tl Xcel Energy®

,vIAY 23 2013

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

L-PI-13-030 10 CFR 50.90

License Amendment Request (LAR) to Add a Methodology to Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)"

Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), hereby requests an amendment to the TS for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, to add a methodology to TS 5.6.5. NSPM evaluated the proposed changes in accordance with 10 CFR 50.92 and concluded that they involve no significant hazards consideration.

The enclosure to this letter, "Evaluation of the Proposed Changes", contains the licensee's evaluation of the proposed changes.

NSPM requests approval of this LAR within one calendar year of the submittal date. Upon NRC approval, NSPM requests 90 days to implement the associated changes. In accordance with 10 CFR 50.91, NSPM is notifying the State of Minnesota of this LAR by transmitting a copy of this letter and enclosure to the designated State Official.

If there are any questions or if additional information is needed, please contact Mr. Dale Vincent, P.E., at 651-388-1121.

1717 Wakonade Drive East • Welch, Minnesota 55089-9642 Telephone: 651.388.1121

Document Control Desk Page 2

Summary of Commitments

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

I declare under p7n~lty of perjury that the foregoing is true and correct. Executed on 5; r23j)S

d!~ Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota

Enclosures (1)

cc: Administrator, Region III, USNRC Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC State of Minnesota

ENCLOSURE

Evaluation of the Proposed Changes

License Amendment Request (LAR) to Add a Methodology to Technical Specification (TS) 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)"

1. SUMMARY DESCRIPTION 2. DETAILED DESCRIPTION

2.1 Proposed Changes 2.2 Background

3. TECHNICAL EVALUATION 4. REGULATORY SAFETY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5. ENVIRONMENTAL CONSIDERATION 6. REFERENCES

ATTACHMENTS:

1. Technical Specification Pages (Markup) 2. Technical Specification Pages (Retyped)

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Enclosure Revise TS 5.6.5

1. SUMMARY DESCRIPTION

NSPM

This evaluation supports a request to amend Renewed Operating Licenses DPR-42 and DRP-60 for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, respectively.

Pursuant to 10 CFR 50.90, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), hereby requests an amendment to the TS for PINGP, Units 1 and 2, to add a methodology to TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)".

2. DETAILED DESCRIPTION

2.1 Proposed Changes

A brief description of the associated proposed TS changes is provided below along with a discussion of the justification for each change. The specific wording changes to the TS are provided in Attachments 1 and 2 to this enclosure.

TS 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)": This LAR proposes to add new references 32 and 33 to the list of NRC-approved topical reports for calculating core operating limits. These changes are acceptable because these references embody an NRC-approved methodology which applies to the reactor core design for pressurized water reactors (PWRs) which utilize uranium fuel. PINGP is a PWR with uranium fuel.

There are no Bases changes associated with this LAR.

In summary these changes are acceptable because they assure the core operating limits have been calculated in accordance with an NRC approved methodology. Use of this methodology will improve plant safety through more accurate core design capability.

2.2 Background

The reactor core for each PINGP unit operating cycle is designed in accordance with NRC-approved methodologies and evaluated pursuant to 10 CFR 50.59. Under the guidance of GL 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications" (ML031150407) and NUREG-1431, Rev. 2.2, Standard Technical Specifications Westinghouse Plants", (ML031270565) the topical reports for these methodologies are listed in TS 5.6.5.b. Westinghouse has developed improved methods for use of the Westinghouse Advance Nodal Computer Code (ANC) which the NRC has reviewed and approved in WCAP-16045-P-A, "Qualification of the Two-

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Dimensional Transport Code PARAGON" and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology". NSPM proposes to add these references to TS 5.6.5.b and use this methodology in the reactor core design at PINGP.

NSPM requests approval of this LAR within one year. If this LAR is approved in the requested timeframe, NSPM will apply this methodology to the Unit 1 reactor core designated as 1 R29 which commences in the autumn of 2014, the Unit 2 reactor core designated as 2R29 which commences in the autumn of 2015, and subsequent reactor core designs.

With the TS changes proposed in this LAR the plant will continue to operate safely, and the health and welfare of the public will continue to be protected.

3. TECHNICAL EVALUATION

PINGP is a two unit plant located on the right bank of the Mississippi River approximately six miles northwest of the city of Red Wing, Minnesota. The facility is owned and operated by Northern States Power Company, a Minnesota corporation (NSPM). Each unit at PINGP employs a two-loop pressurized water reactor designed and supplied by Westinghouse Electric Corporation. The initial PINGP application for a Construction Permit and Operating License was submitted to the Atomic Energy Commission (AEC) in April 1967. The Final Safety Analysis Report (FSAR) was submitted for application of an Operating License in January 1971. Unit 1 began commercial operation in December 1973 and Unit 2 began commercial operation in December 1974.

The PINGP was designed and constructed to comply with the licensee's understanding of the intent of the AEC General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. PINGP was not licensed to NUREG-0800, "Standard Review Plan (SRP)."

Core Operating Limits Report (COLR)

Each PINGP reactor core fuel cycle design requires cycle-specific parameter limits to be defined in the plant TS to assure the plant is operated safely. In accordance with the guidance of NUREG-1431, Revision 4, "Standard Technical Specifications, Westinghouse Plants", each applicable PINGP TS references the COLR as containing the applicable limit. TS 5.6.5.a requires, "Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following ... ", after which follows the list of all TS which reference the COLR to specify core operating limits. TS 5.6.5.b states, "The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents ... ", after which follows the list of NRC-approved topical reports that embody the methodologies used to determine the core operating limits. TS 5.6.5.c

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requires, "The core operating limits shall be determined such that all applicable limits .. . of the safety analysis are met." TS 5.6.5.d requires each revision of the COLR to be provided to the NRC.

PINGP license amendments (LA) 162 and 153 (LA-162/153) issued April 28, 2004 (ML040900209), for Units 1 and 2 respectively, adopted the provisions of NUREG-1431, Revision 2, which revised the content requirements for COLR references. NUREG-1431, Revision 2, incorporated Technical Specifications Task Force (TSTF) industry traveler TSTF-363, "Revise Topical Report references in ITS 5.6.5, COLR", which removed the specific date and revision for topical report references and requires the reference details to be included in the COLR for references used to prepare the COLR. NRC letter dated August 4, 2011, (ML 110660285) allows licensees, which adopted this format prior to the date of the letter, to continue to use the TSTF-363 format. The PINGP COLR is consistent with the guidance of NUREG-1431, Revision 2, as allowed by NRC letter dated August 4, 2011 (ML 110660285), and includes the complete identification for each of the TS referenced topical reports used to prepare the COLR, that is, report number, title, revision, date, and any supplements.

Proposed Changes

Currently, PINGP reactor cores are designed using WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores". Westinghouse has developed a revised methodology for reactor core design using the Westinghouse Advanced Nodal Computer Code (ANC) which was reviewed and approved by the NRC in WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology". This LAR requests approval to use this revised methodology in the design of the PINGP Unit 1 and 2 reactor cores in lieu of WCAP-11596-P-A and include this methodology in the PINGP TS by addition of two references to TS 5.6.5.b as follows:

32. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON"; and

33. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology".

Technical Basis for Change

TS 5.6.5.b lists the analytical methods which are used to determine the core operating limits. The proposed TS change will modify TS 5.6.5.b to add references WCAP-16045-P-A and WCAP-16045-P-A, Addendum 1-A. TS Section 5.6.5.b states that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

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WCAP-16045-P-A confirms the qualifications of the PARAGON code both as a standalone transport code and as a substitute for the PHOENIX-P code, the code currently used in PINGP design, as a nuclear data source for nodal codes. As part of the qualification process, the topical report includes a comparison of PARAGON predicted values to measured data from several plants. Benchmarking has shown that results from the PARAGON/ANC code package are essentially the same as those obtained from the current PHOENIX-P/ANC system. WCAP-16045-P-A concludes that the application of PARAGON does not result in any undesirable changes in predicted fuel performance or safety analysis results. The NRC Safety Evaluation (SE) for the PARAGON nuclear data methodology attached to a letter dated March 18, 2004, (ML040780402) states, " ... the staff considers the new PARAGON code to be well qualified as a stand-alone code replacement for the PHOENIX-P lattice code, wherever the PHOENIX-P code is used in NRC-approved methodologies."

WCAP-16045-P-A, Addendum 1-A, verifies the accuracy of NEXUS for cross section representation. As part of this topical report, different assembly types were calculated using NEXUS, which include the following: both Westinghouse and Combustion Engineering assembly types; U02 fuel; and integral fuel burnable absorber (lFBA); wet annular burnable absorber (WABA); and Gd20 3 burnable absorbers. The k-infinity results from these calculations were compared directly to PARAGON k-infinity results at corresponding conditions. The comparisons demonstrated that the NEXUS cross sections are accurate over the range of temperatures, boron concentrations, and power levels expected to be encountered in PWR core calculations. The NRC SE for the NEXUS nuclear data methodology attached to a letter dated February 23, 2007 (ML070320398) states:

The NRC staff has reviewed the TR [Topical Report] submitted by Westinghouse and determined that the NEXUS/ANC code system is adequate to replace the PARAGON/ANC code system wherever the latter is used in NRC-approved methodologies. The NRC staff, furthermore, has determined that NEXUS/ANC is qualified as a stand-alone code system so long as its use is limited by the provisions listed in Section 4.0 of this safety evaluation.

The provisions listed in Section 4.0 state that NEXUS/ANC shall be limited to uranium­fueled PWR applications (that is, not mixed-oxide fuel). As such, NEXUS/ANC is acceptable for use for PINGP.

Later versions of the Westinghouse ANC code require cross section data which is generated using the PARAGON neutron transport code. The NEXUS/ANC system is a version of the PARAGON/ANC system in that all nuclear data is based on PARAGON and only the methods of representing this data in ANC have been changed from the version of PARAGON/ANC described in WCAP-16045-P-A. Therefore, NEXUS methodology will be used as a replacement to the PHOENIX-P methodology when NSPM moves to a later version of ANC.

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N8PM

The other methodologies used to determine operating limits referenced in the COLR remain applicable with the use of PARAGON and NEXUS. Future changes to the values of the operating limits are controlled by the 10 CFR 50.59 process, may only be developed using NRC approved methodologies, and must remain consistent with all applicable plant safety analysis limits addressed in the Updated Safety Analysis Report (USAR). The consequences of the design basis accidents will continue to be calculated using NRC accepted methodologies. Assumptions used in the safety analysis are not changed by the use of PARAGON and NEXUS. Safety analysis acceptance criteria are not being altered by the use of PARAGON and NEXUS.

Specifically, at PINGP, these proposed methodologies will be used to determine the core operating limits specified in TS or confirm that the safety analyses which set the limits in the core operating limits report bound expected cycle operation.

The COLR contains twelve sections which provide the limits associated with each of twelve TS that reference the COLR to provide applicable limits.

The methodologies proposed in this LAR will be used to determine the following core operating limits:

• COLR Section 2 addresses shutdown margins required by TS 3.1.1. The actual limits are provided in COLR Table 1;

• COLR Section 6 addresses shutdown margins required by TS 3.1.8. The actual limits are provided in COLR Table 1;

• COLR Section 9 addresses axial flux difference limits required by TS 3.2.3. The actual limits are provided in COLR Figures 5 and 6;

• COLR Section 12 (a) addresses the maximum core reactivity allowed for the refueling boron concentrations required by TS 3.9.1; and

• COLR Section 12 (c) addresses minimum shutdown margin requirements for the refueling boron concentration required by TS 3.9.1. The shutdown margins used to determine the refueling boron concentrations are provided in COLR Table 1.

The methodologies proposed in this LAR will be used to confirm that the safety analyses, which set the following core operating limits, bound expected cycle operation:

• COLR Section 1 addresses reactor core safety limits required by TS 2.1.1. COLR Figure 1 provides the actual limits;

• COLR Section 3 provides isothermal temperature upper and lower limits required by TS 3.1.3;

• COLR Section 5 addresses control bank insertion limits required by TS 3.1.6. The actual limits are provided in COLR Figures 2,3 and 4;

• COLR Section 7 provides the heat flux hot channel factor limit required by TS 3.2.1;

• COLR Section 8 provides the nuclear enthalpy rise hot channel factor (F ~HN) limit required by TS 3.2.2;

• COLR Section 10 provides parameter values for reactor trip system

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instrumentation as required by TS 3.3.1; and

NSPM

• COLR Section 11 provides reactor coolant system pressure, temperature and flow limits as required by TS 3.4.1.

The methodologies proposed in this LAR will be used to determine the penalty factors to be applied prior to determining if the limit has been met:

• COLR Section 7 addresses heat flux hot channel factor (Fo(Z)) required by TS 3.2.1. Tables 2 through 6 specify the penalty factors applied to the steady state Fo (TS Fe o(Z)) before determining if the limit specified in COLR Section 7 is met. Figures 5 and 6 specify which of these Tables are used in the determination.

The methodologies proposed in this LAR are not used to determine the core operating limits in COLR Section 4 and COLR Section 12 (b).

The proposed change from use of PHOENIX-P/ANC to PARAGON-NEXUS/ANC requires updating from ANC Version 8 to ANC Version 9. Since PINGP TS 5.6.5.b Reference 21 does not specify the applicable ANC version, no change to this reference is required or proposed.

Currently core operating limits provided in the PINGP COLR are based on the methodology in PINGP TS 5.6.5.b Reference 23. This LAR does not propose to delete Reference 23. PINGP is a two unit plant governed by a single set of Technical Specifications. This LAR requests NRC approval of the proposed changes by June 2014 after which the proposed TS changes would be incorporated into the TS and the current methodology and the methodology proposed in this LAR would both be available for use in determining PINGP core operating limits. NSPM plans to apply the new methodology in this LAR to the Unit 1 fuel cycle 1 R29 which commences in the autumn of 2014. At that time PINGP Unit 2 will continue to require use of the current methodology until the next Unit 2 refueling outage in the autumn of 2015, 2R29, after which Unit 2 will apply the methodology proposed in this LAR. Since there is an unavoidable transition period when the current and proposed methodologies are both required to support core operating limits, NSPM does not propose in this LAR to remove Reference 23. Removal of Reference 23 would require another LAR after the transition period which would not serve a safety function. The NRC and NSPM resources required to process an LAR to remove Reference 23 would better serve plant safety applied to other issues.

Conclusions

This LAR requests approval to use a new methodology with the Westinghouse Advanced Nodal Computer Code (ANC) as described in topical reports WCAP-16045-P-A and WCAP-16045-P-A, Addendum 1-A, and proposes to add these topical reports as References 32 and 33 in PINGP TS 5.6.5.b. The NRC has reviewed and approved this new methodology for design of pressurized water reactor cores with uranium fuel, and since PINGP is a pressurized water reactor which utilizes uranium fuel, this

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methodology is applicable to PINGP. With the addition of these references to PINGP TS 5.6.5.b, core operating limits will be determined using this methodology. Operation of the Prairie Island Nuclear Generating Plant with the proposed TS revisions will continue to protect the health and safety of the public.

4. REGULATORY SAFETY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria

Title 10 Code of Federal Regulations 50.36, "Technical specifications":

(c) Technical specifications will include items in the following categories:

(5) Administrative controls. Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.

This license amendment request proposes to add two topical report references to the COLR list of NRC-approved methodologies in the Technical Specification Administrative Controls, Report Requirements Section 5.6. These methodologies will be used to determine core operating limits which assure the plant is operated in a safe manner and therefore this proposed change meets the requirements of Title 10 CFR 5.36(c)(5).

With these changes, the Technical Specifications will continue to specify limiting conditions for operation which assure safe operation, maintenance and testing of the plant. Thus with the changes proposed in this license amendment request, the requirements of Title 10 CFR 50.36 continue to be met.

NUREG-1431, Revision 4, "Standard Technical Specifications, Westinghouse Plants"

NUREG-1431 provides format and content guidance for Technical Specifications for Westinghouse plants. The current NRC-approved version of NUREG-1431, Revision 4, provides the following COLR content guidance:

5.6.3 CORE OPERATING LIMITS REPORT

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

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NSPM

--------------------------------~E:\lIE:\AlE:~'~ ~OlrE:----------------------------------------

Licensees that have received prior ~RC approval to relocate lropical ~eport revision numbers and dates to licensee control need only list the number and title of the Topical ~eport, and the COL~ will contain the complete identification for each of the lrechnical ~pecification referenced lropical ~eports used to prepare the COL~ (Le., report number, title, revision, date, and any supplements). ~ee ~~C ADAM~ Accession ~o: ML 110660285 for details.

[Identify the lropical ~eport(s) by number, title, date, and ~~C staff approval document or identify the staff ~afety E:valuation ~eport for a plant specific methodology by ~~C letter and date.]

lrhis license amendment request proposes to add two topical report references to the COL~ list of ~~C-approved methodologies which will be used to determine core operating limits which assure the plant is operated in a safe manner. Consistent with the reviewer's note, these references will be identified by the topical report number and title as allowed by PI~GP license amendments 162 and 153 and ~~C letter dated August 4, 2011 (ML 110660285).

lrherefore, with the changes proposed in this license amendment, the lrechnical ~pecifications will identify ~~C-approved topical reports in accordance with the format and content guidance of ~U~E:G-1431 which assure safe operation of the plant.

4.2 Precedent

On August 14, 2012, ~outhern ~uclear Operating Company submitted an LA~ for the Joseph M. Farley ~uclear Plant (ML 12227A884) which proposed to revise their lrechnical ~pecification 5.6.5, "Core Operating Limits Report (COL~)," to reference and allow use of \Alestinghouse \AlCAP-16045-P-A, Addendum 1-A, "Qualification of the ~E:XU~ ~uclear Data Methodology," to determine core operating limits. In their February 28, 2013, (ML 13063A291) response to requests for additional information, ~~OC stated that their submittal also implies use of \AlCAP-16045-P-A.

4.3 Significant Hazards Consideration

~orthern ~tates Power Company, a Minnesota corporation (~~PM), evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 1 0 CF~ 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

~esponse: ~o

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This license amendment request proposes to revise the Technical Specifications to reference and allow use of WCAP-16045-P-A, "Qualification of the Two­Dimensional Transport Code PARAGON", and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology", for determining core operating limits.

The methodologies which this license amendment proposes for determination of core operating limits are improvements over the current methodologies in use at the Prairie Island Nuclear Generating Plant. The NRC staff reviewed and approved these methodologies and concluded that these analysis codes are acceptable as a replacement for the current analysis code. Thus core operating limits determined using the proposed codes continue to assure that the reactor operates safely and, thus, the proposed changes do not involve an increase in the probability of an accident.

Operation of the reactor with core operating limits determined by use of the proposed analysis codes does not increase the reactor power level, does not increase the core fission product inventory, and does not change any transport assumptions. Therefore the proposed methodology and Technical Specification changes do not involve a significant increase in the consequences of an accident.

Therefore, the proposed methodology change and associated Technical Specification changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No

This license amendment request proposes to revise the Technical Specifications to reference and allow use of WCAP-16045-P-A, "Qualification of the Two­Dimensional Transport Code PARAGON",and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology", for determining core operating limits.

The proposed changes provide revised methodology for determining core operating limits, but they do not change any system functions or maintenance activities. The changes do not involve physical alteration of the plant, that is, no new or different type of equipment will be installed. The changes do not alter assumptions made in the safety analyses but ensure that the core will operate within safe limits. These changes do not create new failure modes or mechanisms which are not identifiable during testing, and no new accident precursors are generated.

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Therefore, the proposed methodology change and associated Technical Specification changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No

This license amendment request proposes to revise the Technical Specifications to reference and allow use ofWCAP-16045-P-A, "Qualification of the Two­Dimensional Transport Code PARAGON", and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology", for determining core operating limits.

This license amendment proposes revised methodology for determining core operating limits. The proposed methodology is an improvement that allows more accurate modeling of core performance. The NRC has reviewed and approved this methodology for use in lieu of the current methodology, thus, the margin of safety is not reduced due to this change.

Therefore, the proposed methodology change and associated Technical Specification changes do not involve a significant reduction in a margin of safety.

Based on the above, Northern States Power Company, a Minnesota corporation, (NSPM) concludes that the proposed amendment does not involve a 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. ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant

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hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6. REFERENCES

None

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ENCLOSURE, ATTACHMENT 1

Technical Specification Pages (Markup)

5.0-37

1 page follows

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

Prairie Island Units 1 and 2

29. Caldon Engineering Report ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM or LEFM CheckPlus System";

30. WCAP-126IO-P-A, "VANTAGE+ Fuel Assembly Reference Core Report"; aru:i

31. WCAP-I261O-P-A and CENPD-404-P-A, Addendum I-A, "Optimized ZIRLOTM"~-:

32. WCAP-I6045-P-A. "Oualification of the Two-Dimensional Transport Code PARAGON'; and

33. WCAP-I6045-P-A. Addendum I-A. "Oualification of the NEXUS Nuclear Data Methodology".

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Unit 1 - Amendment No. 168 197!99 5.0-37Unit 2 - Amendment No. 158 186-l-8+

ENCLOSURE, ATTACHMENT 2

Technical Specification Pages (Retyped)

5.0-37

1 page follows

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

Prairie Island Units I and 2

29. Caldon Engineering Report ER-I57P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM or LEFM CheckPlus System";

30. WCAP-I26IO-P-A, "VANTAGE+ Fuel Assembly Reference Core Report";

31. WCAP-I261O-P-A and CENPD-404-P-A, Addendum I-A, "Optimized ZIRLO TM,,;

32. WCAP-I6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON'; and

33. WCAP-I6045-P-A, Addendum I-A, "Qualification of the NEXUS Nuclear Data Methodology".

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

Unit I - Amendment No. -l-9-+ 19-9 5.0-37 Unit 2 - Amendment No. -l-86 +8+