L-PI-23-027, Revision to 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals

Revision to 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals

ML23276B462
Person / Time
Site:	Prairie Island
Issue date:	10/03/2023
From:	Borgen T Northern States Power Company, Minnesota, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-PI-23-027
Download: ML23276B462 (1)

Text

1717 Wakonade Drive Welch, MN 55089

October 3, 2023 L-PI-23-027 10 CFR 50.55a

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Prairie Island Nuclear Generating Plant, Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License Nos. DPR-42 and DPR-60

Revision to 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspec tion (ISI) Intervals for Prairie Island Unit 1 and Unit 2

References:

1) NSPM Letter L-PI-19-002, 10 CFR 50.55a Requests Nos. 1-RR-5-10 and 2-RR-5-10, Proposed Alternative to Reactor Vessel Inservice Inspection (ISI)

Intervals for Prairie Island Unit 1 and Unit 2, dated June 13, 2019. (NRC ADAMS Accession Number ML19164A166)

2) NRC Letter subject: Prairie Island Nuclear Generating Plant, Units 1 AND 2 -

Proposed Alternative To The Requirements of the ASME CODE (EPID: L-2019-LLR-0055), dated November 5, 2019 (EPID L-2022-LLA-0084). (NRC ADAMS Accession Number ML19282A541)

Pursuant to 10 CFR 50.55a(z)(1), Nort hern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), requests U.S. Nuclear Regulatory Commission (NRC) authorization of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda for the Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2. N SPM previously requested this alternative in Reference 1 and NRC authorized the alternative in Reference 2. However, subsequent reviews identified that the informat ion included in Reference 1 was in error.

The specific error identified is with respect to the fluence value reported for the nozzle shell to intermediate shell circumferentia l weld (Weld W2) in Tables 3a and 3b of Reference 1, that documents the calculation of through-wall cracking frequency (TWCF) for Unit 1 and Unit 2. This error results in a change to which circumferential weld is limiting, but does not change the final calculated TWCF. In Reference 1, Tables 3a and 3b, the limiting circumferential weld was deter mined to be the intermediate shell to lower shell circumferential weld (Weld W3); how ever, when the fluence information for Weld W2 is corrected, Weld W2 becomes the limiti ng circumferential weld for Unit 2. Despite Document Control Desk L-PI-23-027 Page 2

the impact of the error on the calculation results, the TWCF contribution of the limiting circumferential weld (TWCF95-CW) remains 0.000E+00. The calculation of TWCF95-CW includes a subtraction of a constant value from the reference temperature for the limiting circumferential weld (RTMAX-CW). Because RTMAX-CW for both welds W2 and W3 is less than the constant, the subtraction results in a negative number and the calculation method sets the TWCF95-CW equal to zero. Therefore, the effect of the corrected fluence on the limiting circumferential weld does not change the result.

Additionally, new information has become available from the test results of surveillance capsules withdrawn from the Unit 1 and Unit 2 reactor vessels. NSPM previously submitted this information in the following two letters:

NSPM Letter L-PI-22-005, Prairie Island Unit 1 Reactor Vessel Material Surveillance Program Report, dated March 7, 2022. (NRC ADAMS Accession Numbers ML22067A148 thru ML22067A155)

NSPM Letter L-PI-23-002, Prairie Island Nuclear Generating Plant (PINGP) Unit 2 Reactor Vessel Material Surveillance Program Report, dated March 16, 2023.

(NRC ADAMS Accession Numbers ML23075A345 thru ML23075A352)

In order to correct the error and incorporate the new information, NSPM had a new calculation of TWCF95-TOTAL completed. NSPM decided to submit the enclosed revised relief request to include the new calculation results following a discussion with the NRC staff in a public meeting on June 5, 2023.

NSPM requests NRC authorization of the proposed alternative for the fifth ten-year interval of the PINGP lnservice Inspection (ISI) Program. Specifically, NSPM requests authorization to extend the Unit 1 and Unit 2 reactor pressure vessel ISI intervals from 10 years to 20 years. The revised relief request enclosed for Unit 1 (1-RR-5-10) and Unit 2 (2-RR-5-10) provides the basis and supporting information for the proposed alternative.

PINGP is currently in the fifth ten-year interval, which began on December 21, 2014, and is currently scheduled to end December 20, 2024. NSPM requests authorization of these 10 CFR 50.55a requests before the expiration of the current fifth ten-year interval.

Please contact Mr. Jeff Kivi at (612) 330-5788 or Jeffrey.L.Kivi@xcelenergy.com if there are any questions or if additional information is needed.

Document Control Desk L-Pl-23-027 Page 3

Summary of Commitments

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

Timothy P. Borgen Plant Manager, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota

Enclosure

cc: Administrator, Region Ill, USNRC Project Manager, Prairie Island, USNRC Resident Inspector, Prairie Island, USNRC L-PI-23-027 NSPM Enclosure Page 1 of 9

10 CFR 50.55a Request 1-RR-5-10 (PINGP Unit 1) Revision 1 10 CFR 50.55a Request 2-RR-5-10 (PINGP Unit 2) Revision 1 Proposed Alternative to Reactor Vessel Inservice Inspection (ISI) Intervals

Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Provides an Acceptable Level of Quality and Safety

1. American Society of Mechanical Engineers (ASME) Code Component(s) Affected

The affected components are the Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2 reactor vessels (RVs), specifically, the following ASME Boiler and Pressure Vessel (BPV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the RVs. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.

Category B-A welds are defined as Pressure Retaining Welds in Reactor Vessel.

Category B-D welds are defined as Full Penetration Welded Nozzles in Vessels.

Examination Item Category Number Description B-A B1.11 Circumferential Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section

2. Applicable Code Edition and Addenda

ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition though 2008 Addenda (Reference 1).

3. Applicable Code Requirement

IWB-2411, Inspection Program, requires volumetric examination of essentially 100% of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each ten-year interval.

The PINGP Unit 1 and Unit 2 fifth ten-year inservice inspection (ISI) interval is scheduled to end on December 20, 2024.

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4. Reason for Request

An alternative is requested from the requirement of IWB-2411, Inspection Program, that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each ten-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will reduce man-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use

Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter "NSPM"), proposes to extend the fifth ISI interval and perform the ASME Code required volumetric examinations of reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds on Units 1 and 2 in 2033 and 2034, respectively.1 The proposed inspection dates for PINGP Unit 1 and Unit 2 are consistent with the implementation plan presented in OG-10-238, Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval. PA-MSC-0120 (Reference 2).

In accordance with 10 CFR 50.55a(z)(1), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis, (Reference 3). Thus, the alternative provides an acceptable level of quality and safety.

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval (Reference 4). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for PINGP Unit 1 and Unit 2 were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for PINGP Unit 1 and Unit 2 are bounded by the results of the Westinghouse pilot plant qualifies PINGP Unit 1 and Unit 2 for an ISI interval extension.

Table 1a below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of PINGP Unit 1. Table 1b below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of PINGP Unit 2. Tables 2a, 2b, 3a and 3b provide additional information that was requested by the NRC and included in Appendix A of Reference 4. Revised or New Additional Information in the tables since Revision 0 is shaded.

1Revision 0 of these Prairie Island Relief Request Nos. 1-RR-5-10 and 2-RR-5-10 were approved by the NRC on November 5, 2019. (Reference 9)

L-PI-23-027 NSPM Enclosure Page 3 of 9

Table 1a Critical Parameters for the Application of Bounding Analysis for PINGP Unit 1 Revised or New Additional Information is shaded Plant-Specific Additional Parameter Pilot Plant Basis Basis Evaluation Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization (PTS) Transients in the NRC PTS Risk (Reference 5) Study (Reference 6) No Study are Applicable Through-Wall Cracking Frequency 1.76E-08 Events per 4.64E-13 Events per (TWCF) year (Reference 4) year (Calculated per No Reference 4)

Frequency and Severity of Design Basis 7 heatup/cooldown Bounded by 7 Transients cycles per year heatup/cooldown No (Reference 4) cycles per year Cladding Layers (Single/Multiple) Single Layer (Reference 4) Single Layer No

Table 1b Critical Parameters for the Application of Bounding Analysis for PINGP Unit 2 Revised or New Additional Information is shaded I 1 Plant-Specific Additional Parameter Pilot Plant Basis Basis Evaluation Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization (PTS) Transients in the NRC PTS Risk (Reference 5) Study (Reference 6) No Study are Applicable Through-Wall Cracking Frequency 1.76E-08 Events per 1.83E-13 Events per (TWCF) year (Reference 4) year (Calculated per No Reference 4)

Frequency and Severity of Design Basis 7 heatup/cooldown Bounded by 7 Transients cycles per year heatup/cooldown No (Reference 4) cycles per year Cladding Layers (Single/Multiple) Single Layer (Reference 4) Single Layer No

Tables 2a and 2b below provide summaries of the latest reactor vessel inspection for PINGP Unit 1 and Unit 2, respectively, and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the PINGP Unit 1 and Unit 2 reactor vessels.

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Table 2a Additional Information Pertaining to Reactor Vessel Inspection for PINGP Unit 1

The latest ISI examinations of PINGP Unit 1 Category B-A and B-D welds were performed to ASME Section XI, Appendix VIII, 1998 Edition with 2000 Addenda Inspection as modified by the PDI program and Federal Register, Part II, NRC 10 CFR methodology: Part 50, Industry Codes and Standards, amended requirements. Future inservice inspections will be performed to ASME Section XI, Appendix VIII methodology.

Number of past inspections: Four ten-year inservice inspections have been performed.

There were 54 indications identified in the beltline region of the RV during the last ISI. These subsurface indications are located in the nozzle to intermediate circumferential weld seam (Item 4 in Table 3a) and the intermediate to lower shell circumferential weld seam (Item 5 in Table 3a). These indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Twenty-three of these indications are within the inner 1/10 th or 1 inch of the reactor vessel thickness. These indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7). A disposition of the twenty-three flaws against the limits of the Alternate PTS Rule is shown in the tables below.

The following indications are located within the weld material of the reactor vessel beltline.

Through-Wall Extent, TWE (in) Scaled maximum Number of Flaws TWEMIN TWEMAX number of weld flaws (Axial/Circ.)

Number of 0 0.075 No Limit 0 indications found: 0.075 0.475 139 6 (0/6) 0.125 0.475 76 5 (0/5) 0.175 0.475 19 3 (0/3) 0.225 0.475 8 2 (0/2) 0.275 0.475 4 1 (0/1) 0.325 0.475 3 0

The following indications are located within the forging material of the reactor vessel beltline.

Through-Wall Extent, TWE (in) Scaled maximum Number of Flaws TWEMIN TWEMAX number of forging (Axial/Circ.)

flaws 0 0.075 No Limit 0 0.075 0.375 45 17 (0/17) 0.125 0.375 18 10 (0/10) 0.175 0.375 5 0

Proposed This inspection will be performed in 2033. The proposed inspection date is inspection consistent with the latest revised implementation plan, OG-10-238 (Reference schedule for 2).

balance of unit life:

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Table 2b Additional Information Pertaining to Reactor Vessel Inspection for PINGP Unit 2

The latest ISI examinations of PINGP Unit 2 Category B-A and B-D welds were performed to ASME Section XI, Appendix VIII, 1998 Edition with 2000 Addenda Inspection as modified by the PDI program and Federal Register, Part II, NRC 10 CFR methodology: Part 50, Industry Codes and Standards, amended requirements. Future inservice inspections will be performed to ASME Section XI, Appendix VIII methodology.

Number of past inspections: Four ten-year inservice inspections have been performed.

There were 17 indications identified in the beltline region of the RV during the last ISI. These subsurface indications are located in the nozzle to intermediate shell circumferential weld seam (Item 4 in Table 3b), and are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Sixteen of these indications are within the inner 1/10 th or 1 inch of the reactor vessel thickness.

These indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7). A disposition of the sixteen flaws against the limits of the Alternate PTS Rule is shown in the tables below.

The following indications are located within the weld material of the reactor vessel beltline.

Through-Wall Extent, TWE (in) Scaled maximum Number of Flaws TWEMIN TWEMAX number of weld flaws (Axial/Circ.)

0 0.075 No Limit 0 0.075 0.475 139 12 (0/12)

Number of 0.125 0.475 76 7 (0/7) indications found: 0.175 0.475 19 2 (0/2) 0.225 0.475 8 2 (0/2) 0.275 0.475 4 1 (0/1) 0.325 0.475 3 0

The following indications are located within the forging material of the reactor vessel beltline.

Through-Wall Extent, TWE (in) Scaled maximum Number of Flaws TWEMIN TWEMAX number of forging (Axial/Circ.)

flaws 0 0.075 No Limit 0 0.075 0.375 45 4 (0/4) 0.125 0.375 18 3 (0/3) 0.175 0.375 5 1 (0/1) 0.225 0.375 2 0

Proposed This inspection will be performed in 2034. The proposed inspection date is inspection consistent with the latest revised implementation plan, OG-10-238 (Reference schedule for 2).

balance of unit life:

L-PI-23-027 NSPM Enclosure Page 6 of 9

Tables 3a and 3b summarize the inputs and outputs for the calculation of through-wall cracking frequency (TWCF) for Unit 1 and Unit 2, respectively.

Table 3a Details of TWCF Calculation for PINGP Unit 1 at 54 Effective Full Power Years (EFPY)

Revised or New Additional Information is shaded I I Inputs(1)

Reactor Coolant System Temperature, TC [°F]: N/A Twall [inches]: 6.889 Region and Material Material Heat Cu Ni R.G. Fluence Number Component ID Number [wt%] [wt%] 1.99 CF [ºF] RTNDT(U) [ºF] [n/cm2, E >

Description Pos. 1.0 MeV]

1 Nozzle Shell Forging B 21744/38384 0.08 0.68 1.1 51.0 -4°F I I 3.37E+19 2 Intermediate Shell Forging C 21918/38566 0.07 0.80 2.1 69.0 14°F 5.63E+19 3 Lower Shell Forging D 21887/38530 0.07 0.66 1.1 44.0 -4°F - -I I 5.53E+19 Nozzle Shell to 4 Intermediate Shell W2 2269 0.15 0.15 1.1 79.5 0°F 3.63E+19 Circumferential Weld -

Intermediate Shell to 5 Lower Shell W3 1752 0.13 0.13 2.1 99.1 -13°F 5.53E+19 Circumferential Weld - -

Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Fluence FF Material Region XX RTMAX-XX [n/cm2, E (Fluence T30 Number [°R] >1.0 MeV] Factor) [ºF] TWCF95-XX (From Above)

Limiting Forging - 572.00 5.63E+19 1.4250 98.33 1.8554E-13 FO 2 2.5000 I -

Limiting Circumferential Weld 5 2.5000 587.56 5.53E+19 1.4217 140.89 0.000E+00 I - I I

- CW TWCF95-TOTAL = (FOTWCF95-FO + CWTWCF95-CW): 4.64E-13 (1) Material properties and fluence inputs are based on plant-spec ific assessment. Revised or new additional information since Revision 0 comes from correction of W2 location error and new reactor vessel surveillance capsule results and is shaded.

L-PI-23-027 NSPM Enclosure 1 Page 7 of 9

Table 3b Details of TWCF Calculation for PINGP Unit 2 at 54 Effective Full Power Years (EFPY)

Revised or New Additional Information is shaded I I Inputs(1)

Reactor Coolant System Temperature, TC [°F]: N/A Twall [inches]: 6.889 Region and Material Material Heat Cu Ni R.G. Fluence Number Component ID Number [wt%] [wt%] 1.99 CF [ºF] RTNDT(U) [ºF] [n/cm2, E >

Description Pos. 1.0 MeV]

1 Nozzle Shell Forging B 22231/39088 0.07 0.73 1.1 44.0 -13°F I I 3.37E+19 2 Intermediate Shell Forging C 22829 0.07 0.75 1.1 44.0 14°F 5.66E+19 3 Lower Shell Forging D - -22642 0.08 0.67 2.1 74.4 -4°F 5.58E+19 I I I I I I Nozzle Shell to 4 Intermediate Shell W2 1752 (1263) 0.13 0.13 2.1 101.4 -13°F 3.61E+19 Circumferential Weld - -

Intermediate Shell to 5 Lower Shell W3 2721 (1263) 0.09 0.11 2.1 87.5 -31°F 5.58E+19 Circumferential Weld - -

Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Fluence FF Material Region XX RTMAX-XX [n/cm2, E (Fluence T30 Number [°R] >1.0 MeV] Factor) [ºF] TWCF95-XX (From Above)

Limiting Forging - 561.57 5.58E+19 1.4233 105.90 7.3356E-14 FO 3 2.5000 I -

Limiting Circumferential Weld 4 2.5000 581.90 3.61E+19 1.3336 135.23 0.000E+00

- CW I I - I I

TWCF95-TOTAL = (FOTWCF95-FO + CWTWCF95-CW): 1.83E-13 (1) Material properties and fluence inputs are based on plant-spec ific assessment. Revised or new additional information since Revision 0 comes from correction of W2 location error and new reactor vessel surveillance capsule results and is shaded.

L-PI-23-027 NSPM Page 8 of 9

6. Duration of Proposed Alternative

In Reference 9, the NRC staff authorized Requests Nos. 1-RR-5-10 and 2-RR-5-10 until December 20, 2034. The proposed revision does not request a change to this duration.

7. Precedents

1. Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574), dated April 30, 2013.

(Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140)

2. Vogtle Electric Generating Plant, Units 1 and 2 - Request for Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT-06 (TAC Nos. MF2596 and MF2597), dated March 20, 2014. (ADAMS Accession Number ML14030A570)

3. Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAC Nos. MF1922 and MF1923),

dated March 26, 2014. (ADAMS Accession Number ML14079A546)

4. Sequoyah Nuclear Plant, Units 1 and 2 - Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval (TAC Nos. MF2900 and MF2901), dated August 1, 2014. (ADAMS Accession Number ML14188B920)

5. Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596), dated December 10, 2014. (ADAMS Accession Number ML14303A506)

6. Wolf Creek Generating Station - Request for Relief Nos. I3R-08 and I3R-09 for the Third 10-Year Inservice Inspection Program Interval (TAC Nos. MF3321 and MF3322),

dated December 10, 2014. (ADAMS Accession Number ML14321A864)

7. Callaway Plant, Unit 1 - Request for Relief I3R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAC No. MF3876), dated February 10, 2015. (ADAMS Accession Number ML15035A148)

8. Braidwood Station, Units 1 and 2 - Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) (CAC Nos. MF8191 and MF8192), dated March 15, 2017. (ADAMS Accession Number ML17054C255)

9. South Texas Project, Units 1 and 2 - Relief from the Requirements of the ASME Code Regarding the Third 10-Year Inservice Inspection Program Interval (EPID L-2018-LLR-0010), dated July 24, 2018. (ADAMS Accession Number ML18177A425)

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8. References

1. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda, American Society of Mechanical Engineers, New York.

2. PWROG Letter OG-10-238, Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval. PA-MSC-0120, July 12, 2010. (ADAMS Accession Number ML11153A033)

3. NRC Regulatory Guide 1.174, Revision 1, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, U.S. Nuclear Regulatory Commission, November 2002. (ADAMS Accession Number ML003740133)

4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, October 2011. (ADAMS Accession Number ML11306A084)

5. NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), U.S. Nuclear Regulatory Commission, March 2010. (ADAMS Accession Number ML15222A848)

6. NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS)

Risk Results to Additional Plants, U.S. Nuclear Regulatory Commission, December 14, 2004. (ADAMS Accession Number ML042880482)

7. Code of Federal Regulations, 10 CFR Part 50.61a, Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, Number 1, dated January 4, 2010, and Number 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010.

8. NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988. (ADAMS Accession Number ML003740284)

9. Letter from NRC to NSPM re: Prairie Island Nuclear Generating Plant, Units 1 and 2 -

Proposed Alternative to the Requirements of the ASME Code (EPID: L-2019-LLR-0055), dated November 5, 2019. (ADAMS Accession Number ML19282A541)