Text

License Amendment Request for Spent Fuel Pool Criticality Technical Specification Changes
	ML15327A228
Person / Time
Site:	Prairie Island
Issue date:	11/17/2015
From:	Davison K; Northern States Power Co, Xcel Energy
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15327A244	List: L-PI-15-087, Enclosures 7 & 8: Modeling of Fuel Assemblies Containing Both Ifba and Gadolinia Absorber Rods with Westinghouse Core Design Code Systems and Affidavit; ML15327A228; ML15327A229;
References
	L-PI-15-087
	Download: ML15327A228 (34)
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ENCLOSURES 4 AND 6 CONTAIN PROPRIETARY INFORMATION

-WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390~Prairie Island Nuclear Generating Plant Xcel ne_.*rav1717 Wakonade Drive East Xcel~ ergyWelch, MN 55089 L-P I-1 5-087 November 17, 2015 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 License Amendment Request for Spent Fuel Pool Criticality Technical Specification Changes Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby requests an amendment to the renewed operating licenses for Prairie Island Nuclear Generating Plant (PINGP).Specifically, NSPM proposes to revise Technical Specification (TS) 3.7.16, "Spent Fuel Storage Pool Boron Concentration," and TS 4.3.1, "Fuel Storage Criticality," to allow spent fuel pool (SFP) storage of nuclear fuel containing a boron-based neutron absorber in the form of zirconium diboride (ZrB 2) Integral Fuel Burnable Absorber (IF BA).Enclosure 1 to this letter provides the evaluation of the proposed TS changes and their supporting justifications, including a no significant hazards determination.

Enclosure 2 provides the current TS pages marked-up to show the proposed changes. Enclosure 3 provides, for information only, the current TS Bases pages marked-up to show the associated proposed Bases changes. Final TS Bases changes will be implemented pursuant to TS 5.5.12, "Technical Specifications (TS) Bases Control Program,".

at the time the amendment is implemented.

Enclosure 4 provides Westinghouse Electric Company, LLC (WE£C) report WCAP-.17400-P, Supplement 1, Revision 1, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Analysis -Supplemental Analysis for the Storage of IFBA Bearing Fuel," dated October 2015. This report provides the analytical basis for the revised TS. This report contains proprietary information.

Enclosure 5 provides the non-proprietary version of the Westinghouse Report,.WCAP-17400-NP, Supplement 1, Revision 1.Enclosure 6 provides a WEC document to explain how the primary neutronic codes used in the supporting spent fuel criticality analysis remain valid for modeling fuel Document Control Desk Page 2 assemblies containing both IFBA and gadolinia absorber rods at PINGP. The document is entitled, "Modeling of Fuel Assemblies Containing both IFBA and Gadolinia Absorber Rods with Westinghouse Core Design Code Systems." Enclosure 6 contains proprietary information.

Enclosure 7 provides the non-proprietary version of the document.Enclosure 8 contains the Westinghouse Applications for Withholding Proprietary Information from Public Disclosure, accompanying Affidavits, Proprietary Information Notices, and Copyright Notices. These WEC affidavits set forth the basis on which the information may be withheld from public disclosure by the NRC and .addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).

NSPM requests that the proprietary information in Enclosures 4 and 6 be withheld from public disclosure in accordance with 10 CFR 2.390(a)4, as authorized by 10 CFR 9.17(a)4.

Accordingly, it is respectfully requested that the information which is proprietary to WEC be withheld from public disclosure in accordance with 10 CFR 2.390.Correspondence with respect to the copyright or proprietary aspects of the items provided in Enclosures 4 and 6 of this letter or the supporting Westinghouse affidavit should reference the respective WEC letter number (CAW-1 5-4311 or CAW-1 5-4308)" and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.NSPM has determined that the proposed amendment does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment meets the categorical exclusion requirements of 10 CFR 51 .22(c)(9) and an environmental impact assessment need not be prepared.A copy of this submittal, including the Determination of No Significant Hazards Consideration, without Enclosures 2 through 8, is being forwarded to the designated State of Minnesota official pursuant to 10 CFR 50.91 (b)(1).NSPM requests approval of this proposed amendment by November 30, 2017. Once approved, .the amendment will be implemented within 120 days.If there are any questions or if additional information is needed, please contact Glenn Adams at 612-330-6777.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

Document Control Desk Page 3 I declare under penalty of perjury that the foregoing is true and correct.Executed on November 17, 2015 Kevin Davison Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company -Minnesota Enclosures (8)cc: Regional Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC State of Minnesota (without enclosures 2 through 8)

L-PI-1 5-087 NSPM Enclosure 1 ENCLOSURE 1 Evaluation of the Proposed Change License Amendment Request for Spent Fuel Pool Criticality Technical Specification Chanqes 1.0

SUMMARY

DESCRIPTION

2.0 DETAILED

DESCRIPTION

2.1 Proposed

Change to TS 3.7.16, "Spent Fuel Storage Pool Boron Concentration" 2.2 Proposed Change to TS 4.3.1, "Fuel Storage Criticality" 2.3 Proposed Change to Criticality Analysis Methodology

2.4 Other

Proposed Changes to the Current Licensing Basis 3.0 TECHNICAL EVALUATION

3.1 Design

Description

3.2 Current

Licensing Basis 3.3 Justification for the Proposed Changes 3.4 Conclusion

4.0 REGULATORY EVALUATION

4.1 Applicable

Regulatory Requirements/Criteria

4.2 Precedent

4.3 Significant Hazards Consideration

4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATIONS

6.0 REFERENCES

ATTACHMENT 1 Page 1 ofi15 L-PI-1 5-087 NSPM Enclosure 1 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby requests an amendment to the renewed operating licenses for Prairie Island Nuclear Generating Plant (PINGP). Specifically, NSPM proposes to revise Technical Specification (TS) 3.7.16, "Spent Fuel Storage Pool Boron Concentration" and TS 4.3.1, "Fuel Storage Criticality" to allow spent fuel pool (SFP) storage of nuclear fuel containing a boron-based neutron absorber in the form of zirconium diboride (ZrB 2) Integral Fuel Burnable Absorber (IFBA).2.0 DETAILED DESCRIPTION The proposed changes to the TS and current licensing basis are as follows: 2.1 Proposed Change to TS 3.7.16, "Spent Fuel Storage Pool Boron Concentration" The proposed change will increase the value of minimum concentration of soluble boron required in the spent fuel pool from 1800 parts per million (ppm) to 2500 ppm. This increase would provide sufficient niegative reactivity to maintain the required subcriticality margin for a more conservative misloading accident than previously analyzed.2.2 Proposed Change to TS 4.3.1, "Fuel Storage Criticality" The proposed change to TS Table 4.3.1-3 involves a complete set of new coefficients for calculating the minimum required fuel assembly burnup as a function of decay time and enrichment, specifically for fuel not operated in PINGP operating Cycles 1 through 4. The revised coefficients result in burnup values that are up to 4 GWD/MTU higher than existing requirements.

2.3 Proposed

Change to Criticality Analysis Methodology The proposed change involves an explicit change to the criticality analysis methodology.

As described in Enclosure 4 (Section S4.1.2.1 .4), the methodology has been revised to capture regulatory guidance (NUREG/CR-71

09) and adopt a certain bias for minor actinide and fission product nuclides.

Herein, NSPM requests approval of this methodology change.As described in Enclosure 6, the Westinghouse Electric Company, LLC (WEC)neutronic codes used to determine axial power shapes and burnup profiles for the spent fuel criticality analysis remain valid for the combination of boron and gadolinia.

The code suite used to calculate the spent fuel criticality depletion models only IFBA in the fuel, as further discussed in Section 2.4. With respect to this combination of neutron absorbers, the proposed amendment does not involve Page 2 of 15 L-PI-1 5-087 NSPM Enclosure 1 any change to the computer codes that comprise the evaluation methodology currently described in the Updated Safety Analysis Report (USAR).2.4 Other Proposed Changes to the Current Licensing Basis In addition to the specific changes to TS and analysis methodology discussed above, two conservative changes are introduced to the licensing basis as inputs to the models used in the spent fuel criticality analysis (SFCA). These changes were discussed with NRC Staff at a pre-application meeting (Reference 6.6): 1. Mode lin~q the effects of the neutron absorber.

The current licensing basis spent fuel criticality analysis conservatively ignores the effect of the current neutron absorber (gadolinia) because it is a net poison throughout the operating cycle.However, this effect is not valid for the new proposed neutron absorber which is boron, in the form of zirconium diboride IFBA. Therefore, the licensee has conservatively included the IFBA neutron absorber in the depletion models (as it hardens the neutron spectrum to increase reactivity), and conservatively ignored the negative reactivity effect of residual IFBA in the SFP criticality analysis.2. Multiple-Assembly Misloadinq Accidents.

The proposed amendment also involves the analysis of a new accident that extends beyond the Double Contingency Principle (the regulatory basis for nuclear fuel storage criticality analyses that states two unlikely independent and concurrent incidents or postulated accidents are beyond the scope and need not be analyzed).

Whereas the current licensing basis limits the misloading accident to just a single fuel assembly, the proposed amendment would conservatively adopt a multiple-misloading event in lieu of attempting to justify the low probability of such an event. In effect, the proposed criticality analysis (provided in Enclosure

4) analyzes a conservative array of fuel that bounds any possible combination of misloading events.3.0 TECHNICAL EVALUATION

3.1 Design

Description Prairie Island Units 1 and 2 share a common spent fuel pool that employs one modular storage rack design throughout.

As described in PINGP USAR Section 10.2.1, the storage rack design originally credited Boraflex neutron absorber panels between the storage cells to help meet subcriticality criteria.

These Boraflex panels are degraded and have not been credited in the current design basis. The rack design does benefit from a dedicated "flux-trap" design that provides a minimum gap between cells. Key design parameters for the storage racks are provided in USAR Section 10.2.1 and Reference 6.1.To ensure stored fuel remains in a subcritical configuration during any normal or Page 3 of 15 L-PI-1 5-087 NSPM Enclosure 1 accident condition, strict administrative controls require that any fresh (new) fuel assembly or spent fuel assembly loaded into a storage rack is first evaluated to ensure it meets the loading restrictions of TS 3.7.17 and 4.3.1. Currently, each fuel assembly is qualified for a storage location based on several key parameters that include initial enrichment, burnup, and decay time. Certain parameters (e.g., initial enrichment) are determined from fuel records. Other parameters (e.g., burnup and decay time) are determined from core operating records. The value of burnup is the average assembly exposure in megawatt days per metric ton uranium (MWD/MTU) and is currently calculated using an industry standard core power distribution system called BEACONTM (Best Estimate Analyzer for Core Operations

-Nuclear);

however, other suitable methods have been used previously.

, Once an assembly is selected for placement based on the required characteristics, procedures ensure that the fuel assembly is qualified for its new location, and that it is safely placed in the designated location.The spent fuel storage racks are designed so that it is impossible to insert assemblies between rack modules or between rack modules and the spent fuel pool wall. Besides the procedural controls on fuel selection and placement in accordance with allowable storage arrays, criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack that limits fuel assembly interaction.

This is done by fixing the minimum separation between assemblies and/or maintaining soluble neutron poison (i.e., boron) in the spent fuel pool water.The required subcriticality margin of safety for the stored fuel is assured with the soluble boron present in the spent fuel pool. TS 3.7.16 presently requires a minimum soluble boron concentration of 1800 ppm whenever fuel is present in the spent fuel pool. This boron concentration provides significant margin above the current value (359 ppm) required to maintain an effective neutron multiplication factor (keff) < 0.95 under normal conditions.

Further, this TS value of 1800 ppm boron also provides margin above the current value (910 ppm) required to maintain keff < 0.95 under the limiting accident conditions.

Additionally, plant design features and operator responsiveness ensure that the credible spent fuel pool dilution event (initiated at the TS minimum concentration of 1800 ppm) will be terminated before the Spent Fuel Pool (SFP) boron concentration reaches 750 ppm. This termination point provides ample margin to the current boron concentration (359 ppm) that ensures the limiting normal configuration stays below keff 0.95.Fuel designs employed at PINGP are described in USAR Section 3.1. The original design was Westinghouse 14x14 Standard, and the most recent design in use is the Westinghouse 422 Vantage+ (422V+). However, several variations of 14x14 fuel have been used, including several Exxon designs. In addition to fuel design changes, several core design and operational changes have been made over the Page 4 of 15 L-PI-1 5-087 NSPM Enclosure 1 plant's operating history that would have a bearing how the nuclear fuel is depleted during operation.

For instance, Burnable Poison Rods (BPRs) were inserted into certain unrodded assembly positions for several cycles as a fixed burnable neutron poison. All applicable design variations and operating variations are evaluated in Reference 6.1, WCAP-1 7400 (hereafter referred to as the SFCA).Another variation in fuel design applicable to the SFCA resulted from the fuel consolidation campaign that was conducted in 1987. This consolidation project involved removing the fuel rods from two fuel assemblies and reconfiguring them into a close-packed triangular array; packaged into a specially-design canister.

In this manner, 36 assemblies were consolidated into 18 canisters.

The project is further described in USAR Section 10.2.1.5.Consolidated fuel assemblies and other variations on fuel design (failed fuel baskets) and other spent fuel pool materials of interest (e.g., assembly structural materials from the fuel consolidation project) are described further in the SEGA and supporting calculations.

The proposed amendments involve no physical modifications to the SFP storage racks or to any other system, structure, or component.

3.2 Current

Licensing Basis At a regulatory level, 10 CFR 50.68(a) requires licensees to select one of two options to satisfy criticality accident requirements:

(1) 10 CFR 70.24, or (2) 10 CFR 50.68(b).

In PINGP License Amendments 209/196, NSPM transitioned to fully adopt 10 CFR 50.68(b).

The applicable criticality criteria for the spent fuel storage racks are represented in TS 4.3.1 .1 and summarized below: a. Maximum fuel assembly U-235 enrichment of 5.0 weight percent;b. keff < 1 .0 if fully flooded with unborated water, which includes an allowance for uncertainties as described in USAR Section 10.2;C. keff < 0.95 if fully flooded with water borated to 400 ppm, which includes an allowance for uncertainties as described in USAR Section 10.2;d. A nominal 9.5 inch~ center to center distance between fuel assemblies placed in the fuel storage racks; and e. New or spent fuel assemblies, fuel inserts, and hardware loaded in accordance with TS Figure 4.3.1-1.For the criticality analysis of spent fuel pool abnormal and accident conditions, the current licensing basis uses soluble boron credit and applies the double contingency principle to demonstrate a keff < 0.95 for all postulated scenarios.

This criterion is described in USAR Section 10.2.1. This keff < 0.95 criterion for accidents is more conservative than regulatory guidance which establishes subcriticality (keff < 1.0) as an acceptable limit for accidents.

Page 5 of 1.5 L-PI-15-087 NSPM Enclosure 1 The USAR describes the applicable PINGP General Design Criterion (GDC-66) as follows: Criticality in new and spent fuel storage shall be prevented by physical systems or processes.

Such means as geometrically safe configurations shall be emphasized over procedural controls.

The design and analytical approach to satisfying this criterion is described in USAR Section 10.2.1.The Prairie Island spent fuel racks have been analyzed to allow storage of fuel assemblies with nominal enrichments up to 5.0 weight percent (wlo) uranium-235 (U-235) in all storage cell locations using credit for specific storage arrays, initial enrichment, burnup, and decay time. The analysis does not take any credit for the presence of the spent fuel rack Boraflex neutron absorber panels which are believed to be in a degraded condition.

Currently, the TS and USAR (Section 10.2.1) describe special fuel configurations that deviate from standard fuel assembly construction.

These configurations include the Fuel Rod Storage Canister (FRSC), the Failed Fuel Pin Basket (FFPB), and the Consolidated Rod Storage Canister (CRSC). These have been evaluated for storage limitations as part of the SFCA.3.3 Justification for the Proposed Changes 3.3.1 Justification for Technical Specification Changes In a broad sense, the proposed revisions to TS 3.7.16 (SFP minimum boron concentration) and to TS Table 4.3.1-3 (coefficients to calculate the minimum required fuel assembly burnup) are justified because the new values are supported by approved spent fuel criticality analysis methods (with conservative changes as noted below) and because the resulting changes to TS are incremental to current specifications.

As described in more detail below, these revised TS values can be implemented with little or no change to existing fuel selection and SFP loading procedures.

Therefore, no new human factors considerations are created by the proposed changes.With respect to TS 3.7.16, the increase of SFP minimum soluble boron concentration from 1800 ppm to 2500 ppm is justified because: a. The use of SFP soluble boron to accommodate accidents is justified by the regulation 10 CFR 50.68(b)(4) as well as the current licensing basis which now demonstrates that a soluble boron concentration of 1800 ppm accommodates the limiting non-dilution accident (single assembly misloading accident).

b. Notwithstanding the Double Contingency Principle, extending the licensing basis to include multiple-assembly misloading accidents is a conservative accommodation for an event that may be considered difficult to preclude considering industry operating experience and the Page 6 of 15 L-PI-1 5-087 NSPM Enclosure I fundamental reliance on procedural controls to ensure proper placement of fuel assemblies in the PINGP SFP. NSPM has adopted this change to the misloading analysis (and the accompanying increase in SFP minimum boron concentration limit) because it reduces the effect of human performance errors that might contribute to a misloading event.c. The new soluble boron limit was established to provide margin above the soluble boron concentration calculated for the limiting non-dilution accident (i.e., the 2030 ppm calculated for the multiple-assembly misload).

As discussed in Enclosure 4, the value calculated for the limiting multiple-assembly misload used previously-approved analytical methodologies with appropriate input and model changes to incorporate the IFBA-Gd fuel designs. As discussed in Enclosure 6, the analytical methodologies were sufficiently benchmarked to support analysis of gadolin ia-based neutron absorbers in proximity with boron-based neutron absorbers.

d. Operationally, the increase of soluble boron concentration to 2500 ppm is inconsequential because water chemistry guidelines do not place a maximum limit on the SFP boron concentration, and a level greater than 2500 ppm has been normally maintained for operational convenience to accommodate the minimum concentration required for refueling operations.

e. Increasing the minimum TS concentration from 1800 to 2500 ppm will effectively increase operational margin for mitigating a boron dilution accident , which is analyzed from a starting point of 1800 ppm to an end point of 750 ppm. Enforcing a TS minimum of 2500 ppm will provide plant operators additional time to identify and mitigate a boron dilution event.See Attachment I of this Enclosure for more explanation of the SFP soluble boron concentrations required for the proposed condition, and the available margins. Attachment 1 also includes a comparison to the current condition.

With respect to TS Table 4.3.1-3, the changes to the coefficients for calculating the minimum required fuel assembly burnup are justified because: a. The use of coefficients for calculating the minimum required fuel assembly burnup has been previously approved and implemented at PINGP. A change to the coefficient values does not constitute a new process of any kind; it is incremental to a currently-approved process.SThe boron dilution event analysis supports the current as well as the proposed SFP soluble boron requirements.

Therefore, no revision is required to support the proposed amendment.

Refer to Attachment I of this Enclosure to see how the boron dilution event relates to the current and proposed SFP soluble boron requirements.

Page 7 of 15 L-PI-1 5-087 NSPM Enclosure 1 Thus, the revised coefficient values do not require any new human factors considerations.

b. The objective of these revised coefficients is to achieve the subcriticality criteria prescribed by regulation 10 CFR 50.68(b)(4) with consideration of the planned use of IFBA-Gd fuel design. Enclosure 4 demonstrates how these criteria will continue to be met with the proposed change to coefficients.

c. The new coefficients were calculated using previously-approved analytical methodologies with appropriate input and model changes to incorporate the IFBA-Gd fuel designs. As discussed in Enclosure 6, the analytical methodologies were sufficiently benchmarked to support analysis of gadolinia-based neutron absorbers in proximity with boron-based neutron absorbers.

Enclosure 4 summarizes the analysis that provides the new coefficient values for TS Table 4.3.1-3.d. The revised coefficients result in changes to burnup requirements that are up to 4 GWD/MTU higher than existing requirements.

Such a change will not significantly affect the current spent fuel pool configuration.

Based on a preliminary estimate, few spent fuel assemblies would have to be re-assigned to a more-reactive fuel category and relocated in the spent fuel pooi to align with the revised coefficients.

e. For Fuel Not Operated in Cycles 1-4, the revised coefficients and reanalysis of loading patterns would result in a new reactivity condition for the normal loading configurations that requires a soluble boron-concentration (to achieve keff 0.95) that is lower than previously analyzed.

However, as described in Enclosure 4, the results for Fuel Operated in Cycles 1-4 (which is unaffected by IFBA) sustain the limiting soluble boron condition of 359 ppm. Refer to See Attachment 1 of this Enclosure for more explanation of the SFP soluble boron concentrations required for the proposed condition, and the available margins. Attachment 1 also includes a comparison to the current condition.

3.3.2 Justification

for Spent Fuel Criticality Analysis Methodology Changes The adoption of a certain bias for minor actinide and fission product worth is consistent with the regulatory guidance (NUREG/CR-7109) and precedent established by the precedent analysis submitted in support of Reference 6.5. Refer to Enclosure 4 Section S4.1 .2.1.4 for further explanation of this biased treatment of actinide and fission product worth.Whereas boron-based (i.e., IFBA) fuel rods have not been explicitly modeled and analyzed in combination with gadolinia-based fuel rods for PINGP, Westinghouse Electric Company, LLC (WEC) reviewed the applicability of the neutronic code suite (ALPHA / PHOENIX-P or PARAGON / ANC) for determining axial power shapes and burnup profiles Page 8 of 15 L-PI-1 5-087 NSPM Enclosure 1 of this configuration and concluded that the currently-approved analytical methods are valid for the intended application proposed herein (i.e., IFBA in combination with gadolinia fuel rods). This evaluation is provided in Enclosure 6.3.3.3 Justification for Other Changes to the Current Licensing Basis The current licensing basis spent fuel criticality analysis conservatively ignores the effect of the current neutron absorber (gadolinia) because it is a net poison throughout its exposure over the operating cycle. However, the nature of the new proposed neutron absorber (boron, in the form of zirconium diboride IFBA) depletes differently such that it cannot always be viewed as a net poison throughout the operating cycle. Thus, IFBA is explicitly modeled in the approved computer codes that analyze the nuclear fuel as it depletes in the reactor, and it is conservatively ignored as a neutron absorber in the computer models that analyze criticality of fuel in the SFP storage configurations.

3.4 Conclusion

The proposed changes to the Technical Specifications and to the SFCA model are incremental to the current licensing basis and are readily justified because the methods and results continue to meet the prevailing standards.

None of the changes affect a system, structure, or component, and none result in a change to how systems are operated.

in that regard, the proposed changes do not create a new challenge to human performance nor increase the probability of a previously-evaluated accident or malfunction.

4.0 REGULATORY EVALUATION

4.1 Applicable

Regulatory RequirementslCriteria The US Atomic Energy Commission (AEC) issued their Safety Evaluation (SE) of the PINGP on September 28, 1972. The SE, Section 3.1, "Conformance with AEC General Design Criteria," described the conclusions the AEC reached associated with the General Design Criteria in effect at the time. The AEC stated: The Prairie Island plant was designed and constructed to meet the in tent of the AEC's General Design Criteria, as originally proposed in July 1967. Construction of the plant was about 50% complete and the Final Safety Analysis Report (Amendment No. 7) had been filed with the Commission before publication of the revised General Design Criteria in February 1971 and the present version of the criteria in July 1971. We did not require the applicant to reanalyze the plant or resubmit the FSAR. Howe ver, our technical review did assess the plant against the General Design Criteria now in effect and we are satisfied that the plant design generally conforms to the intent of these criteria.Page 9 of 15 L-PI-1 5-087 NSPM Enclosure 1 Based on the above, the applicable PINGP GDC states: Criticality in spent fuel storage shall be prevented by physical systems or processes.

Such means as geometrically safe configurations shall be emphasized over procedural controls.On September 29, 2011, the NRC staff issued the Interim Staff Guidance (ISG)DSS-ISG-2010-01 (Reference 6.2). The purpose of the ISG is to provide updated review guidance to the NRC staff to address the increased complexity of recent SFP nuclear criticality analyses and operations.

The ISG rebaselines NRC's expectations for spent fuel criticality analysis.

The expectations of the ISG were further reinforced in subsequent NRC Information Notice 2011-03 (Reference 6.3).The Commission's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36. The TS requirements in 10 CFR 50.36 include the following categories:

(1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

The requirements for system operability during movement of irradiated fuel are included in the TSs in accordance with 10 CFR 50.36(c)(2), Limiting Conditions for Operation.

As required by 10 CFR 50.36(c)(4), design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not*covered in categories described in paragraphs (c)(1), (2), and (3) of 10 CFR 50.36.This amendment request concerns 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(4).

Paragraph 50.68(b)(1) of 10 CFR requires, "Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water." Paragraph 50.68(b)(4) of 10 CFR requires, "If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water." The U.S. Atomic Energy Commission (AEC) issued its Safety Evaluation (SE) for PINGP before the revised General Design Criteria (GDCs) were published in 1971.A PINGP GDC requires that, "Criticality in new and spent fuel storage shall be prevented by physical systems or processes.

Such means as geometrically safe configurations shall be emphasized over procedural controls." As guidance for reviewing criticality analyses of fuel storage at light-water reactor power plants, the NRC staff issued an internal memorandum on August 19, 1998 (ADAMS Accession No. ML00372B001).

This memorandum is known as the Page 10 of 15 L-Pl-1 5-087 NSPM Enclosure 1"Kopp Letter." The Kopp Letter provides guidance on salient aspects of a criticality analysis.

The guidance is germane to boiling-water reactors and pressurized-water reactors, and to borated and unborated conditions.

Additional guidance is available in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," particularly Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," Revision 3, issued March 2007. Section 9.1.1 provides the existing recommendations for performing the review of the nuclear criticality safety analysis of SFPs.4.2 Precedent There is little precedent that is applicable to the proposed activity because of the following factors: a. Based on the recent review and approval of PINGP Unit 1 and 2 license amendments (in 2013 per Reference 6.4) that explicitly addressed the cited Interim Staff Guidance and contemporaneous precedent, there has been little opportunity for new developments.

b. The incremental changes of this LAR are of such limited scope that the potential for impacts from other licensing activities (whether plant-specific or topical) is small.Notwithstanding the above, one precedent licensing activity with practical impact on the proposed amendment stems from the regulatory review performed for Comanche Peak (Reference 6.5) with respect to human performance errors that could lead to a SFP misloading event where several assemblies are misloaded in series due to a common cause. Whereas Comanche Peak made an extensive justification of its fuel selection and inventory process to effectively preclude such an event, NSPM has chosen an analytical approach.

Accordingly, this precedent was addressed in Enclosure 4 with due consideration and analysis of a multiple fuel assembly misload event in the PINGP spent fuel criticality analysis.The Comanche Peak amendment also set precedent for adopting a certain bias for minor actinide and fission product nuclides.

This precedent is addressed in Enclosure

4.4.3 Significant

Hazards Consideration Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, proposes to amend the renewed operating licenses of Prairie Island Nuclear Generating Plants (PINGP) Units 1 and 2. The purpose of this amendment is to modify the PINGP Technical Specifications (TS) to allow spent fuel pooi (SFP) storage of nuclear fuel containing a boron-based neutron absorber in the form of zirconium diboride Integral Fuel Burnable Absorber (IFBA).Page 11 of 15 L-PI-1 5-087 NSPM Enclosure 1 The proposed revisions involve an incremental increase to the minimum required value for Spent Fuel Pool (SFP) boron concentration and incremental change to the coefficients used to calculate the minimum required fuel assembly burnup for establishing fuel storage categories for safe loading patterns.

These revised TS values can be implemented with minimal change to existing fuel selection and SEP loading procedures, and do not involve any change to plant systems, structures, components or to the processes for fuel handling.NSPM has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No.The proposed amendments do not change or modify the fuel, fuel handling processes, fuel storage racks, number of fuel assemblies that may be stored in the spent fuel pool (SFP), decay heat generation rate, or the SFP cooling and cleanup system. The proposed amendment was evaluated for impact on the following previously-evaluated criticality events and accidents and no impacts were identified:

(1) fuel assembly misloading, (2) loss of spent fuel pool cooling, and (3) spent fuel boron dilution.Operation in accordance with the proposed amendment will not change the.probability of a fuel assembly misloading because fuel movement will continue to be controlled by approved fuel selection and fuel handling procedures.

These procedures continue to require identification of the initial and target locations for each fuel assembly and fuel assembly insert that is moved. The consequences of a fuel misloading event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the worst-case fuel misloading event.Operation in accordance with the proposed amendment will not change the probability of a loss of spent fuel pool cooling because the change in fuel burnup requirements and SFP boron concentration have no bearing on the systems, structures, and components involved in initiating such an event. The proposed amendment does not change the heat load imposed by spent fuel assemblies nor does it change the flow paths in the spent fuel pool. Finally, a criticality analysis of the limiting fuel loading configuration confirmed that the condition would remain subcritical at the resulting temperature value.Therefore, the accident consequences are not increased for the proposed amendment.

Page 12 of 15 L-PI-1 5-087 NSPM Enclosure 1 Operation in accordance with the proposed amendment will not change the probability of a boron dilution event because the incremental changes in TS values have no bearing on the systems, structures, and components involved in initiating or sustaining the intrusion of unborated water to the spent fuel pool.The consequences of a boron dilution event are unchanged because the proposed amendment has no bearing on the systems that operators would use to identify and terminate a dilution event. Also, 'implementation of the proposed amendment will not affect any of the other key parameters of the boron dilution analysis which includes SFP water inventory, volume of SFP contents, the assumed initial boron concentration of the accident, and the sources of dilution water. Finally, a criticality analysis of the limiting fuel loading configuration confirmed that the dilution event would be terminated at a soluble boron concentration value that ensured a subcritical condition.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of a criticality accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No.The proposed changes involve incremental changes to TS values, and represent minimal change to existing fuel selection and SEP loading procedures.

Further, the proposed changes involve no change to plant systems, structures, components or to the processes for fuel handling.

The proposed changes do not involve new SFP loading configurations and do not change or modify the fuel, fuel handling processes, fuel storage racks, number of fuel assemblies that may be stored in the pool, decay heat generation rate, or the spent fuel pool cooling and cleanup system. As such, the proposed changes introduce no new material interactions, man-machine interfaces, or processes that could create the potential for an accident of a new or different type.3. Do the proposed changes involve a significant reduction in a margin of safety?Response:

No.The proposed change was evaluated for its effect on current margins of safety as they relate to criticality.

The margin of safety for subcriticality required by 10 CFR 50.68 (b)(4) is unchanged.

The new criticality analysis confirms that operation in accordance with the proposed amendment continues to meet the required subcriticality margin. Increasing the minimum SFP soluble boron concentration ensures that subcriticality margins will be preserved, and increases the margin of safety associated with a boron dilution event.Page 13 of 15 L-PI-1 5-087 NSPM Enclosure 1 Therefore, the proposed changes do not involve a significant reduction in the margin of safety.Therefore, based on the above, NSPM has concluded that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly a finding of "no significant hazards consideration" is justified.

4.4 Conclusions

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATIONS 10 CFR 51 .22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment.

A proposed amendment of an operating license for a facility requires no environmental assessment if the operation of the facility in accordance with the proposed amendment does not: (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (3)result in a significant increase in individual or cumulative occupational radiation exposure.

NSPM has reviewed this LAR and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9).

Pursuant to 10 CFR 51 .22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The basis for this determination follows.1. As demonstrated in the 10 CFR 50.92 evaluation, the proposed amendment does not involve a significant hazards consideration.

2. The proposed amendment does not result in a significant change in the types or increase in the amounts of any effluents that may be released offsite.Implementation of the proposed changes involves no physical change to the nuclear fuel or the types of exposure it would receive. Nor does it involve the physical change to any system, structure, or component.

3. The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure.

Implementation of the proposed amendment will not involve a significant amount of fuel movements.

Aside from the small amount of individual and cumulative occupational radiation exposure resulting from such movements, the proposed changes will not result in any unusual spent fuel pool operations that would result in a Page 14 of 15 L-PI-1 5-087 NSPM Enclosure 1 permanent effect to increase occupational exposure.

The proposed fuel storage configurations do not fundamentally change the inventory or radiological source term of the spent fuel. In addition, based on NSPM's experience with routine fuel movement campaigns during refueling outages, the cumulative exposure from the proposed activities is expected to be minimal.

6.0 REFERENCES

6.1 Westinghouse

Report WCAP-1 7400-P, Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis, Revision 0, dated July 2011 (submitted as Enclosure to Xcel Energy Letter to NRC dated August 19, 2011 (ADAMS Accession No. MLl12360231)

6.2 Interim

Staff Guidance DSS-ISG-2010-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, dated September 29, 2011 (ADAMS Accession No. ML1 10620086)6.3 NRC Information Notice 2011-03, Nonconservative Criticality Safety Analyses for Fuel Storage, dated February 16, 2011 (ADAMS Accession No. ML1 03090055)6.4 Prairie Island Units I and 2 Operating License Amendment Nos. 209/1 96 and NRC Safety Evaluation Report (SER) dated August 29, 2013 6.5 Comanche Peak Units 1 and 2 Operating License Amendment No. 162 and NRC SER dated July 1, 2014 (ADAMS Accession No. ML14160A035) 6.6 NRC (Terry Beltz) letter to Xcel Energy, "Summary of the April 14, 2015, Public Meeting with Xcel Energy and Westinghouse to Discuss a. Potential Future License Amendment Request Regarding the Use of Integral Fuel Burnable Absorber Neutron Absorbers in Westinghouse 422V+ Fuel Assembly Design (TAC NOS.MF5839 AND MF5840)," dated May 15, 2015 (ADAMS Accession No.ML1 51 07A059)Page 15 ofI15 L-PI-1 5-087 Enclosure 1, Attachment I, Comparison of SFP Boron Requirements NSPM Purpose: This attachment describes how the revised Spent Fuel Pool (SFP) soluble boron requirements and the revised TS 3.7.16 limit for minimum SFP boron concentration affect the margins to limiting conditions in the SFP. Please refer to the graphic below (Figure A-I) and note that "Current Licensing Basis" relates to the current conditions, and "Proposed Licensing Basis" relates to the conditions proposed in the license amendment request.Figure A-I Comparison of SFP Boron Requirements (Current vs. Proposed)Current Licensing Basis I SFP Boron Concentration (ppm) r , ...2500 TS Minimum (2500)Licensing Basis 2400 -4 2300 --2200 -"TS Minimum (1800)---Start 2100 -2000 -1900 -1700-15600-E- Minimum for Non_-Dilution Accidents (2030)(Analytical value for multiole assembly misload)Start 1300--Minimum for No._n-Dilution Accidents (910) (analytical value for assembly misload)Stop! 0-700 ---m.o m Stop 600 ---500 --Minimum "IS Value for Normal Keff 0.95 (400) ->4 00 Minimum TS Value (400) for Normal Keff 0.95 Limiting Configuration Normal Keff 0.95 (359) 300 -LitngCfgutonNrlKe 09(3)200 --100 --0-Page 1 of 2 L-PI-1 5-087 NSPM Enclosure 1, Attachment 1, Comparison of SEP Boron Requirements

Soluble Boron Concentration (SBC) margqin for the normal SFP conditions.

As described in the TS Bases, the TS 4.3.1.1 .c value for maintaining keff< 0.95 under normal conditions (i.e., 400 ppm) was conservatively chosen to be higher than the limiting normal SFP criticality condition in the criticality analysis.

The difference between 400 ppm and the SBC at the limiting normal condition in the analysis provides administrative margin to accommodate a future analysis error. The LAR proposes no change to the TS 4.3.1.1 .c value and no change to the value that achieves the limiting SBC for the normal condition.

Therefore, the margin is unchanged by the proposed amendment.

SBC mar qin for Boron Dilution Event mitigation.

As shown in Figure A-i, the Boron Dilution Event has not been reanalyzed for the proposed amendment; the event is still postulated to start at 1800 ppm and progress to 750 ppm. Within this SBC range in the dilution analysis, the analysis shows that the time of dilution provides ample time for operators to identify and mitigate the event. Thus, any margin above 1800 ppm provides incrementally more time for operators to respond to the dilution event. Thus, the proposed TS 3.7.16 change to a minimum SBC of 2500 ppm increases the margin considerably (700 ppm).*SBC margqin for Non-Dilution Accidents.

The Double Contingency Principle precludes a boron dilution event in combination with a non-dilution event such as the limiting misloading.

Thus, the SBC margin to accommodate a misloading (non-dilution event) is irrelevant because no dilution need be considered.

Nevertheless, a nominal discussion of margin is provided below.The proposed amendment takes a more conservative approach to postulating non-dilution accidents by accepting the possibility of multiple fuel assembly misloading, when only one misloading was previously assumed. Therefore, the revised analysis of the multiple misloading requires a much higher SBC of 2030 ppm (to maintain a keff < 0.95), which appears to significantly reduce the margin to the TS minimum SBC as follows:* In the current condition, the margin from TS 3.7.16 minimum SBC to the minimum required for the misloading event is 890 ppm (1800 ppm minus 910 ppm).* In the proposed condition, the margin is reduced to 470 ppm (2500 ppm minus 2030 ppm).This reduction in margin should not be concerning because: (1) the Double Contingency Principle neutralizes the effect (as discussed above), and (2) the reduction in margin resulted from the conservative adoption of a multiple-misloading event that could have been previously included in the licensing basis. There is nothing inherent to the proposed use of IFBA that would increase the probability of misloading accidents.

Page 2 of 2 L-PI-1 5-087 NSPM Enclosure 2 Enclosure 2 Marked-Up Technical Specification Pages 3 pages follow 3.7.16-1 4.0-7 Insert Spent Fuel Storage Pool Boron Concentration 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Storage Pool Boron Concentration LCO 3.7.16 The spent fuel storage pool boron concentration shall be > 4-800 ppm.APPLICABILITY:

When fuel assemblies are stored in the spent fuel storage pool.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool--------NOTE-----

boron concentration not LCO 3.0.3 is not applicable.

within limit.A. 1 Suspend movement of fuel Immediately assemblies in the spent fuel storage pool.AND A.2 Initiate action to restore spent Immediately fuel storage pool boron concentration to within limit.Prairie Island Units 1 and 2 Unit 1 -Amendment No. 4-48 Unit 2 -Amendment No. 4-49 3.7.16-1 Replce al deetedvalus lDesign Features withvales n th atachd ~Table 4.3.1-3 (page 1 of 1)For Fuel Not Operated In Units 1 and 2 Cycles 1 -4 Coefficients to Calc late the Minimum Required Fuel Assembly Burnup (Bu) as a Fu) ction of Decay Time and Enrichment (En)FUEL DECAY TIME \COEFFICIENTS lA 3A 2 0 -066 9.4 -3.8 0 -0,42,0 4-30 320 -0,404-$ 2,4 4. -4..46 15 .4.81 2.4 4.6 20 .441 4.3 .484 40 4-3 4486 510 -2.4 2484-3.0 15 443 4404.24 20 -042 2.2 2.44-44 Notes: 1.All relevant uncertainties are explicitly included in the criticality analysis.

For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required.

For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed "minimum burnup" (GWdI/MTU) given by the curve fit for the assembly "decay time" and"initial enrichment".

The specific minimum burnup required for each fuel assembly is calculated from the following equation for each increment of decay time: Bu = A *En 3 + A 2*En 2 + A 3*En + A4 2. Initial enrichment (En) is the nominal U-235 enrichment.

Any enrichment between 1.7 and 5.0 weight percent U-235 may be used. If the computed Bu value is negative, zero shall be used.3. Decay Time is in years. An assembly with a cooling time greater than 20 years must use 20 years. No extrapolation is permitted.

4. If Decay Time value fails between increments of the table, the lower Decay Time value shall be used or a linear interpolation may be performed as follows: Compute the Bu value using the coefficients associated with the Decay Time values that bracket the actual Decay Time. Interpolate between Bu values based on the increment of Decay Time between the actual Decay Time value and the computed Bu results.5. This table applies to fuel assemblies that were not operated in the Unit 1 or Unit 2 core during operating Cycles 1 through 4.Prairie Island Units 1 and 2 Unit 1 -Amendment No. 20-Unit 2 -Amendment No. 94 4.0-7 Insert to Table 4.3.1-3 Replace the Coefficient (A 1 , A 2 , A 3 , A 4) values with those shown below Fuel Coefficients Category Decay Time A1 A2 A3 A4 2 0 -1.9089 22.9292 -81.9646 91.4193 o -0.0536 0.5516 8.2824 -23.3157 5 -0.0372 0.2803 9.0736 -23.8543 310 -0.0408 0.2587 9.0667 -23.6452 15 -0.0893 0.7485 7.2536 -21.4102 20 -0.1011 0.8822 6.6122 -20.4468 4 0 1.3659 -14.9709 63.0347 -72.9223 0 0.2744 -3.7275 29.5218 -41.7174 5 0.0533 -1 .3478 20.6704 -32.3235 5 10 -0.0407 -0.3472 16.7092 -27.9591 15 -0.1809 1.0636 11.8632 -23.0476 20 -0.0897 0.2312 13.9007 -24.5529 0 0.4604 -5.9192 38.3216 -50.3021 5 0.4161 -5.2825 34.6238 -45.6381 6 10 0.3716 -4.7154 31.7812 -42.2260 15 0.1816 -2.7038 24.7285 -35.1164______ 20 0.1318 -2.1711 22.5833 -32.7644 L-PI-1 5-087 NSPM Enclosure 3 Enclosure 3 Marked-Up Technical Specification Bases Pages 8 pages follow B3.7.1 6-2 B3.7.1 6-3 B3.7.1 6-4 B3.7.1 6-5 B3.7.1 7-2 B3.7.1 7-5 B3.7.1 7-6 B3.7.1 7-10 Insert Fuel Storage Pool Boron Concentration B 3.7.16 BASES (continued)-and Ref 5I APPLICABLE SAFEITY ANALYSES The spent fuel pool criticality analysis (Ref. 4) addresses all the fuel types currently stored in the spent fuel pool and in use in in the reactor. The fuel types considered in the analysis include the Westinghouse Standard (STD), OFA, and Vantage Plus designs, (both 0.400" and 0.422" O.D. designs) and the Exxon fuel assembly types in storage in the spent fuel pool.~Accident conditions which could increase the keff were evaluated Dropped and. icung..misplaced fresh incluing:locations; and Ifuel assemblies; a-". A new fuel, sse.m.blv, drop on t~he of the racks,;'misloaded betw-een rack modules;,misloaded into au incorrect storage rack d. Tntramodule w..ate=r gap reduc,,tio-n to. a o, .ei ...c evn; and@. Spent fuel pool temperature greater than 150 °F.J INew fueFor an occurrence of these postulated accident conditions, the double contingency principle of Reference 2 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.

Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 359 ppm required to maintain 1kf less than 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its lanalytically presence would be a second unlikely event.--,,L_./-lan Ref5/Calculations were performed (Ref. 4) to determine the amount of praph [soluble boron required to offset the highest reactivity increase the lastI caused by these postulated accidents and to maintain k~f less than or equal to 0.95. It was found that a spent fuel pool boron cocnrtono

-- ppm (assuming a conservatively low boron-l10 12030 Iatom percent of 19.4) was adequate to mitigate these postulated criticality related accidents and to maintain k~ff less than or equal to 0.95. This specification ensures the spent fuel pool contains Insert parag provided on page in this Enclosure Prairie Island Units 1 and 2 Unit Reviion I Unit 2 -Revision 2,-1 B 3.7.16-2 Fuel Storage Pool Boron Concentration B 3.7.16 BASES APPLICABLE SAFETY AMNA 1 Vq1l'Z adequate dissolved boron to compensate for the increased reactivity caused by these accidents.(c ontinued) spent fuel pool boron dilution analysis was performed which/confirmed that sufficient time is available to detect and mitigate a/dilution of the spent fuel pool before the 0.95 keff design basis is/exceeded.

The spent fuel pool boron dilution analysis concluded]that an unplanned or inadvertent event which could result in the/ dilution of the spent fuel pool boron concentration from 1800 ppm/ to 750 ppm is not a credible event./ The current spent fuel rack criticality analysis (Ref. 4) only require] a boron concentration of 359 ppm (assuming a conservatively low[ boron-10 atom percent of 19.4) to ensure that the spent fuel rack ke[ will be less than or equal to 0.95 for the allowable storage[ configuration, excluding accidents.

Therefore the spent fuel pool] boron dilution analysis which assumes 750 ppm as the endpoint of/ the analysis is conservative with respect to the endpoint of 359 ppn/ since a larger volume of water would be required, which would tak~more time to dilute the spent fuel pool to 359 ppm.The concentration of dissolved boron in the fuel storage pool~satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

]iff:e To establish the most limiting non-dilution accident configuration, the criticality analysis assumed an extensive array of fresh unpoisoned fuel. This configuration required a minimum boron concentration of 2030 ppm (at a conservatively low boron-10 concentration of 19.4 atom percent) to achieve keff less than or equal to 0.95. The TS 3.7.16 limit of 2500 ppm ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by this accident.Prairie Island Units 1 and 2 Unit 1 -Revision ;!2-1-Unit 2 -Revision 2241-B 3.7.16-3 Fuel Storage Pool Boron Concentration B 3.7.16 LCOThe fun/storage pooi boron concentration is required to be land__5 I> 800 ppm. The specified concentration of dissolved boron in the /fuel storage pool preserves the assumptions used in the analyses of _i the potential critical accident scenarios as described in Reference 4.'This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.APPLICABIITUY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.ACTIONS A. 1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.When the concentration of boron in the spent fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies.

The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.

This does not preclude movement of a fuel assembly to a safe position.If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable.

If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.Prairie Island Unit 1 -Revision ;2~-1-Units 1 and 2 B 3.7.16-4 Unit 2 -Revision 22-1 Fuel Storage Pool Boron Concentration B 3.7.16 BASES (continued)

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed.

The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time.REFERENCES

1. USAR, Section 10.2.2. ANSI/ANS-8.1-1983.

3. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees from B. k. Grimes, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", April 14, 1978.4. "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Analysis", WCAP- 17400-NP, Revision 0, Westinghouse Electric Company, July 2011.5. Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Analysis ISupplement Analysis for the Storage of IFBA Bearing Fuel, WCAP-17400-P, ISupplement 1, Rev 1, Westinghouse Electric Company, October 2015.Prairie Island Units 1 and 2 Unit 1 -Revision 2,2--Unit 2 -Revision ;!2-4,I B 3.7.16-5 Spent Fuel Pool Storage B 3.7.17 BASES (continued)

APPLICABLE SAFEIY ANALYSES Per Reference 5, the presence of an Integral Fuel Burnable Absorber (IFBA) is considered for the 422V+ fuel design.The hypothetical criticality accidents can only take place during or as a result of the movement of an assembly (Ref. 4 and 5). For these accident occurrences, the presence of soluble boron in the spent fuel storage pool (controlled by LCO 3.7.16, "Fuel Storage Pool Boron Concentration")

prevents criticality.

By closely controlling the movement of each assembly and by verifying the appropriate checkerboarding after each fuel handling campaign, the time period for potential accidents may be limited to a small fraction of the total operating time. During the remaining time period with no potential for criticality accidents, the operation may be under the auspices of the accompanying LCO. ir-and 5 The spent fuel storage racks have been an, zed in accordance with the methodology contained in Reference

4. That methodology ensures that the spent fuel rack multiplication factor, lkf, is less than the values required by 10 CFR 50.68(b).

The codes, methods and techniques contained in the methodology are used to satisfy these criteria for keff. The resulting Prairie Island spent fuel rack criticality analysis allows for the storage of fuel assemblies with enrichments up to a maximum of 5.0 (nominal 4.95% + 0.05%) weight percent U-235 while maintaining lYf < 1.0 (including uncertainties) if flooded with unborated water and k~f <0.95 (including uncertainties) with credit for soluble boron. The analysis determined at" a minimum soluble boron concentration of 359 pm(at a con v'eatively low boron-l0 atom percent of 19.4) will ensure any loaded c figuration k~f will be < 0.95. In addition, the analysis differentiate~a f uel assembly operated during Operating Cycle 1 -4 from an nassembl..rated after Cycle 4 in determining the assembly's reactivity.

Credit is taken for the radioactive decay time of the spent fuel. No credit is given for any gadoliniurrburnable poison in the fuel. [rIB /The criticality analysis (Ref. 4 ,specifically analyzed each of the following storage to ensure that the spent fuel pool will remain subcritical when fud is placed in accordance with Specification 4.3.1.1.an5 Prairie Island Units 1 and 2 Unit 1 -Revision ;22-1-Unit 2 -Revision 2,2-1-.B 3.7.17-2 Spent Fuel Pool Storage B 3.7.17 BASES APPIJCABLE SA-ETY ANALYSES (continued) modules because all the racks in the SFP have identical fuel cell design and the actual physical gap between rack modules is ignored in the analysis (i.e., there is no credit taken for the gaps between rack modules).Array interface requirements:

Technical Specifications provide only one special interface requirement between different arrays.This specific interface is described in Figure 4.3.1-1 Note 7 (Array F shall interface only with Array A) and was specifically analyzed.

Otherwise, the Technical Specifications do not provide any unique rules for the interface between arrays.Rather, the Technical Specifications require that all fuel in the spent fuel pool satisfy one of the required arrays, even in transitions between two major arrays.Specification 3.7.17 and Specification

4.3 ensure

that fuel is stored in the spent fuel racks in accordance with the storage configurations assumed in the spent fuel rack criticality analysis (Ref. The spent fuel pool criticality analysis addresses all the fuel types currently stored in the spent fuel pool and in use in the reactor. The fuel types considered in the analysis include the Westinghouse Standard (STD), OFA, and Vantage Plus designs (both 0.400" and 0.422" O.D. designs), and the Exxon fuel assembly types in storage in the spent fuel pool.Accident conditions which could increase the keff were evaluated Dropped and misplaced fresh fuel assembly a. A new fuel assembly drop on the top of the racks;Inadertet cb. A new fuel assembly misloaded b removal of an c.A new fuel assembly, misloaded ii RCCA!oain:etween rack modules: nto an-incorrect storage rack New fuel assemblies

! !Prairie Island Units 1 and 2 Unit 1 -Revision ;2,2-1-Unit 2 -Revision 22-1-B 3.7.17-5 Spent Fuel Pool Storage B 3.7.17 BASES APPLICABLE

d. ntramodu..

water... ga reductA,,-ion, due, to ... seismic e.ent. and SAFETY ANALYSES Spent fuel pool temperature greater than 150°F.(continued)

For an occurrence of these postulated accident conditions, the double contingency principle of Reference 2 can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.

Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 359 ppm required to maintain kef less than 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.Iand Ref 5 l f Westinghouse Electric Company LLC calculations (Ref. 4) were ..2030 Insert paragraph performed to determine the amount of soluble boron required4 o provided on last offset the highest reactivity increase caused by these po s~ated page in this [accidents and to maintain keff less than or equal t .7. It was found Enclosure Ithat a spent fuel pool boron concentration of 94 ppm (assuming a conservatively low boron-lO atom percent of 19.4) was adequate to mitigate these postulated criticality related accidents and to maintain lqff less than or equal to 0.95.Specification 3.7.16 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by a-mispositioned fuel aemxyor aloss of spent fuel pool cooling.Specification

4.3 requires

that the spent fuel rack keff be less than or equal to 0.95 when flooded with water borated to 400 ppm. This value was selected to provide a nominal margin above the calculated limiting value of 359 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 keff design basis is exceeded.

The spent fuel pool boron dilution analysis concluded that sufficient time would be available for operators to recognize and terminate a dilution event that started at Prairie Island Unit 1 -Revision 224-1 Units 1 and 2 B 3.7.17-6 Unit 2 -Revision 22.1-Spent Fuel Pool Storage B 3.7.17 BASES REFERENCES (continued)

4. "Prairie Island Units 1 2 et Fuel Pool Criticality Analysis", WCAP-1740 -NP, R lision 0, Westinghouse Electric Company, July 5. Not 6. A Nuclear Society, "American National Standard Desig Requirements for Light Water Reactor Fuel Storage at Nuclear Power Plants", ANSI/ANS-57.2-1983, October 193."Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Analysis Supplement Analysis for the Storage of IFBA Bearing Fuel," WCAP-1 7400-P, Supplement 1, Rev 1, Westinghouse Electric Company, October 2015.Prairie Island Units 1 and 2 Unit 1 -Revision ;!2-1 Unit 2 -Revision 2,-1.B 3.7.17-10 Insert the followincq paracqraph on pacqes 3.7.16-2 and 3.7.17-6: In recognition of industry operating experience that multiple fuel assemblies have been coincidentally misloaded in spent fuel pools, the PING P licensing basis criticality analysis has adopted the possibility of a multiple-assembly misloading accident and determined that subcriticality requirements can be met with a concentration of soluble boron that does not exceed the TS 3.7.16 minimum concentration.

Thus, consistent with the double contingency principle, a multiple-assembly misloading is adopted as an unlikely event that need not be assumed to occur coincidentally with another unlikely, independent event (such as a dilution event).