Section 7 of Westinghouse Calculation Note Number CN-CRIT-244, Revision 0. The specific areas that were removed include the discussion of core operating conditions, bumup shape applicability less than 30 gigawatt day per metric ton uranium (GWd/MTU), the discussion of the SFP temperature bias, and the justification of the 5% assembly burnup uncertainty. These portions were removed from this response because they are addressed in the responses to the other RAI Items.

Soluble Boron Credit The total soluble boron credit needed to maintain the kjff less than 0.95 with 95% probability of a 95% confidence interval, including the worst case postulated accident, is determined using Equation 1, shown below. Separate terms are used to account for a 5% reactivity reduction, assembly reactivity uncertainties, and postulated accidents. The reactivity worth of each term is determined and then converted to a boron concentration using the calculated Beaver Valley Power Station Unit 2 (BVPS-2) soluble boron worth.

SBCTolaI z SBC95195 + SBCRE + SBCPA Equation (1)

Where: SBCTota! = total soluble boron concentration requirement, SBC 95/95 = soluble boron concentration required to lower SFP reactivity by 5%,

SBCRE = soluble boron concentration required to offset assembly reactivity uncertainties, SBCPA = soluble boron concentration required to mitigate the limited postulated accident.

, L-08-100 Page 2 of 20 The first step in the soluble boron credit methodology is to determine the worth of the soluble boron in the spent fuel pool. The worth is conservatively determined by loading the entire pool with the "3x3" loading configuration with 5 w/o 235U assemblies depleted to 55,000 MWd/MTU burnup. As discussed in Reference 1, this configuration is selected because the depleted fuel has a harder (higher energy) neutron spectrum, which is less sensitive to the primarily thermal neutron absorption from boron-10 in the soluble boron. In other words, using depleted fuel for this calculation minimizes the incremental reactivity worth of the soluble boron, conservatively maximizing the soluble boron requirement. Calculations are performed for a range of boron concentrations and a polynomial is fit to this data. This polynomial is used to translate the reactivity worth determined for each SBC term into boron concentrations.

The second step in the methodology is to determine the reactivity worth associated with each term in Equation 1.

The first term, SBC 95 /95 is simply 5% (0.05 Akeff units). This value is derived from Reference 2, which states that the SFP must be subcritical with unborated water (ken < 1.0) and with soluble boron present, ker must not be more than 0.95 . This difference defines the 5% in reactivity.

The second term, SBCRE, accounts for assembly reactivity uncertainties. The SBCRE term was originally used to.account for the reactivity penalty inherent in reactivity equivalencing, but this practice is no longer used. This term is currently used to account for two specific uncertainties:

Bumup uncertainty - the burnup uncertainty is determined as the reactivity associated with a 5% decrease in assembly burnup. The reactivity impact of the 5% reduction in bumup is fuel storage configuration specific because of both the maximum credited burnup in that configuration and the reactivity response to the bumup change. The largest uncertainty is selected from all configurations analyzed.

• Depletion uncertainty - The depletion uncertainty is calculated as 1% reactivity per 30,000 MWd/MTU. The largest burnup credited in this analysis is used for determining this uncertainty.

The reactivity effects of the burnup and depletion uncertainties are combined algebraically to yield a bounding assembly reactivity uncertainty.

The third term, SBCPA, is the boron concentration necessary to mitigate the worst postulated accident scenario. The double contingency principle, detailed in References 3 and 4, allows the use of soluble boron to mitigate these other accidents, since a boron dilution event and the postulated accident scenarios are separate low-probability events. The specific events considered are provided in Reference 1. Each accident is considered from unborated pool conditions as this produces the highest reactivity in the accident condition. The reactivity insertion of the accident is calculated as the difference between the accident ken and the base case unborated keff.

The final step is the summing of the three boron concentration terms.

, L-08-100 Page 3 of 20 There are several conservatisms implicit to the Westinghouse soluble boron credit methodology, and these include the following:

1. The burnup uncertainty is included in the SBCRE term (as described above), but is also included in the overall criticality analysis bias and uncertainty rackup. Since the uncertainty is already accounted for, adding it in the SBCRE term is additional conservatism.

2. The algebraic treatment of the two portions of the SBCRE term is conservative relative to a statistical root-sum-square (RSS) combination of two independent uncertainties.

3. Biases and uncertainties are not recalculated in the presence of soluble boron. Westinghouse experience is that the sum of biases and uncertainties is nearly invariant to the presence of soluble boron and, in most cases, decreases slightly, as shown in the current criticality safety analysis of record for BVPS-2. Thus, using the biases and uncertainties from unborated conditions provides a slight conservatism for most analyses.

4. Conservatism exists in using a single boron worth determined with fuel depleted to the maximum credited burnup of any configuration. Particularly for an accident condition, which is typically limited by a misloaded fresh 5 w/o assembly, this single soluble boron worth value can be higher than that calculated for the entire spent pool model. The mitigation of the accident will be driven by the boron within the misloaded assembly, which will respond more than the depleted fuel for which the worth was calculated. This is the largest source of conservatism in the Westinghouse soluble boron credit methodology. While the limiting Reference I accident occurs in the "I -out-of-4 5.0 w/o at 15,000 MWd/MTU" configuration, a conservatively low boron worth determined with the "3x0" configuration (requiring increased burnup) is used to calculate the boron concentration necessary to mitigate the accident. If the lower burnup requirements for "1 -out-of-4 5.0 w/o at 15,000 MWd/MTU" configuration were used, then the soluble boron worth would increase and the soluble boron requirements would decrease.

Table I presents a series of soluble boron worth calculations performed in a similar method as described in Reference 1, with the exception that the maximum concentration considered is increased to 1000 ppm and SCALE 4.4 is used. The differential boron worth varies from 0.00017 Akeff/ppm at 200 ppm to 0.00014 Akef/ppm at 1000 ppm. Table 2 shows that the differential boron worth calculated in the limiting accident condition ("I-out-of-4 5.0 w/o at 15,000 MWd/MTU" configuration) is 0.00019 Akeff/ppm. Since a lower boron worth value was utilized in the Reference I analysis, the soluble boron requirements are conservatively higher than necessary.

Table 1. Soluble Boron Worth in the BVPS-2 Spent Fuel Pool Boron Integral Differential Concentration keff ( Boron Worth Boron Worth (ppm) (Akff) (Akinppm) 0 0.97081 0.00034 0 0 200 0.93619 0.00032 0.03462 0.00017 400 0.90586 0.00029 0.06495 0.00016 600 0.87832 0.00030 0.09249 0.00015 800 0.85389 0.00030 0.11692 0.00015 1000 0.83167 0.00027 0.13914 0.00014

, L-08-100 Page 4 of 20

5. The accident scenarios presented in Reference I were conservatively considered in a pool with no soluble boron present. The limiting postulated accident was determined in Reference I to be the misloading of a fresh 5 w/o assembly, and the reactivity of this assembly was maximized by considering it in the unborated condition. Table 2 presents analysis performed to demonstrate the conservatism inherent in this approach.

Table 2. Accident Condition kff at Different Soluble Boron Concentrations Boron Configuration Concentration krff a Accident Winm) Akff 3x3 - Base 0 0.97081 0.00034 _

3x3 - Misload 1.02579 0.00031 005498 3x3 - Base 441.8 0.89942 0.00031

• 3x3 - Misload 0.94910 0.00034 0.04968 I-out-of-4, 5 w/o, 15k - Base 0 0.95866 0.00025 I-out-of-4, 5 w/o, 15k - Misload 1.02096 ( 0.00033 0.06230 1-out-of-4, 5 w/o, 15k - Base 441.8 0.86213- 0.00023 1-out-of-4, 5 w/o, 15k - Misload 0.93692 (' 0.00029 0.07479

( The differential boron worth of the misload accident case is 0.00019 Aken/ppm The misload accident was considered for the two limiting fuel storage configurations identified in Reference 1 - "3x3" and "I-out-of-4 5.0 w/o at 15,000 MWd/MTU".

For each configuration, the base case and accident k~f were determined at both 0 ppm and 441.8 ppm soluble boron concentration. Per Reference 1, 441.8 ppm is the required minimum concentration not accounting for any accidents.

• For the "3x3" configuration, the accident Akeff decreases in the presence of soluble boron.

• For the "I-out-of-4 5.0 w/o at 15,000 MWd/MTU" configuration, the accident Akeff increases in the presence of soluble boron; however, the accident condition keffis more than 6% subcritical.

These calculations demonstrate that while in some cases it is more conservative to consider the accident condition initiating from the required boron concentration, the resulting accident kfr is significantly subcritical.

6. The three SBC terms are determined first as reactivities and then each is individually

("parallel" application) translated into boron concentrations. It has been commented that since the differential boron worth is reduced as boron concentration increases that a more conservative approach would be to sum the Akeff for each of the three terms ("serial" application), and then translate this overall worth into a boron concentration.

Table 3 presents a series of calculations performed in the limiting accident condition ("l-out-of-4 5.0 w/o at 15 GWd/MTU" configuration) at varying boron concentrations. The results show that at the total soluble boron requirement identified in Reference I (determined using the "parallel application" of boron worths), the SFP is subcriticalby almost 12% Akffg. Table 3 also shows that with "serial application" of the soluble boron worth the SFP is subcritical

, L-08-100 Page 5 of 20 by more than 14% Akeff. While more conservative, this clearly represents an unnecessary level of conservatism in this case.

Table 3. "I-out-of-4 5.0 w/o at 15 GWd/MTU" Accident Condition kff at Different Soluble Boron Concentrations Boron Concentration (ppm) 0 1.02096 0.00033 382.3 (accident mitigation) 0.94664 0.00030 441.8 (required without accident) 0.93692 0.00029 836.3 (parallel application) 0.88099 0.00030 1005.6 (serial application) 0.85965 0.00030 In summary, there are conservatisms in the soluble boron concentration requirement that provides more than 10% Akeff conservatism.

Burnup Profile The burnup profile used in the Reference I analysis is obtained from Reference 5. As elaborated below, the profile selected for use is very conservative and obviates the need for consideration of plant specific burnup shapes above 30 GWd/MTU. The profile is applied to every depleted fuel assembly modeled in the Reference 1 analysis.

A series of 12 bounding profiles are presented in Reference 5, derived from the DOE burnup shape database (Reference 6). Each profile represents the axial burnup distribution which yields the largest calculated kff in the bumup range considered in that group. The profile for Group 5 covers the burnup range from 30 - 34 GWd/MTU, and per Reference 5 can be conservatively applied to all higher bumups as well. The reason for this is that the burnup profiles become flatter (i.e. higher relative burnup levels near the ends of the fuel) as fuel burnup is increased, thereby reducing the reactivity near the ends of the fuel. Therefore, the profile is conservatively applied in the Reference I analysis for each fuel storage configuration at 30 GWd/MTU and higher. This profile is also applied in the Reference I analysis for each fuel storage configuration from 0 to 30 GWd/MTU; additional information demonstrating that this is conservative for BVPS-2 is provided in RAI Response to Items 2 and 5.

The NUREG guidance (Reference 5) notes that the database presented in Reference 6 is an adequate representation of all spent nuclear fuel from U.S. PWRs. This database includes profiles from Westinghouse plants and fuel as well as other vendor plants and fuel types. The bounding Group 5 profile is 4.5 standard deviations (or approximately 0.03 Akefr) more reactive than the average assembly profile within the Group 5 database. The profile is described as a statistical outlier that causes "a considerable increase in reactivity." This limiting profile is from a B&W 15x] 5 fuel assembly that is likely to have experienced control rod shadowing for a substantial portion of its depletion. The most reactive Group 5 profile presented in Reference 6 for Westinghouse 17xl 7 fuel is only about 0.015 Akeff more reactive than the average assembly profile. This demonstrates that actual Westinghouse 17xl 7 fuel assembly burnup profiles are significantly less skewed than the profile used in the Reference I analysis, reflecting the fact that Westinghouse PWRs such as BVPS-2 have traditionally operated at baseload conditions with (near) all-rods-out, and continue to do so.

, L-08-100 Page 6 of 20 The nodalization used for the Reference 1 analysis is a 4-zone model that has three 6-inch nodes at the top of the assembly to capture the end effect. The remaining 126 inches of the fuel assembly is represented as a single node. Benchmarking calculations show that the 4-zone model is statistically identical to a 7-zone model that has three 6-inch zones at both the top and bottom of the fuel assembly. The reason for this is that the slightly less depleted bottom end of the fuel assembly does not contribute to the overall assembly reactivity. This reactivity is largely driven by the top portion of the assembly where the presence of lower moderating conditions creates a harder neutron spectrum and consequently more plutonium production.

Physical Fuel Rod Tolerances The two physical fuel rod tolerances considered here are on the fuel pellet diameter and the cladding thickness. The tolerance on pellet diameter has been evaluated to determine its reactivity effect, though only the positive tolerance is considered because it adds fissile mass.

The tolerance on cladding thickness has also been evaluated, and is considered in both directions.

The reactivity impact of these tolerances is shown in Table 4, although only the minimum cladding thickness result is provided because it results in a positive reactivity uncertainty. The overall impact on the bias and uncertainty rackup would be 0.00004 Akeif, and is not included in the Reference I analysis. This impact is consistent with Westinghouse experience that has shown the net effect of these physical fuel rod tolerances on the bias and uncertainties to be less than 0.00005 Akeff.

Table 4. Fuel Rod Physical Tolerance Results for the "All-Cell" Configuration Configuration I k I F Akff Nominal 0.97346 0.00011 Maximum diameter fuel pellet 0.97371 0.00011 0.00047 Minimum cladding thickness 0.97434 0.00011 0.00110 One source of conservatism available to offset the small impact of the physical fuel rod tolerances is the assumption on fuel pellet theoretical density. The Reference 1 analysis assumes a 97.5% theoretical density in a right circular cylinder with no fuel pellet dishing or chamfering.

However, if the actual bounding theoretical density of all manufactured BVPS-2 fuel is considered, at least a 1% reduction in theoretical density is realized. A further reduction in fissile mass and reactivity would be realized if the 1.1% nominal pellet dishing and chamfering were explicitly considered.

Table 5 presents analysis to quantify the reactivity decrease associated with a 1% reduction in pellet density only, from 97.5% to 96.5%. The reactivity decrease varies from 0.00081 to 0.00145 Akff depending on fuel storage configuration. If one adjusts the Table 5 keff values for the Monte Carlo uncertainties, then the reactivity decrease varies from 0.00035 to 0.00120 Akff depending on fuel storage configuration. In either case, this reactivity decrease is more than sufficient to compensate for the very small reactivity increase associated with physical fuel rod tolerances.

, L-08-100 Page 7 of 20 Table 5. Conservatism Determined from 1% Reduction in Theoretical Density 97.5% 96.5%

Configuration Theoretical Density Theoretical Density Akff

_______________ keit a kerr _

a__ _

All-Cell 0.97346 0.00011 0.97211 0.00011 0.00135 3x3 0.95866 0.00025 0.95785 0.00021 0.00081 1-out-of-4 5 w/o 15,000 0.96894 0.00016 0.96786 0.00017 0.00107 1-out-of-4 3.85 w/o IFBA 0.95404 0.00013 0.95259 0.00012 0.00145 RAI Item 2 After considering the information in CN-CRIT-244 and NUREG/CR-6801 (Reference 5), it appears the axial burnup profile used in WCAP-16518 under predicts ker for any case run at 15 gigawatt day per metric ton uranium (GWD/MTU) or 25 GWD/MTU. Provide a site-specific analysis that demonstrates that BVPS-2 retains reactivity margin with the burnup profile used in the WCAP-16518 analysis.

RAI Item 5 During the depletion phase of the analysis, core operating parameters should be selected to maximize 24 1Pu production and increase the reactivity of the spent fuel. WCAP-16518 did not use core operating parameters which would maximize 241pu production. NUREG/CR-6665, Reference 4, provides some indication of the impact of the core operating parameters.

The information in CN-CRIT-244 indicated that the axial burnup profile used in WCAP-16518 provides sufficient margin to accommodate this issue above, but not below 30 GWD/MTU. Provide a site-specific analysis that demonstrates that BVPS-2 retains reactivity margin with the burnup profile used in the WCAP-16518 analysis.

Response to Items 2 and 5 An analysis has been performed to investigate the reactivity effects of the use of limiting BVPS-2 burnup profiles, between 10 GWd/MTU and 30 GWd/MTU, and the use of limiting core operating temperatures on the criticality safety conclusions presented in Reference '1. This analysis has considered specific depletion effects from BVPS-2 core operation and the actual geometry and pool conditions in the spent fuel pool. The bumup profiles investigated in this analysis considered:

• all non-blanketed fuel assembly discharge burnup profiles that have less relative burnup in the top two zones than the burnup profile utilized in Reference 1.

• all natural-enriched blanket fuel assembly end-of-cycle burnup profiles from applicable cycles.

• all mid-enriched blanket fuel assembly end-of-cycle burnup profiles from applicable cycles.

, L-08-100 Page 8 of 20 Note that nearly all blanketed fuel assembly burnup profiles have less relative bumup in the top three zones than the burnup profile utilized in the Reference 1 analysis, as expected. The analysis explicitly demonstrates that the reduced reactivity resulting from the lower blanket enrichments is more than sufficient to offset the reactivity effects of the most severe BVPS-2 burnup profiles identified. This axial blanket reactivity benefit conservatively bounds less severe blanketed fuel assembly burnup profiles. -

The limiting BVPS-2 bumup profiles were explicitly analyzed using a limiting temperature profile that bounds uprated core conditions and all previous cycles'. operating conditions.

Reactivity comparisons are made relative to the Reference I conditions.

The following conclusions are drawn from the analysis of these burnup profiles and operating conditions.

• The reactivity of all non-blanketed BVPS-2 fuel assemblies, including the effects of bounding core operating temperatures, is demonstrated to be less than that of the associated minimum bumup requirement contained in Reference 1.

• The reactivity resulting from the most severe blanketed (natural and mid-enriched) fuel assembly bumup profiles, including the effects of bounding core operating temperatures, is demonstrated to be less than that of the Reference I burnup profile and conditions.

The results demonstrate that the Reference I analysis is conservative with respect to the use of limiting BVPS-2 bumup profiles and bounding core operating temperatures.

The following conditions are required to be met, prior to, or concurrent with, amendment implementation, for any new BVPS-2 fresh fuel assemblies in order to remain bounded by the Reference 1 analysis conclusions.

• Fresh fuel assemblies with nominal center-zone enrichments of 3.6 w/o to 4.95 w/o must contain a blanket with a minimum nominal length of 6 inches and with a nominal enrichment that does not exceed 2.6 w/o.

• Fresh fuel assemblies with nominal center-zone enrichments less than 3.6 w/o must contain a blanket with a minimum nominal length of 6 inches and with a nominal enrichment that does not exceed 1.0 w/o.

Deviations from these conditions will require evaluation/analysis to demonstrate that the Reference I conclusions remain applicable.

Description of Analysis The analysis is separated into three categories of depletion calculations: non-blanketed fuel assemblies, natural-enriched blanket assemblies, and mid-enriched blanket assemblies. All BVPS-2 burnup profiles evaluated in the analysis consider a limiting temperature profile that is specified in Table 6. The Reference I bumup profile evaluated in the analysis considers the temperature profile from Reference I (values shown in Table 6).

, L-08-100 Page 9 of 20 Table 6. Moderator Temperature Profiles Axial Zone Reference 1 Limiting Temperature Temperature Profile (f) Profile rF 1 (6 inches) 612.86 619.23 2 (6 inches) 608.99 614.88 3 (6 inches) 605.12 610.53 4 (126 inches) 574.14 577.36 Non-Blanketed Fuel Assemblies Non-blanketed fresh fuel was only inserted into BVPS-2 Cycle 1; the last of these assemblies were discharged at end of Cycle 5, and there is no plan to use these fuel assemblies again in BVPS-2. The analysis considers the discharge burnup profiles from all non-blanketed fuel assemblies used in the BVPS-2 core. Most of these discharge burnup profiles were found to be bounded by the burnup profile used in Reference 1; however, five of these discharge bumup profiles were found to be not bounded by the profile used in Reference 1, in that these profiles had less relative burnup in the top two zones of the assembly than the Reference I burnup profile. Since the axial burnup profile reactivity effect is dominated by the relative bumups in the top two zones, these particular profiles will be limiting. These five limiting non-blanketed assembly burnup profiles are shown in Table 7.

Table 7. Bounding Non-Blanketed Assembly Burnup Profiles Axial Zone Reference I Profile V) Profile 2(2) Profile 3(2) Profile 4(2) Profile 5(2)

Profile Cycle 3 Cycle I Cycle I Cycle I Cycle I 1 0.462 0.443 0.471 0.496 0.511 0.487 2 0.738 0.698 0.703 0.724 0.737' 0.719-3 0.971 0.860 0.836 0.847 0.855 0.845 4 1.039 1.048 1.047 1.044 1.043 1.045 TProfile used for Table 9 analysis.

(2) Profile used for Table 10 analysis.

For each of the five fuel assemblies with limiting burnup profiles, the following steps were performed.

A depletion calculation was performed using that assembly's initial enrichment, and using the limiting operating temperatures defined in Table 6.

The isotopics resulting from this depletion calculation were then used in a spent fuel pool storage configuration calculation that also used the actual fuel assembly burnup and that assembly's limiting bumup profile.

The results of this calculation were then compared against a similar calculatikrn that used the actual fuel assembly enrichment, and fuel assembly isotopics obtained from a depletion using the Reference I operating temperature profile, the Reference I burnup profile, and the minimum

, L-08-100 Page 10 of 20 required bumup for the storage configuration in question. The minimum required burnup for each storage configuration was determined from the bumup versus enrichment polynomials in Reference 1.

If the reactivity from the first calculation was less than that from the second calculation, then margin existed within the Reference I analysis to permit storage of that limiting burnup shape.

This set of calculations was repeated for each storage configuration that was permissible for the combination of enrichment and burnup represented by the actual fuel assembly.

Table 9 shows the results of these calculations for the single assembly (at 3.099 w/o 2 35U) discharged from Cycle 3 that is lower than the Reference I burnup profile in zones I and 2.

These results demonstrate that this actual profile is less reactive than that based upon the Reference I profile in all spent fuel pool storage configurations in which this fuel assembly is allowed to be stored in accordance with the Reference I requirements. This supports the conclusion that while a single Cycle 3 bumup shape has less burnup in the top two zones than the Reference I burnup profile, the reactivity determined in the Reference 1 analysis is conservative.

Note that the "RSS cY"column is the root-sum-square of the Monte Carlo standard deviations from the neutronic simulations - this is the uncertainty on each reported Ak value in the table.

Table 10 shows the results for the four assemblies (at 2.105 w/o 235U) discharged from Cycle I that are lower than the Reference 1 burnup profile in zone 2. These results demonstrate that these actual fuel assemblies are less reactive than the limit from Reference I in all spent fuel pool storage configurations. Note that the assembly with burnup Profile 4 and 18,427 MWd/MTU of burnup is bounded by a small reactivity difference in the 3x3 storage configuration that may be statistically insignificant. This assembly was discharged from the core at the end of BVPS-2 Cycle I operation. When the 24 1 Pu decay (and associated 241Am buildup) is credited over a 15 year period (this assembly has more than 15 years of decay time), the reactivity of the assembly is demonstrated to be 0.00924 Ak conservative relative to the Reference 1 limit. No credit is taken for the decay of any other actinides or fission products.

This supports the conclusion that while four Cycle I burnup shapes have been identified that have less burnup in zone 2 than the Reference I burnup profile, the reactivity determined in the Reference 1 analysis is conservative in all cases.

Blanketed Assemblies The blanketed assembly burnup profiles selected for investigation in this analysis are shown in Table 8. These are the burnup profiles that have the lowest relative burnup in the top two zones of the assembly, and therefore result in the highest reactivity. The reactivity behavior is investigated at 15000 MWd/MTU, 20000 MWd/MTU and 25000 MWd/MTU in each instance.

Natural Blanket Assemblies Fresh fuel containing natural blankets was only inserted into BVPS-2 Cycles 2 through Cycle 8.

Table II shows the spent fuel pool ken' values for assembly burnup profiles with natural blankets.

The bounding end-of-cycle bumup profile for all natural blanket assemblies ever used at BVPS-2 was selected. The reactivity of the bounding assembly is compared to that of a depleted assembly with the profile utilized in Reference I (without blankets) at the same absolute burnup in the All-Cell storage configuration. This configuration produces the largest coupling of blanketed fuel burnup profiles and will conservatively represent the effect relative to the other

, L-08-100 Page 11 of 20 storage configurations. The isotopics for the natural blanket assembly were calculated using the limiting operating temperature profile shown in Table 6.

These comparisons are performed for a center-zone 235U enrichment range of 3.2 w/o to 5.0 w/o.

Note that the natural blankets were simulated with 1.0 w/o fuel to conservatively represent their reactivity-dampening effect.

The results in Table 11 demonstrate that this natural blanket burnup profile is less reactive than the Reference I profile in all spent fuel pool storage configurations. This supports the conclusion that while natural blanket burnup shapes have less burnup in the top zones than the Reference 1 burnup profile, the reactivity determined in the Reference I analysis is conservative in all cases.

Mid-Enriched Blanket Assemblies Since Cycle 9, all fresh fuel inserted into BVPS-2 has contained mid-enriched (2.6 w/o) blankets.

Table 12 and Table 13 show the spent fuel pool kefr values for assembly bumup profiles with mid-enriched blankets. The bounding profiles which were used are provided in Table 8. The reactivity of an assembly with each of the bounding profiles is compared to that of a depleted assembly with the profile utilized in Reference I (without blankets) at the same absolute burnup in the All-Cell storage configuration. This configuration produces the largest coupling of blanketed fuel burnup profiles and will conservatively represent the effect relative to the other storage configurations. The isotopics for the mid-enriched assemblies were calculated using the limiting operating temperature profile shown in Table 6.

These comparisons are performed over two 235U enrichment ranges. The bumup profiles occurring in assemblies with lower center-zone enrichment are grouped and analyzed across a range from 3.6 w/o to 4.6 w/o. Burnup profiles occurring in assemblies with higher center-zone enrichment are grouped and analyzed across a range from 4.6 w/o to 5.0 w/o. The enrichment ranges are selected because higher center-zone enrichment assemblies produce more limiting bumup profiles for a given blanket enrichment, and it would be overly conservative to apply these profiles to fuel assemblies with lower center-zone enrichments but containing the same blanket enrichment.

Note that the mid-enriched blankets were all modeled with 2.60 w/o solid fuel pellets. This is the only mid-enriched blanket enrichment that has been utilized at BVPS-2. Also, use of solid fuel pellets conservatively represents the reactivity behavior of annular fuel pellets that are sometimes utilized in blanket fuel at BVPS-2.

The results in Table 12 and Table 13 demonstrate that these mid-enriched blanket burnup profiles are less reactive than the Reference I profile in all spent fuel pool storage configurations.

In most cases, a trend of increasing margin to the Reference I results with increasing burnup was noted. In two cases, however, this trend was not seen. The reactivity differences which establish these trends are noted to be relatively small, and the absence of such a trend is statistically insignificant, given that the one-sigma uncertainties on these reactivity differences are of similar magnitude to the trends themselves. This supports the conclusion that while mid-enriched blanket bumup shapes have less burnup in the top zones than the Reference I burnup profile, the reactivity determined in the Reference I analysis is conservative in all cases.

, L-08-100 Page 12 of 20 Table 8. Bounding Blanketed Assembly Burnup Profiles Natural Blankets (Cycles 2 - 8) Mid-Enriched Blankets (Cycles 9 - 13) 3.20 - 5.00 w/o 3.60 - 4.60 w/o 4.60 - 5.00 w/o Axial Limiting Limiting Limiting Limiting Limiting Zone Zone I & 2 Zone I Zone 2 Zone I Zone 2 Profileo') Profile(2) Profile(2) Profile 3) Profile(3)

Cycle 7 Cycle 9 Cycle 9 Cycle 9 Cycle 13 1 0.151 0.334 0.344 0.331 0.341 2 0.638 0.741 0.698 0.703 0.696 3 0.864 0.836 0.908 0.911 0.871 4 1.064 1.052 1.050 1.050 1.052 (1)

Profile used for Table 11 analysis.

(2)

Profile used for Table 12 analysis.

(3)

Profile used for Table 13 analysis.

, L-08-100 Page 13 of 20 Table 9. kff Comparison of Non-Blanketed Cycle 3 Discharge Shape and Reference 1 Burnup Limit Description Enrichment Burnup kegr +/- 0 Ak Ref. 1 profile, Ref. I temps, non-blanketed, All-Cell 14131 0.97635 +/- 0.00032 Ref. I profile, Ref. I temps, non-blanketed, 1-out-of-4 with 15k 3.099 21377 0.97274 +/- 0.00032 )

Ref. I profile, Ref. I temps, non-blanketed, 3x3, 20 years decay 27003 0.97550 +/- 0.00051 RSS a

&1eý.

3,Pof le4 '

All-Cell 0.89555 +/- 0.00032 -0.08080 +/- 0.00045 1-out-of-4 with 15k 0.95431 +/- 0.00036 -0.01843 +/- 0.00048 3x3, 20 years decay 3.099 27491 0.97424 +/- 0.00047 -0.00126 +/- 0.00069 Assembly with this enrichment & burnup I -out-of-4 with 3.85 w/o T 7 in.otpermituted to be stored in tIs coniflon

< Burnup profile described in Table 7. Used limiting temperature profile described in Table 6.

, L-08-100 Page 14 of 20 Table 10. kef Comparison of Non-Blanketed Cycle 1 Discharge Shapes and Reference 1 Burnup Limit Description Enrichment Burnup keff +/- Ak Ref. I profile, Ref. f temps, non-blanketed, All-Ceil 2940 0.97140 +/- 0.00032 Ref. I profile, Ref. I temps, non-blanketed, l-out-of-4 with 15k 8235 0.97357 +/- 0.00032 Ref. I profile, Ref. I temps-, non-blanketed, I-out-of-4 with 3.85 w/o 15154 0.98219 +/- 0.00039 Ref. 1 profile, Ref. I temps, non-blanketed, 3x3, 0 years decay 2.105 17967 0.97820 +/- 0.00050 Ref. 1 profile, Ref. 1 temps, non-blanketed, 3x3, 5 years decay 16780 0.97747 +/- 0.00050 Ref. I profile, Ref. I teps, non-blanketed, 3x3, 50 years decay 16804 0.97870 - 0.0004 Ref. I profile, Ref. 1 temps, non-blanketed, 3x3, 10 years decay 15044 0.97850 +/- 0.00043 Ref. I profile, Ref. I temps, non-blanketed, 3x3, 15 years decay 14695 0.97866 0.00051 Ref. I rofile, Ref I terns, non-blanketed, 3x, 20 years decay 14332 0.97778 +/- 0.00051 4 RSS a All-Cell 0.86628 +-0.00031 -0.10512 +- 0.00045 1-out-of-4 with 15k 2.105 15874 0.94181 +/- 0.00035 -0.03882 +/- 0.00047 l-out-of-4 with 3.85 w/o 0.97992 +/-_0.00037 -0.00227 +/-_ 0.00054 3x3, 10 years decay 0.97670 0.00054 -0.00180 0.00069 All-Cell 0.85341 +/-_0.00028 -0.11799 +_-_ 0.00043 l-out-of-4 with 15k 2.105 17853 0.93475 +/- 0.00036 -0.03882 +/- 0.00048 Cctc1,Pfile4! 3 >

l-out-of-4 with 3.85.w/o 0.97323 +/- 0.00040 -0.00896 +/- 0.00056 3x3, 5 years decay 0.97544 +/- 0.00050 -0.00203 +/- 0.00071 All-Cell 0.84840 +-0.00028 -0.12300 +/.0.00043 1-out-of-4 with. 15k 0.93225 +-0.00034 -0.04132 +- 0.00047 l-out-of-4 with 3.85 w/o 2.105 18427 0.97252 +-0,00037 -0.00967 +- 0.00054 3x3, 0 years decay 0.97805 +-0.00049 -0.00015 +- 0,00070 3x3, 15 years decay 0.96896 +/- 0.00053 -0.00924 +/- 0.00074 c~Irf1, t 1 All-Cell 0.87381 +/- 0.00029 -0.09759 +/- 0.00043 I1-out-of-4 with 15k 0.94548 +/- 0.00037 -0.02809 +/- 0.00049 2.165 14686 30x Assembly with this enrichment & burnup l-out-of-4 with 3.85 w/o not permittedto be stored in these configurations

(') Burnup profile described in Table 7. Used limiting temperature profile described in Table 6.

, L-08-100 Page 15 of 20 Table 11. keff Comparison of Bounding Natural Blanket Burnup Profile and Reference I Burnup Profile Description Enrichment Burnup kr +/-- a Ak 15000 0.97812 +/- 0.00034 --

3.2 20000 0.94221 +/- 0.00032 25000 0.91075 ,/- 0.00034 15000 1.07106 +/- 0.00030 Ref. 1 profile, Ref. I tenips, non-blanketed 4.6 20000 1.03841 +/- 0.00033 -

25000 1.00682 +/- 0.00032 Z 15000 1.09134 +/- 0.00031 5.0 20000 1.05857 +/- 0.00034 -*

25000 1.02857 +/- 0.00029 - RSS a 15000 0.97480 +/- 0.00032 -0.00332 +/- 0.00047 3.2 20000 0.93679 +/- 0.00030 -0.00542 +/- 0.00044 25000 0.90285 +/- 0.00031 -0.00790 +/- 0.00046 15000 1.06796 +/- 0.00031 -0.00310 +/- 0.00043 Table 8 Natural Blanket Limiting Profile

• 4.6 20000 1.03345 +/- 0.00032 -0.00496 +/- 0.00046 25000 1.00068 +/- 0.00031 -0.00614 +/- 0.00045 15000 1.08847 +/- 0.00032 -0.00287 +/- 0.00045 5.0 20000 1.05478 +/- 0.00031 -0.00379 +/- 0.00046 25000 1.02343 +/- 0.00031 -0.00514 +/- 0.00042 235 I)lsotopics determined using limiting temperature profile described in Table 6. Natural blanket conservatively represented using 1 w/o U fuel.

, L-08-100 Page 16 of 20 Table 12. kff Comparison of Bounding Mid-Enriched Blanket Burnup Profiles and Reference I Burnup Profile at Low Enrichments (3.6- 4.6 w/o)

Description Enrichment Burnup kr + a a- Ak 15000 1.00853 +/- 0.00032 3.6 20000 0.97339 +/- 0.00030

_________ 25000 0.94150 +-0.00032 o Ref. 1 profile, Ref. 1temps, non-blanketed 15000 107150 0.00030 4.6 20000 1.03841 +/- 0.00033 25000 1.00682 +/- 0.00032 RSS 15000 1.00786 +/- 0.00031 -0.00067 +/- 0.00045 3.6 20000 0.97175 +/- 0.00031 -0.00164(2) +/- 0.00043 Table 8 Mid-Enriched Blanket, 3.6-4.6 w/o, Limiting Zone 1 Profile (') 25000 0.94064 +/- 0.00032 -0.00086 +/- 0.00045 15000 1.06995 +/- 0.00033 -0.00111 +/- 0.00045 4.6 20000 1.03582 +/- 0.00033 -0.00259 +/- 0.00047 25000 1.00301 +/- 0.00031 -0.00381 +/- 0.00045 15000 1.00794 +/- 0.00031 -0.00059 +/- 0.00045 3.6 20000 0.97206 +/- 0.00030 -0.00133 +/- 0.00042 Table 8 Mid-Enriched Blanket, 3.6-4.6 w/o, Limiting Zone 2 Profile (_) 25000 0.93915 +/- 0.00032 -0.00235 +/- 0.00045 15000 1.06953 +/- 0.00030 -0.00153 +/- 0.00042.

4.6 20000 1.03545 +/- 0.00031 -0.00296 +/- 0.00045 25000 1.00314 +/- 0.00030 -0.00368 +/- 0.00044

• Isotopics determined using limiting temperature profile described in Table 6.

(2) Note that the Ak result does not follow the general trend of increasing conservatism with increasing burnup due to the statistical variation of the calculations.

, L-08-100 Page 17 of 20 Table 13. knff Comparison of Bounding Mid-Enriched Blanket Burnup Profiles and Reference 1 Burnup Profile at High Enrichments (4.6 - 5.0 w/o)

Description Enrichment Burnup keff +/" . Ak 15000 1.07106 +/- 0.00030 4.6 20000 1.03841 +/- 0,00033 Ref. 1 profile, Ref. I temps, non-blanketed 25000 1.00682 +/- 0.00032 .

15000 1.09134 +/- 0.00031 '

5.0 20000 1.05857 +/- 0.00034 25000 1.02857 +/- 0.00029 15000 1.06948 +/- 0.00032 -0.00158 +/- 0.00044 4.6 20000 1.03560 +/- 0.00032 -0.00281 +/- 0.00046 Table 8 Mid-Enriched Blanket, 4.6-5.0 w/o, Limiting Zone 1 Profile o) 25000 1.00344 +/- 0.00031 -0.00338 +/- 0.00045 15000 1.08954 +/- 0.00035 -0.00180 +/- 0.00047 5.0 20000 1.05640 +/- 0.00031 -0.00217 +/- 0.00046 25000 1.02566 +/- 0.00030 -0.00291 +/- 0.00042 M i-n i B (16 NýQ))

15000 1.06961 +/- 0.00033 -0.00145 +/- 0.00045 4.6 20000 1.03611 +/- 0.00033 -0.00230 +/- 0.00047 Table 8 Mid-Enriched Blanket, 4.6-5.0 w/o, Limiting Zone 2 Profile) 25000 1.00369 +- 0.00031 -0.00313 +- 0.00045 15000 1.08945 +/- 0.00034 -0.00189 +/- 0.00046 5.0 20000 1.05705 +/- 0.00034 -0.00152 (2) +/- 0.00048 25000 1.02540 +/- 0.00032 -0.00317 +/- 0.00043

( Isotopics determined using limiting temperature profile described in Table 6.

(2) Note that the Ak result does not follow the general trend of increasing conservatism with increasing bumup due to the statistical variation of the calculations.

, L-08-100 Page 18 of 20 RAI Item 3 WCAP-16518 calculates the burnup uncertainty in a method different than that specified in the NRC staff guidance, Reference 3. TheNRC staff has reviewed the justification for the new method and determined it is inadequate to justify the method. Calculate the burnup uncertainty in accordance with the NRC staff guidance in Reference 3.

Response to Item 3 The 5% Reactivity Decrement method suggested in the NRC guidance document (Reference 4) has been used to determine a revised burnup uncertainty. In this method, fresh fuel assemblies (5 w/o, 0 GWd/MTU burnup) are substituted into each "Depleted Fuel" location in the storage configurations presented in Section 3.1 of Reference 1. The total reactivity credit associated with fuel burnup is determined by comparing the burnup-credited k~f of the configuration (i.e.,

0.995 minus the Sum of Biases and Uncertainties) to the keff of the same configuration with the Depleted Fuel locations replaced with the most reactive possible fuel assemblies (5 w/o, 0 GWd/MTU bumup). This reactivity difference is multiplied by 5% to determine the decremental burnup uncertainty.

As summarized in Table 14, using the revised burnup uncertainty, the overall Sum of Biases and Uncertainties is increased by 0.00222 to 0.00303 Akeff, depending on storage configuration, compared to the methodology presented in Reference 1.

Table 14. Sum of Biases and Uncertainties Confiauration (units in Ak~f.)

1-out-of-4 1-out-of-4 Parameter All Cell 3x3 3.85 w/o Fresh 5.0 w/o at with IFBA 15 GWD/MTU Increased Enrichment 0.00284 0.00173 0.00338 0.00226 Decreased Cell Pitch 0.00674 0.00472 0.00538 0.00678 Decreased Rack Thickness 0.00609 0.00645 0.00411 0.00505 Increased Rack ID 0.00033 0.00090 0.00111 0.00037 Off-Center Positioning 0.00740 0.01602 0.00670 0.00760 Wrapper Thickness 0.00326 0.00362 0.00294 0.00278 Burnup Uncertainty 0.01246 0.01217 0.01112 0.01097 Methodology Uncertainty 0.00643 0.00645 0.00643 0.00643 Statistical Sum of Uncertainties 0.01878 0.02296 0.01664 0.01743 Methodology Bias 0.00310 0.00310 0.00310 0.00310 Temperature Bias 0.01077 0.00120 0.00534 0.00983 5% Reactivity Decrement Method - Sum 0.03265 0.02726 0.02508 0.03036 of Biases and Uncertainties Reference I Method- "

Sumefe BIesnd-'

c U0.03043 0.02423 0.02217 Sum of Biases and Uncertainties 0.02758 Akeff Between 5% Reactivity Decrement 0.00222 0.00303 0.00291 0.00278 Method and Reference_ Method

/

, L-08-100 Page 19 of 20 RAI Item 4 WCAP-16518 calculates a typical temperature bias for each storage configuration. The staff has reviewed the justification for method calculating the temperature bias and determined it is inadequate for the staff to reach a reasonable assurance conclusion that the limiting temperature bias has been determined for each storage configuration. Provide a site-specific analysis that demonstrates that the limiting temperature bias has been determined for each storage configuration.

Response to Item 4 Additional temperature bias calculations have been performed for each of the four storage configurations presented in Reference 1. For each storage configuration, a temperature bias has been calculated as a function of fuel enrichment (3, 4, and 5 w/o), fuel burnup (at each enrichment level, two burnup levels are selected that bracket above, and below the required fuel burnup versus enrichment curves presented in Reference 1), and temperature (50'F and 185°F).

Therefore, the reactivity effects of spent fuel pool temperature have been calculated for a total of 48 unique combinations of storage configuration (4), fuel enrichment (3), fuel burnup (2), and temperature (2).

It is observed that; for each unique storage configuration and fuel enrichment, there is a consistent trend wherein the temperature bias decreases with increasing fuel burnup. Therefore, for each configuration, the temperature biases calculated at the two burnup levels (a span of only 10 GWD/MTU) are interpolated to obtain the temperature bias at the minimum required fuel burnup to satisfy the fuel storage requirement for that configuration as presented in Reference 1.

As summarized in Table 15, the temperature bias has no impact on the reactivity of two of the storage configurations (All Cell and I -out-of-4 3.85 w/o Fresh with IFBA), compared to the values presented in Reference 1. For the other two configurations (3x3 and I-out-of-4 5.0 w/o at 15 GWD/MTU), the temperature bias increases the reactivity of these storage configurations by 0.00025 Akef and 0.00090 Akef, respectively, compared to the values presented in Reference 1.

Table 15. Sum of Temperature Biases Configuration (units in Akeff) 1-out-of-4 I-out-of-4 Parameter All Cell 3x3 3.85 w/o Fresh 5.0 w/o at with IFBA 15 GWD/MTU Increase in Temperature Bias _<O.00000 0.00025 0.00090 *0.00000 The conclusion for Items 3 and 4 is that the combination of the increased temperature bias and the increase in bumup uncertainty by applying the 5% reactivity decrement method is less than the 0.00500 Akiff administrative margin included in the Reference I analysis.

, L-08-100 Page 20 of 20 RAI References

1. FirstEnergy Nuclear Operating Company letter L-07-084, Peter P. Sena III, Site Vice President, Beaver Valley Power Station, to USNRC document control desk re: "Beaver Valley Power Station, Unit Nos. 1 and 2, BV-1 Docket No. 50-334, License No. DPR-66, BV-2 Docket No. 50-412, License No. NPF-73, Responses to a Request for Additional Information (RAI dated May 21, 2007) in Support of License Amendment Request Nos.

333 and 204 (TAC Nos. MD2377 and MD2378)," July 20, 2007. (ADAMS ML072050213)

2. FirstEnergy Nuclear Operating Company letter L-07-103, Edward H. Hubley, Acting Director, Maintenance, Beaver Valley Power Station, to USNRC document control desk, re: "Beaver Valley Power Station, Unit Nos. 1 and 2, BV-I Docket No. 50-334, License No. DPR-66, BV-2 Docket No. 50-412, License No. NPF-73, Supplemental Information for License Amendments Request Nos. 333 and 204 (Revision 2 of WCAP-16518) (TAC Nos. MD2377 and MD2378)" July 26, 2007. (ADAMS ML073320036)

3. NRC Memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998. (ADAMS ML003728001)

4. NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel." (ADAMS ML003688150)

5. NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis." (ADAMS ML03110292)

Response References

1. WCAP-1 6518-P, Revision 2, "Beaver Valley Unit 2 Spent Fuel Pool Criticality Analysis,"

V.N. Kucukboyaci, July 2007.

2. Code of Federal Regulations, Title X, Part 50.68, "Criticality Accident Requirements,"

November 16, 2006.

3. ANSI/ANS-8.17-2004, "Criticality Safety Criteria for Handling, Storage, and Transportation of LWR Fuel Outside Reactors," November 3, 2004.

4. NRC Memorandum from L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998.

5. NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses," J.C. Wagner, et. al., March 2003.

6. YAEC-1937, Yankee Atomic Electric Company, "Axial Bumup Profile Database for Pressurized Water Reactors," R. J. Cacciapouti and S. Van Volkinburg, May 1997.

ATTACHMENT 2 L-08-100 Regulatory Commitment List Page 1 of 1 The following list identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for Beaver Valley Power Station (BVPS) Unit No. 2 in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC. They are described only as information and are not Regulatory Commitments. Please notify Mr.

Thomas A. Lentz, Manager - Licensing, at (330) 761-6071 of any questions regarding this document or associated Regulatory Commitments.

Regulatory Commitments Due Date The following conditions are required to be met Prior to, or concurrent with, for any new BVPS-2 fresh fuel assemblies in amendment implementation.

order to remain bounded by the Reference I analysis conclusions.

• Fresh fuel assemblies with nominal center-zone enrichments of 3.6 w/o to 4.95 w/o must contain a blanket with a minimum nominal length of 6 inches and with a nominal enrichment that does not exceed 2.6 w/o.

• Fresh fuel assemblies with nominal center-zone enrichments less than 3.6 w/o must contain a blanket with a minimum nominal length of 6 inches and with a nominal enrichment that does not exceed 1.0 w/o.