Text

Prairie Island, Units 1 and 2, 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 PlantXcel ne_.*rav1717 Wakonade Drive EastXcel~ ergyWelch, MN 55089L-P I-1 5-087November 17, 2015 10 CFR 50.90U.S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2Dockets 50-282 and 50-306Renewed License Nos. DPR-42 and DPR-60License Amendment Request for Spent Fuel Pool Criticality Technical Specification ChangesPursuant 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 FuelStorage Pool Boron Concentration,"

and TS 4.3.1, "Fuel Storage Criticality,"

to allowspent fuel pool (SFP) storage of nuclear fuel containing a boron-based neutronabsorber in the form of zirconium diboride (ZrB2) Integral Fuel Burnable Absorber(IF BA).Enclosure 1 to this letter provides the evaluation of the proposed TS changes and theirsupporting justifications, including a no significant hazards determination.

Enclosure 2provides the current TS pages marked-up to show the proposed changes.

Enclosure 3provides, for information only, the current TS Bases pages marked-up to show theassociated proposed Bases changes.

Final TS Bases changes will be implemented pursuant to TS 5.5.12, "Technical Specifications (TS) Bases Control Program,".

at thetime 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 PoolCriticality Analysis

-Supplemental Analysis for the Storage of IFBA Bearing Fuel,"dated October 2015. This report provides the analytical basis for the revised TS. Thisreport 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 codesused in the supporting spent fuel criticality analysis remain valid for modeling fuel Document Control DeskPage 2assemblies containing both IFBA and gadolinia absorber rods at PINGP. The documentis entitled, "Modeling of Fuel Assemblies Containing both IFBA and Gadolinia AbsorberRods with Westinghouse Core Design Code Systems."

Enclosure 6 containsproprietary information.

Enclosure 7 provides the non-proprietary version of thedocument.

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 theinformation may be withheld from public disclosure by the NRC and .addresses withspecificity the considerations listed in 10 CFR 2.390(b)(4).

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

Accordingly, itis respectfully requested that the information which is proprietary to WEC be withheldfrom public disclosure in accordance with 10 CFR 2.390.Correspondence with respect to the copyright or proprietary aspects of the itemsprovided 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 ofeffluent

release, or result in any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendment meets thecategorical exclusion requirements of 10 CFR 51 .22(c)(9) and an environmental impactassessment need not be prepared.

A copy of this submittal, including the Determination of No Significant HazardsConsideration, 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. Onceapproved,

.the amendment will be implemented within 120 days.If there are any questions or if additional information is needed, please contact GlennAdams at 612-330-6777.

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

Document Control DeskPage 3I declare under penalty of perjury that the foregoing is true and correct.Executed on November 17, 2015Kevin DavisonSite Vice President, Prairie Island Nuclear Generating PlantNorthern States Power Company -Minnesota Enclosures (8)cc: Regional Administrator, Region Ill, USNRCProject Manager, Prairie Island Nuclear Generating Plant, USNRCResident Inspector, Prairie Island Nuclear Generating Plant, USNRCState of Minnesota (without enclosures 2 through 8)

L-PI-1 5-087 NSPMEnclosure 1ENCLOSURE 1Evaluation of the Proposed ChangeLicense Amendment Request forSpent Fuel Pool Criticality Technical Specification Chanqes1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Proposed Change to TS 3.7.16, "Spent Fuel Storage Pool BoronConcentration" 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 Basis3.0 TECHNICAL EVALUATION 3.1 Design Description 3.2 Current Licensing Basis3.3 Justification for the Proposed Changes3.4 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 1Page 1 ofi15 L-PI-1 5-087 NSPMEnclosure 11.0 SUMMARY DESCRIPTION Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota Corporation (NSPM), doing business as Xcel Energy, hereby requests anamendment to the renewed operating licenses for Prairie Island NuclearGenerating Plant (PINGP).

Specifically, NSPM proposes to revise Technical Specification (TS) 3.7.16, "Spent Fuel Storage Pool Boron Concentration" and TS4.3.1, "Fuel Storage Criticality" to allow spent fuel pool (SFP) storage of nuclearfuel containing a boron-based neutron absorber in the form of zirconium diboride(ZrB2) 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 BoronConcentration" The proposed change will increase the value of minimum concentration of solubleboron required in the spent fuel pool from 1800 parts per million (ppm) to 2500ppm. This increase would provide sufficient niegative reactivity to maintain therequired subcriticality margin for a more conservative misloading accident thanpreviously 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 newcoefficients for calculating the minimum required fuel assembly burnup as afunction of decay time and enrichment, specifically for fuel not operated in PINGPoperating Cycles 1 through 4. The revised coefficients result in burnup values thatare 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 analysismethodology.

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 acertain bias for minor actinide and fission product nuclides.

Herein, NSPMrequests 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 thespent fuel criticality analysis remain valid for the combination of boron andgadolinia.

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 tothis combination of neutron absorbers, the proposed amendment does not involvePage 2 of 15 L-PI-1 5-087 NSPMEnclosure 1any 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 BasisIn addition to the specific changes to TS and analysis methodology discussed above, two conservative changes are introduced to the licensing basis as inputs tothe models used in the spent fuel criticality analysis (SFCA). These changes werediscussed 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 spentfuel criticality analysis conservatively ignores the effect of the current neutronabsorber (gadolinia) because it is a net poison throughout the operating cycle.However, this effect is not valid for the new proposed neutron absorber whichis boron, in the form of zirconium diboride IFBA. Therefore, the licensee hasconservatively included the IFBA neutron absorber in the depletion models (asit 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 alsoinvolves the analysis of a new accident that extends beyond the DoubleContingency Principle (the regulatory basis for nuclear fuel storage criticality analyses that states two unlikely independent and concurrent incidents orpostulated accidents are beyond the scope and need not be analyzed).

Whereas the current licensing basis limits the misloading accident to just asingle fuel assembly, the proposed amendment would conservatively adopt amultiple-misloading event in lieu of attempting to justify the low probability ofsuch an event. In effect, the proposed criticality analysis (provided inEnclosure

4) analyzes a conservative array of fuel that bounds any possiblecombination 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 onemodular storage rack design throughout.

As described in PINGP USAR Section10.2.1, the storage rack design originally credited Boraflex neutron absorberpanels between the storage cells to help meet subcriticality criteria.

TheseBoraflex panels are degraded and have not been credited in the current designbasis. The rack design does benefit from a dedicated "flux-trap" design thatprovides a minimum gap between cells. Key design parameters for the storageracks are provided in USAR Section 10.2.1 and Reference 6.1.To ensure stored fuel remains in a subcritical configuration during any normal orPage 3 of 15 L-PI-1 5-087 NSPMEnclosure 1accident condition, strict administrative controls require that any fresh (new) fuelassembly or spent fuel assembly loaded into a storage rack is first evaluated toensure it meets the loading restrictions of TS 3.7.17 and 4.3.1. Currently, eachfuel 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 ofburnup is the average assembly exposure in megawatt days per metric tonuranium (MWD/MTU) and is currently calculated using an industry standard corepower distribution system called BEACONTM (Best Estimate Analyzer for CoreOperations

-Nuclear);

however, other suitable methods have been usedpreviously.

,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 thatit is safely placed in the designated location.

The spent fuel storage racks are designed so that it is impossible to insertassemblies between rack modules or between rack modules and the spent fuelpool wall. Besides the procedural controls on fuel selection and placement inaccordance with allowable storage arrays, criticality of fuel assemblies in a fuelstorage rack is prevented by the design of the rack that limits fuel assemblyinteraction.

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 thesoluble boron present in the spent fuel pool. TS 3.7.16 presently requires aminimum soluble boron concentration of 1800 ppm whenever fuel is present in thespent fuel pool. This boron concentration provides significant margin above thecurrent value (359 ppm) required to maintain an effective neutron multiplication factor (keff) < 0.95 under normal conditions.

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

Additionally, plant design features and operator responsiveness ensure that thecredible spent fuel pool dilution event (initiated at the TS minimum concentration of1800 ppm) will be terminated before the Spent Fuel Pool (SFP) boronconcentration reaches 750 ppm. This termination point provides ample margin tothe current boron concentration (359 ppm) that ensures the limiting normalconfiguration stays below keff 0.95.Fuel designs employed at PINGP are described in USAR Section 3.1. The originaldesign was Westinghouse 14x14 Standard, and the most recent design in use isthe Westinghouse 422 Vantage+

(422V+).

However, several variations of 14x14fuel have been used, including several Exxon designs.

In addition to fuel designchanges, several core design and operational changes have been made over thePage 4 of 15 L-PI-1 5-087 NSPMEnclosure 1plant's operating history that would have a bearing how the nuclear fuel isdepleted during operation.

For instance, Burnable Poison Rods (BPRs) wereinserted into certain unrodded assembly positions for several cycles as a fixedburnable 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 fuelconsolidation campaign that was conducted in 1987. This consolidation projectinvolved removing the fuel rods from two fuel assemblies and reconfiguring theminto a close-packed triangular array; packaged into a specially-design canister.

Inthis manner, 36 assemblies were consolidated into 18 canisters.

The project isfurther described in USAR Section 10.2.1.5.

Consolidated fuel assemblies and other variations on fuel design (failed fuelbaskets) and other spent fuel pool materials of interest (e.g., assembly structural materials from the fuel consolidation project) are described further in the SEGAand supporting calculations.

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

3.2 Current Licensing BasisAt a regulatory level, 10 CFR 50.68(a) requires licensees to select one of twooptions to satisfy criticality accident requirements:

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

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

The applicable criticality criteria for the spent fuelstorage 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 foruncertainties as described in USAR Section 10.2;C. keff < 0.95 if fully flooded with water borated to 400 ppm, which includes anallowance for uncertainties as described in USAR Section 10.2;d. A nominal 9.5 inch~ center to center distance between fuel assemblies placed inthe fuel storage racks; ande. 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, thecurrent licensing basis uses soluble boron credit and applies the doublecontingency principle to demonstrate a keff < 0.95 for all postulated scenarios.

Thiscriterion is described in USAR Section 10.2.1. This keff < 0.95 criterion foraccidents 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 NSPMEnclosure 1The USAR describes the applicable PINGP General Design Criterion (GDC-66) asfollows:

Criticality in new and spent fuel storage shall be prevented by physicalsystems or processes.

Such means as geometrically safe configurations shall beemphasized over procedural controls.

The design and analytical approach tosatisfying this criterion is described in USAR Section 10.2.1.The Prairie Island spent fuel racks have been analyzed to allow storage of fuelassemblies 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, initialenrichment, burnup, and decay time. The analysis does not take any credit for thepresence of the spent fuel rack Boraflex neutron absorber panels which arebelieved 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 Changes3.3.1 Justification for Technical Specification ChangesIn a broad sense, the proposed revisions to TS 3.7.16 (SFP minimum boronconcentration) and to TS Table 4.3.1-3 (coefficients to calculate theminimum required fuel assembly burnup) are justified because the newvalues 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 inmore detail below, these revised TS values can be implemented with littleor no change to existing fuel selection and SFP loading procedures.

Therefore, no new human factors considerations are created by theproposed changes.With respect to TS 3.7.16, the increase of SFP minimum soluble boronconcentration from 1800 ppm to 2500 ppm is justified because:a. The use of SFP soluble boron to accommodate accidents is justified bythe regulation 10 CFR 50.68(b)(4) as well as the current licensing basis which now demonstrates that a soluble boron concentration of1800 ppm accommodates the limiting non-dilution accident (singleassembly misloading accident).

b. Notwithstanding the Double Contingency Principle, extending thelicensing basis to include multiple-assembly misloading accidents is aconservative accommodation for an event that may be considered difficult to preclude considering industry operating experience and thePage 6 of 15 L-PI-1 5-087 NSPMEnclosure Ifundamental reliance on procedural controls to ensure properplacement of fuel assemblies in the PINGP SFP. NSPM has adoptedthis change to the misloading analysis (and the accompanying increase in SFP minimum boron concentration limit) because itreduces the effect of human performance errors that might contribute to a misloading event.c. The new soluble boron limit was established to provide margin abovethe 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 thelimiting multiple-assembly misload used previously-approved analytical methodologies with appropriate input and model changes toincorporate the IFBA-Gd fuel designs.

As discussed in Enclosure 6,the analytical methodologies were sufficiently benchmarked to supportanalysis of gadolin ia-based neutron absorbers in proximity with boron-based neutron absorbers.

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

e. Increasing the minimum TS concentration from 1800 to 2500 ppm willeffectively increase operational margin for mitigating a boron dilutionaccident

, which is analyzed from a starting point of 1800 ppm to anend point of 750 ppm. Enforcing a TS minimum of 2500 ppm willprovide plant operators additional time to identify and mitigate a borondilution event.See Attachment I of this Enclosure for more explanation of the SFP solubleboron 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 forcalculating the minimum required fuel assembly burnup are justified because:a. The use of coefficients for calculating the minimum required fuelassembly burnup has been previously approved and implemented atPINGP. A change to the coefficient values does not constitute a newprocess 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 boronrequirements.

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

Refer toAttachment I of this Enclosure to see how the boron dilution event relates to the current and proposedSFP soluble boron requirements.

Page 7 of 15 L-PI-1 5-087 NSPMEnclosure 1Thus, the revised coefficient values do not require any new humanfactors considerations.

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

c. The new coefficients were calculated using previously-approved analytical methodologies with appropriate input and model changes toincorporate the IFBA-Gd fuel designs.

As discussed in Enclosure 6,the analytical methodologies were sufficiently benchmarked to supportanalysis of gadolinia-based neutron absorbers in proximity with boron-based neutron absorbers.

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

Such achange will not significantly affect the current spent fuel poolconfiguration.

Based on a preliminary

estimate, few spent fuelassemblies would have to be re-assigned to a more-reactive fuelcategory and relocated in the spent fuel pooi to align with the revisedcoefficients.

e. For Fuel Not Operated in Cycles 1-4, the revised coefficients andreanalysis 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 FuelOperated in Cycles 1-4 (which is unaffected by IFBA) sustain thelimiting soluble boron condition of 359 ppm. Refer to See Attachment 1 of this Enclosure for more explanation of the SFP soluble boronconcentrations required for the proposed condition, and the available margins.

Attachment 1 also includes a comparison to the currentcondition.

3.3.2 Justification for Spent Fuel Criticality Analysis Methodology ChangesThe adoption of a certain bias for minor actinide and fission product worth isconsistent 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 thisbiased 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 forPINGP, Westinghouse Electric

Company, LLC (WEC) reviewed theapplicability of the neutronic code suite (ALPHA / PHOENIX-P orPARAGON / ANC) for determining axial power shapes and burnup profilesPage 8 of 15 L-PI-1 5-087 NSPMEnclosure 1of this configuration and concluded that the currently-approved analytical methods are valid for the intended application proposed herein (i.e., IFBA incombination with gadolinia fuel rods). This evaluation is provided inEnclosure 6.3.3.3 Justification for Other Changes to the Current Licensing BasisThe current licensing basis spent fuel criticality analysis conservatively ignores the effect of the current neutron absorber (gadolinia) because it is anet poison throughout its exposure over the operating cycle. However, thenature of the new proposed neutron absorber (boron, in the form ofzirconium diboride IFBA) depletes differently such that it cannot always beviewed as a net poison throughout the operating cycle. Thus, IFBA isexplicitly modeled in the approved computer codes that analyze the nuclearfuel as it depletes in the reactor, and it is conservatively ignored as aneutron absorber in the computer models that analyze criticality of fuel inthe SFP storage configurations.

3.4 Conclusion

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

None of thechanges affect a system, structure, or component, and none result in a change tohow systems are operated.

in that regard, the proposed changes do not create anew 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) ofthe PINGP on September 28, 1972. The SE, Section 3.1, "Conformance with AECGeneral 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 theAEC'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 therevised General Design Criteria in February 1971 and the present version of thecriteria in July 1971. We did not require the applicant to reanalyze the plant orresubmit the FSAR. Howe ver, our technical review did assess the plant againstthe General Design Criteria now in effect and we are satisfied that the plant designgenerally conforms to the intent of these criteria.

Page 9 of 15 L-PI-1 5-087 NSPMEnclosure 1Based on the above, the applicable PINGP GDC states: Criticality in spent fuelstorage shall be prevented by physical systems or processes.

Such means asgeometrically 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 updatedreview guidance to the NRC staff to address the increased complexity of recentSFP nuclear criticality analyses and operations.

The ISG rebaselines NRC'sexpectations for spent fuel criticality analysis.

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

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

The requirements for systemoperability during movement of irradiated fuel are included in the TSs inaccordance with 10 CFR 50.36(c)(2),

Limiting Conditions for Operation.

Asrequired by 10 CFR 50.36(c)(4),

design features to be included are those featuresof 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 thehandling and storage at any one time of more fuel assemblies than have beendetermined 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, thek-effective of the spent fuel storage racks loaded with fuel of the maximum fuelassembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percentconfidence level, if flooded with borated water, and the k-effective must remainbelow 1.0 (subcritical),

at a 95 percent probability, 95 percent confidence level, ifflooded with unborated water."The U.S. Atomic Energy Commission (AEC) issued its Safety Evaluation (SE) forPINGP before the revised General Design Criteria (GDCs) were published in 1971.A PINGP GDC requires that, "Criticality in new and spent fuel storage shall beprevented by physical systems or processes.

Such means as geometrically safeconfigurations shall be emphasized over procedural controls."

As guidance for reviewing criticality analyses of fuel storage at light-water reactorpower plants, the NRC staff issued an internal memorandum on August 19, 1998(ADAMS Accession No. ML00372B001).

This memorandum is known as thePage 10 of 15 L-Pl-1 5-087 NSPMEnclosure 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 theReview of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Edition,"

particularly Section 9.1.1, "Criticality Safety of Fresh and SpentFuel Storage and Handling,"

Revision 3, issued March 2007. Section 9.1.1provides the existing recommendations for performing the review of the nuclearcriticality safety analysis of SFPs.4.2 Precedent There is little precedent that is applicable to the proposed activity because of thefollowing factors:a. Based on the recent review and approval of PINGP Unit 1 and 2 licenseamendments (in 2013 per Reference 6.4) that explicitly addressed the citedInterim Staff Guidance and contemporaneous precedent, there has been littleopportunity for new developments.

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

Accordingly, this precedent was addressed in Enclosure 4 with due consideration and analysis of a multiplefuel assembly misload event in the PINGP spent fuel criticality analysis.

The Comanche Peak amendment also set precedent for adopting a certain bias forminor actinide and fission product nuclides.

This precedent is addressed inEnclosure 4.4.3 Significant Hazards Consideration Northern States Power Company, a Minnesota Corporation (NSPM), doingbusiness as Xcel Energy, proposes to amend the renewed operating licenses ofPrairie Island Nuclear Generating Plants (PINGP) Units 1 and 2. The purpose ofthis amendment is to modify the PINGP Technical Specifications (TS) to allowspent fuel pooi (SFP) storage of nuclear fuel containing a boron-based neutronabsorber in the form of zirconium diboride Integral Fuel Burnable Absorber (IFBA).Page 11 of 15 L-PI-1 5-087 NSPMEnclosure 1The proposed revisions involve an incremental increase to the minimum requiredvalue for Spent Fuel Pool (SFP) boron concentration and incremental change tothe coefficients used to calculate the minimum required fuel assembly burnup forestablishing fuel storage categories for safe loading patterns.

These revised TSvalues can be implemented with minimal change to existing fuel selection and SEPloading 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 involvedwith the proposed changes by focusing on the three standards set forth in 10 CFR50.92(c) as discussed below:1. Does the proposed change involve a significant increase in theprobability or consequences of an accident previously evaluated?

Response:

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

(1) fuel assembly misloading, (2) loss of spent fuel poolcooling, 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 continueto be controlled by approved fuel selection and fuel handling procedures.

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

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

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

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

Response:

No.The proposed changes involve incremental changes to TS values, andrepresent minimal change to existing fuel selection and SEP loadingprocedures.

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

Theproposed changes do not involve new SFP loading configurations and do notchange or modify the fuel, fuel handling processes, fuel storage racks, numberof 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 proposedchanges introduce no new material interactions, man-machine interfaces, orprocesses 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 ofsafety?Response:

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

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

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

4.4 Conclusions

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

A proposed amendment of an operating license for afacility requires no environmental assessment if the operation of the facility inaccordance 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 proposedamendment meets the eligibility criteria for categorical exclusion set forth in 10CFR 51 .22(c)(9).

Pursuant to 10 CFR 51 .22(b), no environmental impactstatement or environmental assessment needs to be prepared in connection withthe 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 typesor increase in the amounts of any effluents that may be released offsite.Implementation of the proposed changes involves no physical change to thenuclear fuel or the types of exposure it would receive.

Nor does it involve thephysical 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 theproposed 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 willnot result in any unusual spent fuel pool operations that would result in aPage 14 of 15 L-PI-1 5-087 NSPMEnclosure 1permanent effect to increase occupational exposure.

The proposed fuelstorage configurations do not fundamentally change the inventory orradiological source term of the spent fuel. In addition, based on NSPM'sexperience with routine fuel movement campaigns during refueling outages,the cumulative exposure from the proposed activities is expected to beminimal.

6.0 REFERENCES

6.1 Westinghouse Report WCAP-1 7400-P, Prairie Island Units 1 and 2 Spent FuelPool Criticality Safety Analysis, Revision 0, dated July 2011 (submitted asEnclosure to Xcel Energy Letter to NRC dated August 19, 2011 (ADAMSAccession No. MLl12360231) 6.2 Interim Staff Guidance DSS-ISG-2010-01, Staff Guidance Regarding the NuclearCriticality 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 forFuel 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 andNRC Safety Evaluation Report (SER) dated August 29, 20136.5 Comanche Peak Units 1 and 2 Operating License Amendment No. 162 and NRCSER dated July 1, 2014 (ADAMS Accession No. ML14160A035) 6.6 NRC (Terry Beltz) letter to Xcel Energy, "Summary of the April 14, 2015, PublicMeeting with Xcel Energy and Westinghouse to Discuss a. Potential Future LicenseAmendment Request Regarding the Use of Integral Fuel Burnable AbsorberNeutron 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 5-087Enclosure 1, Attachment I, Comparison of SFP Boron Requirements NSPMPurpose:

This attachment describes how the revised Spent Fuel Pool (SFP) solubleboron requirements and the revised TS 3.7.16 limit for minimum SFP boronconcentration affect the margins to limiting conditions in the SFP. Please refer to thegraphic below (Figure A-I) and note that "Current Licensing Basis" relates to the currentconditions, and "Proposed Licensing Basis" relates to the conditions proposed in thelicense amendment request.Figure A-IComparison of SFP Boron Requirements (Current vs. Proposed)

Current Licensing Basis ISFP Boron Concentration (ppm) r , ...2500 TS Minimum (2500)Licensing Basis2400 -42300 --2200 -"TS Minimum (1800)---Start 2100 -2000 -1900 -1700-15600-E- Minimum for Non_-Dilution Accidents (2030)(Analytical value for multiole assembly misload)Start1300--Minimum for No._n-Dilution Accidents (910) (analytical value for assembly misload)Stop! 0-700 ---m.omStop600 ---500 --Minimum "IS Value for Normal Keff 0.95 (400) ->4 00 Minimum TS Value (400) for Normal Keff 0.95Limiting Configuration Normal Keff 0.95 (359) 300 -LitngCfgutonNrlKe 09(3)200 --100 --0-Page 1 of 2 L-PI-1 5-087 NSPMEnclosure 1, Attachment 1, Comparison of SEP Boron Requirements

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

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

The difference between 400 ppm and the SBC at the limiting normal condition in the analysisprovides administrative margin to accommodate a future analysis error. The LARproposes no change to the TS 4.3.1.1 .c value and no change to the value thatachieves the limiting SBC for the normal condition.

Therefore, the margin isunchanged by the proposed amendment.

SBC mar qin for Boron Dilution Event mitigation.

As shown in Figure A-i, the BoronDilution Event has not been reanalyzed for the proposed amendment; the event isstill postulated to start at 1800 ppm and progress to 750 ppm. Within this SBCrange in the dilution

analysis, the analysis shows that the time of dilution providesample time for operators to identify and mitigate the event. Thus, any margin above1800 ppm provides incrementally more time for operators to respond to the dilutionevent. Thus, the proposed TS 3.7.16 change to a minimum SBC of 2500 ppmincreases 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 thelimiting misloading.

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

Nevertheless, anominal 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 analysisof the multiple misloading requires a much higher SBC of 2030 ppm (to maintain akeff < 0.95), which appears to significantly reduce the margin to the TS minimumSBC as follows:* In the current condition, the margin from TS 3.7.16 minimum SBC to theminimum required for the misloading event is 890 ppm (1800 ppm minus 910ppm).* In the proposed condition, the margin is reduced to 470 ppm (2500 ppm minus2030 ppm).This reduction in margin should not be concerning because:

(1) the DoubleContingency Principle neutralizes the effect (as discussed above), and (2) thereduction in margin resulted from the conservative adoption of a multiple-misloading event that could have been previously included in the licensing basis. There isnothing inherent to the proposed use of IFBA that would increase the probability ofmisloading accidents.

Page 2 of 2 L-PI-1 5-087 NSPMEnclosure 2Enclosure 2Marked-Up Technical Specification Pages3 pages follow3.7.16-14.0-7Insert Spent Fuel Storage Pool Boron Concentration 3.7.163.7 PLANT SYSTEMS3.7.16 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.ACTIONSCONDITION REQUIRED ACTION COMPLETION TIMEA. 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 fuelstorage pool.ANDA.2 Initiate action to restore spent Immediately fuel storage pool boronconcentration to within limit.Prairie IslandUnits 1 and 2Unit 1 -Amendment No. 4-48Unit 2 -Amendment No. 4-493.7.16-1 Replce al deetedvalus lDesign Featureswithvales n th atachd ~Table 4.3.1-3 (page 1 of 1)For Fuel Not Operated In Units 1 and 2 Cycles 1 -4Coefficients to Calc late the Minimum Required Fuel Assembly Burnup (Bu) as aFu) ction of Decay Time and Enrichment (En)FUEL DECAY TIME \COEFFICIENTS lA 3A2 0 -066 9.4 -3.80 -0,42,0 4-30 320 -0,404-$

2,4 4. -4..4615 .4.81 2.4 4.620 .441 4.3 .48440 4-3 4486 510 -2.4 2484-3.015 443 4404.2420 -042 2.2 2.44-44Notes:1.All relevant uncertainties are explicitly included in the criticality analysis.

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

For a fuel assembly to meet the requirements of a Fuel Category, theassembly 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 *En3 + A2*En2 + A3*En + A42. Initial enrichment (En) is the nominal U-235 enrichment.

Any enrichment between 1.7 and 5.0 weight percent U-235 maybe 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 ispermitted.

4. If Decay Time value fails between increments of the table, the lower Decay Time value shall be used or a linearinterpolation may be performed as follows:

Compute the Bu value using the coefficients associated with the Decay Timevalues that bracket the actual Decay Time. Interpolate between Bu values based on the increment of Decay Time betweenthe 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 IslandUnits 1 and 2Unit 1 -Amendment No. 20-Unit 2 -Amendment No. 944.0-7 Insert to Table 4.3.1-3Replace the Coefficient (A1, A2, A3, A4) values with those shown belowFuelCoefficients CategoryDecay TimeA1A2A3A42 0 -1.9089 22.9292 -81.9646 91.4193o -0.0536 0.5516 8.2824 -23.31575 -0.0372 0.2803 9.0736 -23.8543310 -0.0408 0.2587 9.0667 -23.645215 -0.0893 0.7485 7.2536 -21.410220 -0.1011 0.8822 6.6122 -20.44684 0 1.3659 -14.9709 63.0347 -72.92230 0.2744 -3.7275 29.5218 -41.71745 0.0533 -1 .3478 20.6704 -32.32355 10 -0.0407 -0.3472 16.7092 -27.959115 -0.1809 1.0636 11.8632 -23.047620 -0.0897 0.2312 13.9007 -24.55290 0.4604 -5.9192 38.3216 -50.30215 0.4161 -5.2825 34.6238 -45.63816 10 0.3716 -4.7154 31.7812 -42.226015 0.1816 -2.7038 24.7285 -35.1164______ 20 0.1318 -2.1711 22.5833 -32.7644 L-PI-1 5-087 NSPMEnclosure 3Enclosure 3Marked-Up Technical Specification Bases Pages8 pages followB3.7.1 6-2B3.7.1 6-3B3.7.1 6-4B3.7.1 6-5B3.7.1 7-2B3.7.1 7-5B3.7.1 7-6B3.7.1 7-10Insert Fuel Storage Pool Boron Concentration B 3.7.16BASES (continued)

-andRef 5IAPPLICABLE SAFEITYANALYSESThe spent fuel pool criticality analysis (Ref. 4) addresses all thefuel types currently stored in the spent fuel pool and in use inin the reactor.

The fuel types considered in the analysis include theWestinghouse Standard (STD), OFA, and Vantage Plus designs,(both 0.400" and 0.422" O.D. designs) and the Exxon fuel assemblytypes in storage in the spent fuel pool.~Accident conditions which could increase the keff were evaluated Dropped and. icung..misplaced fresh incluing:locations; andIfuel 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 rackd. Tntramodule w..ate=r gap reduc,,tio-n to. a o, .ei ...c evn; and@. Spent fuel pool temperature greater than 150 °F.JINew fueFor an occurrence of these postulated accident conditions, the doublecontingency 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, forthese postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 359 ppmrequired to maintain 1kf less than 0.95 under normal conditions) canbe assumed as a realistic initial condition since not assuming itslanalytically presence would be a second unlikely event.--,,L_./-lan Ref5/Calculations were performed (Ref. 4) to determine the amount ofpraph [soluble boron required to offset the highest reactivity increasethe lastI caused by these postulated accidents and to maintain k~f less thanor equal to 0.95. It was found that a spent fuel pool boroncocnrtono

-- 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 to0.95. This specification ensures the spent fuel pool containsInsert paragprovided onpage in thisEnclosure Prairie IslandUnits 1 and 2Unit Reviion IUnit 2 -Revision 2,-1B 3.7.16-2 Fuel Storage Pool Boron Concentration B 3.7.16BASESAPPLICABLE SAFETYAMNA 1 Vq1l'Zadequate dissolved boron to compensate for the increased reactivity caused by these accidents.

(continued) 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:eTo 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 lowboron-10 concentration of 19.4 atom percent) to achieve keff less than or equal to0.95. The TS 3.7.16 limit of 2500 ppm ensures the spent fuel pool containsadequate dissolved boron to compensate for the increased reactivity caused by thisaccident.

Prairie IslandUnits 1 and 2Unit 1 -Revision

!2-1-Unit 2 -Revision 2241-B 3.7.16-3 Fuel Storage Pool Boron Concentration B 3.7.16LCOThe 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 _ithe potential critical accident scenarios as described in Reference 4.'This concentration of dissolved boron is the minimum requiredconcentration for fuel assembly storage and movement within thefuel storage pool.APPLICABIITUY This LCO applies whenever fuel assemblies are stored in the spentfuel storage pool.ACTIONS A. 1 and A.2The Required Actions are modified by a Note indicating thatLCO 3.0.3 does not apply.When the concentration of boron in the spent fuel storage pool isless than required, immediate action must be taken to preclude theoccurrence of an accident or to mitigate the consequences of anaccident in progress.

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

Theconcentration of boron is restored simultaneously with suspending movement of fuel assemblies.

This does not preclude movement ofa fuel assembly to a safe position.

If the LCO is not met while moving irradiated fuel assemblies inMODE 5 or 6, LCO 3.0.3 would not be applicable.

If movingirradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuelmovement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason torequire 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 3.7.16BASES (continued)

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

The 7 day Frequency isappropriate because no major replenishment of pool water isexpected 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 ReactorLicensees from B. k. Grimes, "OT Position for Review andAcceptance 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 AnalysisISupplement Analysis for the Storage of IFBA Bearing Fuel, WCAP-17400-P, ISupplement 1, Rev 1, Westinghouse Electric

Company, October 2015.Prairie IslandUnits 1 and 2Unit 1 -Revision 2,2--Unit 2 -Revision

!2-4,IB 3.7.16-5 Spent Fuel Pool StorageB 3.7.17BASES (continued)

APPLICABLE SAFEIYANALYSESPer Reference 5, thepresence of anIntegral FuelBurnable Absorber(IFBA) is considered for the 422V+ fueldesign.The hypothetical criticality accidents can only take place duringor as a result of the movement of an assembly (Ref. 4 and 5). Forthese accident occurrences, the presence of soluble boron in thespent fuel storage pool (controlled by LCO 3.7.16, "Fuel StoragePool Boron Concentration")

prevents criticality.

By closelycontrolling the movement of each assembly and by verifying theappropriate checkerboarding after each fuel handling

campaign, thetime period for potential accidents may be limited to a small fractionof the total operating time. During the remaining time period withno potential for criticality accidents, the operation may be under theauspices of the accompanying LCO. ir-and 5The spent fuel storage racks have been an, zed in accordance withthe methodology contained in Reference

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

The codes, methods andtechniques contained in the methodology are used to satisfy thesecriteria 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 percentU-235 while maintaining lYf < 1.0 (including uncertainties) ifflooded 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 acon v'eatively low boron-l0 atom percent of 19.4) will ensure anyloaded c figuration k~f will be < 0.95. In addition, the analysisdifferentiate~a f uel assembly operated during Operating Cycle 1 -4from an nassembl..rated after Cycle 4 in determining theassembly's reactivity.

Credit is taken for the radioactive decay timeof the spent fuel. No credit is given for any gadoliniurrburnable poison in the fuel. [rIB /The criticality analysis (Ref. 4 ,specifically analyzed each of thefollowing storage to ensure that the spent fuel poolwill remain subcritical when fud is placed in accordance withSpecification 4.3.1.1.an5 Prairie IslandUnits 1 and 2Unit 1 -Revision

22-1-Unit 2 -Revision 2,2-1-.B 3.7.17-2 Spent Fuel Pool StorageB 3.7.17BASESAPPIJCABLE SA-ETYANALYSES(continued) modules because all the racks in the SFP have identical fuelcell design and the actual physical gap between rack modules isignored in the analysis (i.e., there is no credit taken for the gapsbetween rack modules).

Array interface requirements:

Technical Specifications provideonly 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 notprovide any unique rules for the interface between arrays.Rather, the Technical Specifications require that all fuel in thespent fuel pool satisfy one of the required arrays, even intransitions between two major arrays.Specification 3.7.17 and Specification 4.3 ensure that fuel is storedin 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 typescurrently stored in the spent fuel pool and in use in the reactor.

Thefuel types considered in the analysis include the Westinghouse Standard (STD), OFA, and Vantage Plus designs (both 0.400" and0.422" O.D. designs),

and the Exxon fuel assembly types in storagein the spent fuel pool.Accident conditions which could increase the keff were evaluated Dropped and misplaced freshfuel assembly

a. A new fuel assembly drop on the top of the racks;Inadertet cb. A new fuel assembly misloaded bremoval of an c.A new fuel assembly, misloaded iiRCCA!oain

etween rack modules:nto an-incorrect storage rackNew fuel assemblies

! !Prairie IslandUnits 1 and 2Unit 1 -Revision

2,2-1-Unit 2 -Revision 22-1-B 3.7.17-5 Spent Fuel Pool StorageB 3.7.17BASESAPPLICABLE

d. ntramodu..

water...

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

For an occurrence of these postulated accident conditions, the doublecontingency principle of Reference 2 can be applied.

This states thatone is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.

Thus, forthese postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 359 ppmrequired to maintain kef less than 0.95 under normal conditions) canbe assumed as a realistic initial condition since not assuming itspresence would be a second unlikely event.Iand Ref 5 lf Westinghouse Electric Company LLC calculations (Ref. 4) were ..2030Insert paragraph performed to determine the amount of soluble boron required4 oprovided on last offset the highest reactivity increase caused by these po s~atedpage in this [accidents and to maintain keff less than or equal t .7. It was foundEnclosure Ithat a spent fuel pool boron concentration of 94 ppm (assuming aconservatively low boron-lO atom percent of 19.4) was adequate tomitigate these postulated criticality related accidents and to maintainlqff less than or equal to 0.95.Specification 3.7.16 ensures the spent fuel pool contains adequatedissolved boron to compensate for the increased reactivity caused bya-mispositioned fuel aemxyor aloss of spent fuel pool cooling.Specification 4.3 requires that the spent fuel rack keff be less than orequal to 0.95 when flooded with water borated to 400 ppm. Thisvalue was selected to provide a nominal margin above the calculated limiting value of 359 ppm. A spent fuel pool boron dilution analysiswas performed which confirmed that sufficient time is available todetect and mitigate a dilution of the spent fuel pool before the 0.95keff design basis is exceeded.

The spent fuel pool boron dilutionanalysis concluded that sufficient time would be available foroperators to recognize and terminate a dilution event that started atPrairie Island Unit 1 -Revision 224-1Units 1 and 2 B 3.7.17-6 Unit 2 -Revision 22.1-Spent Fuel Pool StorageB 3.7.17BASESREFERENCES (continued)

4. "Prairie Island Units 1 2 et Fuel Pool Criticality Analysis",

WCAP-1740

-NP, R lision 0, Westinghouse Electric

Company, July5. Not 6. A Nuclear Society, "American National StandardDesig 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 AnalysisSupplement Analysis for the Storage of IFBA Bearing Fuel," WCAP-1 7400-P, Supplement 1, Rev 1, Westinghouse Electric

Company, October 2015.Prairie IslandUnits 1 and 2Unit 1 -Revision

!2-1Unit 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 beencoincidentally misloaded in spent fuel pools, the PING P licensing basis criticality analysis has adopted the possibility of a multiple-assembly misloading accident anddetermined that subcriticality requirements can be met with a concentration of solubleboron 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 anunlikely event that need not be assumed to occur coincidentally with another unlikely, independent event (such as a dilution event).