ML23013A224

Proposed License Amendment Request to Revise the Applicability Term for Reactor Coolant System Heatup and Cooldown Pressure Temperature Limitation Figures

ML23013A224
Person / Time
Site:	Millstone
Issue date:	01/13/2023
From:	James Holloway Dominion Energy Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
22-361
Download: ML23013A224 (1)

Text

Dominion Energy Nucl ea r Connecticut, Inc.

5000 Dom inion Boulevard, Glen Allen, VA 23060 ~ Dominion Dominion Energy.com

~ Energy" January 13, 2023 U.S. Nuclear Regulatory Commission Serial No.22-361 Attention: Document Control Desk NRA/SS: RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 PROPOSED LICENSE AMENDMENT REQUEST TO REVI SE THE APPLICABILITY TERM FOR REACTOR COOLANT SYSTEM HEATUP AN D COOLDOWN PRESSURE-TEMPERATURE LIMITATIONS FIGURES Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENG) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed change revises MPS3 TS 3.4.9.1, "Reactor Coolant System Pressure/Temperature Limits," to reflect that Figures 3.4-2 and 3.4-3 (Heatup and Cooldown Limitations, respectively) are applicable up to 54 effective full power years (EFPY). Additional changes are proposed to correct typographical errors. provides DENC's description and assessment of the proposed change. provides the marked-up TS pages. Attachment 3 provides the evaluation of Adjusted Reference Temperature and Reference Temperature Shifts for MPS3, which was performed using the methodology of Regulatory Guide 1.99, Revision 2. Attachment 4 provides the Projected Upper Shelf Energy Values for 54 EFPY.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in . DENG has also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite, or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with approval of the proposed change.

The proposed amendment has been reviewed and approved by the station's Facility Safety Review Committee.

DENG calculates that MPS3 will reach 32 EFPY as early as the spring of 2024. DENG requests approval of this LAR by January 13, 2024, with a 60-day implementation period, to support an update to the EFPY limit prior to expiration.

Serial No.22-361 Docket No. 50-423 Page 2 of 3 In accordance with 10 CFR 50.91 (b), a copy of this LAR is being provided to the State of Connecticut.

If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.

Sincerely, James E. Holloway Vice President - Nuclear Engineering and Fleet Support COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by James E. Holloway who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this !gthday of :Jt,Muo v-1 ,2023.

1:______

My Commission Expires: __,)=2---1-(3=-1_.l...;_t.......

CRAIG D SLY Notary Public . .

{;§mmonwealth of Vlrgm1a R~ , # 7518653 Attachments:

1. Description and Assessment of Proposed Change

2. Marked-up Technical Specification Pages

3. Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts

4. Projected Upper Shelf Energy Values for 54 EFPY Commitments made in this letter: None

Serial No.22-361 Docket No. 50-423 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road, Suite 105 King of Prussia, PA 19406-1415 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.22-361 Docket No. 50-423 Attachment 1 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGE Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 1 of 11 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed change revises MPS3 TS 3.4.9.1, "Reactor Coolant System Pressure/Temperature Limits," to reflect that Figures 3.4-2 and 3.4-3 (Heatup and Cooldown Limitations, respectively) are applicable up to 54 effective full power years (EFPY). Additional changes are proposed to correct typographical errors.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation All components of the Reactor Coolant System (RCS) are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. MPS3 TS Limiting Condition for Operation (LCO) 3.4.9.1 limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

MPS3 TS Figures 3.4-2 and 3.4-3 contain pressure/temperature (P/T) limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operation. The curves are typically used for operational requirements during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. A heatup or cooldown is defined as a temperature increase or decrease of greater than or equal to 10 oF in any one hour period. This definition of heatup and cooldown is based upon the definition of isothermal conditions described in American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PVC),

Section XI, Appendix E, Evaluation of Unanticipated Operating Events.

2.2 Current Technical Specification Requirement MPS3 TS LCO 3.4.9.1 requires the Reactor Coolant System (except the pressurizer) temperature, pressure, and heatup and cooldown rates of ferritic materials be limited in accordance with the limits shown in TS Figures 3.4-2 and 3.4-3. These figures are currently stated as applicable for 32 EFPY.

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 2 of 11 2.3 Reason for the Proposed Change A LAR dated April 23, 2001 (Reference 1) replaced MPS3 TS Figures 3.4-2 and 3.4-2, which had been applicable to 10 EFPY, with new curves applicable to 32 EFPY. The NRC approved this change with the issuance of Amendment 197, dated August 27, 2001 (Reference 2). Subsequently, an application was submitted by letter dated January 20, 2004 (Reference 3), for renewal of the MPS3 operating license for an additional 20 years.

The NRC issued a Renewed Facility Operating License for MPS3 by letter dated November 28, 2005 (Reference 4), which permits unit operation until November 25, 2045.

DENC calculates that MPS3 will reach 32 EFPY as early as the spring of 2024. Prior to this date, NRC approval is needed to increase the applicable EFPY value for the P/T curves in TS 3.4.9.1. DENC has determined that the current P/T limit curves are applicable for the extended period of 54 EFPY.

Extending the applicability date of the existing P/T limit curves is preferable to using new P/T limit curves, since it would avoid revisions to plant procedures and retraining of plant operators.

2.4 Description of Proposed Change The proposed change revises the figure title and x-axis for TS Figures 3.4-2 and 3.4-3 as follows (added text is shown below in bold, underlined type, deleted text is shown in strikethrough):

• Figure 3.4-2, MILLSTONE 3 REACTOR COOLANT SYSTEM, Heatup Limitations for Fluence up to 2.72 1.97E+19 n/cm2 (32 EFPY) (54 EFPY)

• Figure 3.4-3, MILLSTONE 3 REACTOR COOLANT SYSTEM, Cooldown Limitations for Fluence up to 2.72 1.97E+19 n/cm2 (32 EFPY) (54 EFPY)

The administrative change from cm to cm2 corrects a typographical error to reflect the standard units for fluence.

The proposed change revises the x-axis in TS Figures 3.4-2 and 3.4-3 as follows:

• Indicated Cold Leg Temperature (x°F)

This administrative change corrects a typographical error to reflect the standard symbol for degrees Fahrenheit.

Attachment 2 provides the marked-up TS pages.

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 3 of 11

3.0 TECHNICAL EVALUATION

By letter dated November 9, 2021, the NRC approved License Amendment 280 for MPS3 (Reference 6). This amendment approved a Measurement Uncertainty Recapture (MUR)

Power Uprate, which increased the units authorized reactor core power level by approximately 1.6% rated thermal power from 3650 megawatts thermal (MWt) to 3709 MWt. A Reactor pressure vessel (RPV) integrity evaluation based on neutron fluence, Pressurized Thermal Shock (PTS), Adjusted Reference Temperature (ART), and Upper Shelf Energy (USE) was provided for the MUR LAR (Reference 7). The technical basis for reassessment of the limiting ART and PTS screening is documented in Attachment 3 of this LAR. Attachment 4 provides the Projected Upper Shelf Energy values for 54 EFPY.

Heatup and cooldown limit curves are calculated using the most limiting value of the reference nil-ductility transition temperature (RTNDT) corresponding to the limiting material in the beltline region of the reactor pressure vessel (RPV). The most limiting RTNDT of the material in the core (beltline) region of the RPV was determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced reference temperature shift ( RTNDT). RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactors life, RTNDT due to the radiation exposure associated with that time period was added to the original unirradiated RTNDT. Using the ART values, P/T limit curves were determined in accordance with the requirements of 10 CFR 50, Appendix G.

The existing 32 EFPY P/T limit curves are acceptable for continued operation up to 54 EFPY because the limiting ART values at the 1/4T and 3/4T locations (i.e., tips of the ASME Code reference flaw when the flaw is assumed at the inside diameter and outside diameter locations) for 54 EFPY are bounded by the 1/4T and 3/4T ART values used to develop the existing 32 EFPY P/T limit curves. The methodology used to develop the 32 EFPY P/T limit curves is based on the 1995 edition of ASME B&PVC,Section XI, Appendix G and Code Case N-640. The reference critical stress intensity (KIc) is a direct function of ART. Since the limiting ART at 54 EFPY is bounded, the values of KIc at 54 EFPY are bounded and the resulting 54 EFPY P/T limits curves would be bounded by the 32 EFPY curves. Because the 54 EFPY ART values are bounded by the 32 EFPY ART values, the 54 EFPY values for PTS, USE, Cold Overpressure Protection System (COPS) relief valve setpoints and COPS enable temperature are also bounded by their respective 32 EFPY values.

The remaining factors that can potentially impact the P/T limits contained in the MPS3 TS Figure 3.4-2 and 3.4-3 and the associated COPS relief valve setpoints, such as heatup and cooldown rates, instrument uncertainties, pressure correction factors, mass and energy addition transients, relief valve capacity and stroke times, etc., are unchanged.

In Attachment 3, the limiting reactor vessel material ART values for 54 EFPY, considering the MUR power uprate and the credible reactor vessel surveillance capsule W data documented in WCAP-16629-NP (Reference 9), were shown to be less than the limiting

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 4 of 11 beltline material ART values used in development of the existing 32 EFPY P/T limit curves (contained in Reference 1). In addition, the chemistry factor (CF) values for plates B9820-1, B9820-2, and B9820-3 were recalculated using unrounded Copper (Cu) and Nickel (Ni) chemistry data taken directly from the applicable Certified Material Test Reports (CMTRs). The recalculated results are shown in Table 6 and Figures 6, 7 and 8 of . The use of unrounded average percent Cu and percent Ni chemistry for each of these materials provides a more accurate calculation of the associated CF. The Regulatory Guide 1.99, Revision 2 (Reference 5) methodology was then used along with the surface fluence from the MUR LAR (Reference 7), the revised CF, and the credible surveillance capsule data to calculate ART values for the MPS3 reactor vessel materials at 54 EFPY. The revised 54 EFPY ART calculations are summarized in Tables 6 and 7 of Attachment 3.

As shown in Attachment 3, the calculated limiting material ART values are less than those used for the development of the current TS P/T limit curves in Reference 1. The new ART values include consideration of credible surveillance data, reflect a recalculation of chemistry factors, and include the consideration of MUR power uprate fluence projections. This reassessment of the ART values resulted in the lower shell plate B9820-2 becoming the limiting beltline material. Prior to the availability of credible surveillance data, the surveillance material intermediate shell plate B9805-1 had been the limiting beltline material. The lower ART values are primarily due to the use of the MPS3 credible surveillance data.

The extension of the P/T Limits to 54 EFPY is supported by Attachment 3 of this LAR. As evidenced in the Attachment 3 analysis, the limiting ART values at 54 EFPY remain less than the limiting ART values used in the development of the existing MPS3 32 EFPY TS P/T limit curves. This means that at 54 EFPY, the ART values used for the development of Figures 3.4-2 and 3.4-3 remain conservative and bounding.

In addition to ART values, the development of the TS P/T Limits must consider instrument uncertainty, hydraulic pressure corrections. The conversion from gauge to absolute pressure must also be applied to the unadjusted allowable pressure as determined using the methodology of ASME B&PVC,Section XI, Appendix G. The pressure and temperature indicator uncertainties were reviewed and verified to be valid for use in the development of the TS Figures. The hydraulic pressure corrections and the conversion from gauge to absolute pressure were also reviewed and verified to be valid. Therefore, the P/T limits currently documented in TS Figures 3.4-2 and 3.4-3 remain valid for 54 EFPY. A more detailed discussion is contained in Attachment 3 to this LAR.

For the three reactor vessel beltline plate materials (B9820-1, B9820-2, and B9820-3),

the use of unrounded average chemistry values for weight percent Cu and Ni also has the potential to impact the 54 EFPY estimates of the bounding PTS reference temperature (RTPTS) and USE. 10 CFR 50.61 defines a maximum PTS screening criteria for plate material of 270°F. The bounding RTPTS estimate for 54 EFPY was determined to be 124.1°F in B9020-1, which continues to satisfy the above criterion. Attachment 3 provides a detailed reassessment of the RTPTS projection for 54 EFPY. For USE, which is an

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 5 of 11 indicator of maximum material toughness, 10 CFR 50, Appendix G defines a minimum criteria of 50 ft-lbs. The bounding USE estimate for 54 EFPY is 60.0 ft-lbs., which is applicable to plate B9820-2. Thus, the USE criterion defined in 10 CFR 50, Appendix G continues to be satisfied at 54 EFPY. Attachment 4 provides the Projected Upper Shelf Energy values for 54 EFPY.

Additionally, 10 CFR 50, Appendix G requires that P/T limits be developed to bound all ferritic materials in the RPV. Components beyond the beltline projected to experience fluence greater than 1x1017 n/cm2 (including nozzles and vessel plates far from the core mid-line) require an assessment to determine whether irradiation embrittlement will have a large effect on the ART and the associated P/T Limit Curves. The Pressurized Water Reactor Owners Group (PWROG) issued report PWROG-15109-NP-A (Reference 8) to address the RPV Nozzles potentially in the Extended Beltline for Appendix G Evaluations.

The report provides substantiation such that plant-specific nozzle analysis for evaluation is not required for generation of P/T Limit Curves, within certain criteria.

The maximum fluence for the MPS3 Inlet and Outlet Nozzles and Nozzle Welds (as documented in the MUR submittal (Reference 7)) for 54 EFPY is 1.29 x 1017 n/cm2. The criterion for non-evaluation is a maximum permissible surface fluence of 4.28 x 1017 n/cm2. Because the fluence on the extended beltline nozzles and welds is below 4.28 x 1017 n/cm2, the MPS3 extended beltline nozzles and nozzle welds require no further analysis or evaluation, and are bounded by the results of PWROG-15109-NP-A. The nozzle shell plates and associated longitudinal welds are bounded by the assessment of the limiting beltline material B9820-2. Section 3.4 of Attachment 3 provides a more detailed discussion of the nozzle shell and associated longitudinal welds. The peak 54 EFPY fluence value of 2.72 x 1019, as documented in MUR LAR Tables IV-1 and IV-2, is reflected as part of the proposed change to the title for TS Figures 3.4-2 and 3.4-3.

The pressurizer pressure relief valve (PORV) setpoint curves for COPS have been evaluated and determined to remain bounding and conservative for the purpose of limiting the maximum pressure in the reactor vessel during reactor start-up and shutdown operation in accordance with ASME B&PVC,Section XI, Appendix G, Section G-2215.

The COPS enable temperature has been reevaluated with consideration of the revised limiting ART. Since the 54 EFPY 1/4T ART is less than the 1/4T ART currently used for the 32 EFPY P/T curves, the COPS enable temperature remains unchanged. The COPS enable temperature is based on a minimum coolant temperature of 200°F plus instrument uncertainty as specified in Section G-2215.

In conclusion, DENC has determined that the P/T curves remain valid through 54 EFPY considering the neutron fluence projections documented in MUR LAR Table IV-2, primarily because the revised limiting ART values remain bounded by the ART values used to develop the 32 EFPY P/T limits. DENC has reviewed the pressure and temperature corrections and conversions and verified them to be valid. DENC has also concluded that the 10 CFR 50, Appendix G screening criteria applicable to neutron fluence, RTPTS, and USE continue to be satisfied for MPS3 through 54 EFPY. The existing

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 6 of 11 COPS relief valve setpoint curves and COPS enable temperature also remain valid through 54 EFPY.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria

• 10 CFR 50, Appendix G requires that the P/T limits for the facilitys reactor pressure vessel (RPV) be at least as conservative as those obtained by following the linear elastic fracture mechanics methodology of ASME B&PVC,Section XI, Appendix G. The P/T limits in TS 3.4.9.1 are unchanged. Only the applicability term is proposed to be increased to 54 EFPY for MPS3 instead of 32 EFPY. Therefore, DENC concludes that the MPS3 RPV will continue to meet RPV integrity regulatory requirements through 54 EFPY.

• 10 CFR 50, Appendix H establishes requirements for each facility related to its RPV material surveillance. These regulatory requirements will continue to be met, with no changes to the surveillance capsule removal schedule.

• Regulatory Guide 1.99, Revision 2 Radiation Embrittlement of Reactor Vessel Materials, contains guidance on methodologies the NRC considers acceptable for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. This RG was used for the calculation of ART values at the 1/4T and 3/4T locations that were previously calculated in MPS3 MUR LAR (Reference 7). The proposed change does not alter the existing 32 EFPY ART values used to develop TS Figures 3.4-2 and 3.4-3. The revised 54 EFPY ART values have been determined to be bounded by the existing 32 EFPY ART values. The proposed change only increases the applicability term to 54 EFPY for MPS3. Therefore, the proposed change has no effect on the application of RG 1.99, Revision 2.

4.2 Precedents Dominion Energy Virginia submitted a LAR for Surry Power Station Units 1 and 2 by letter dated September 19, 2019 (ADAMS Accession Number ML19269B775), which updated the cumulative core burnup applicability limit from 48 to 68 EFPY for TS Figures 3.1-1 and 3.1-2. The NRC approved the LAR as Amendments 302 and 302, by letter dated December 8, 2020 (ADAMS Accession Number ML20148M359).

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 7 of 11 Duke Energy Carolinas, LLC. submitted a LAR for McGuire Units 1 and 2 by letter dated December 20, 2021 (ADAMS Accession Number ML21355A362),

to extend the applicability of the P/T limit curves in Unit 1 TS Figures 3.4.3-1, 3.4.3-2, and 3.4.3-5 from 34 EFPY to 54 EFPY, and Unit 2 TS Figures 3.4.3-3, 3.4.3-4, and 3.4.3-6 for Unit 2 from 34 EFPY from 34 to 38.6 EFPY. Similar to this LAR, the P/T curves were unchanged, and only the applicability of the existing curves was updated. The NRC approved the LAR as Amendment 326 (Unit 1) and 305 (Unit 2) by letter dated November 29, 2022 (ADAMS Accession Number ML22290A101).

4.3 No Significant Hazards Consideration Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed change revises MPS3 TS 3.4.9.1, "Reactor Coolant System Pressure/Temperature Limits," to reflect that Figures 3.4-2 and 3.4-3 (Heatup and Cooldown Limitations, respectively) are applicable up to 54 effective full power years (EFPY). Additional changes are proposed to correct typographical errors.

DENC has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No No changes are being made to the existing P/T limit curves in TS 3.4.9.1 Figures 3.4-2, and 3.4-3. The P/T limit curves are only being revised to change the applicability period from 32 to 54 EFPY and correct typographical errors.

The changes to the TS figures are applicable to normal plant operations and do not influence the probability of occurrence or safety analysis considerations for design basis accidents. Consequently, there will be no change to the probability or consequences of accidents previously evaluated. Operating the facility in accordance with the P/T limit curves ensures that stresses caused by the thermal gradient through the Reactor Pressure Vessel (RPV) beltline material remain bounded by the stress analyses. The proposed amendment does not involve operation of required structures, systems, or components in a manner or configuration different

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 8 of 11 than previously recognized or evaluated. No radiological barriers are affected by the change.

Therefore, the proposed change does 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 No changes are being made to the existing P/T limit curves in TS 3.4.9.1 Figures 3.4-2, and 3.4-3. The P/T limit curves are only being revised to change the applicability period from 32 to 54 EFPY and correct typographical errors.

The change does not involve a modification of plant structures, systems, or components. The change will not affect the manner in which the plant is operated and will not degrade the reliability of structures, systems, or components. Equipment protection features will not be deleted or modified, equipment redundancy or independence will not be reduced, and supporting system performance will not be affected. No new failure modes or mechanisms will be introduced as a result of this proposed change.

Therefore, the proposed 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 No changes are being made to the existing P/T limit curves in TS 3.4.9.1 Figures 3.4-2, and 3.4-3. The P/T limit curves are only being revised to change the applicability period from 32 to 54 EFPY and correct typographical errors.

The proposed amendment does not involve: 1) a physical alteration of the plant, 2) a change to any setpoints for parameters associated with protection or mitigation actions or 3) any adverse impact on the fission product barriers or parameters associated with licensed safety limits. There are no changes

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 9 of 11 to either the containment analysis or to the analysis of any design basis event.

Appendix G to 10 CFR 50 describes the conditions that require P/T limits and provides the general bases for these limits. Operating limits based on the criteria of Appendix G, as defined by applicable regulations, codes and standards, provide reasonable assurance that non-ductile or rapidly propagating failure will not occur. The P/T limits are prescribed for all plant modes to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause non-ductile failure of the reactor coolant pressure boundary.

Calculation of P/T limits in accordance with the criteria of Appendix G to 10 CFR 50 and applicable regulatory requirements ensures that adequate margins of safety are maintained and there is no significant reduction in a margin of safety.

No change is being made to the existing P/T limit curves, only the applicability period associated with the P/T limits is being extended. Since the P/T limits remain unchanged and conservative and bounding there is no reduction in a margin of safety.

The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. There is no change or impact on any safety analysis assumption or on any other parameter affecting the course of an accident analysis supporting the basis of any TS. The proposed change does not involve any increase in calculated off-site dose consequences.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above information, DENC concludes that the proposed changes do not involve a significant hazards consideration, under standards set forth in 10 CFR 50.92(c), Issuance of Amendment, and accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion Based on the above discussions, (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

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 10 of 11 Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

S 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 hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents 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.

6.0 REFERENCES

1. Letter from Eugene S. Grecheck (Dominion Nuclear Connecticut, Inc.) to USNRC, Millstone Nuclear Power Station, Unit No. 3, Technical Specification Change Request 3-11-00, Reactor Coolant System Heatup and Cooldown Curves, April 23, 2001 (ADAMS Accession Number ML011210201).

2. Letter from Victor Nerses (USNRC) to R. P. Necci (Dominion Nuclear Connecticut, Inc.), Millstone Nuclear Power Station, Unit No. 3 - Issuance of Amendment Re:

Reactor Coolant System Heatup and Cooldown Curves (TAC No. MB1785),

August 27, 2001 (ADAMS Accession Number ML012060343).

3. Letter from David A. Christian (Dominion Nuclear Connecticut, Inc.) to USNRC, Millstone Power Station, Units 2 and 3, Applications for Renewed Operating Licenses, January 20, 2004 (ADAMS Accession Number ML040260070).

4. Letter from Johnny Eads (USNRC) to David A. Christian (Dominion Nuclear Connecticut, Inc.), Millstone Power Station, Units 2 and 3 - Issuance of Renewed Facility Operating License Nos. DPR-65 and NPF-49), November 28, 2005 (ADAMS Accession Package Number ML053220382).

5. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, May 1988.

Serial No.22-361 Docket No. 50-423 Attachment 1, Page 11 of 11

6. Letter from Richard V. Guzman (USNRC) to Daniel G. Stoddard (Dominion Energy Nuclear Connecticut, Inc), Millstone Power Station Unit No. 3 - Issuance of Amendment No. 280 Regarding Measurement Uncertainty Recapture Power Uprate (EPID L-2020-LLS-0002), November 9, 2021 (ADAMS Accession Number ML21262A001).

7. Letter from Mark D. Sartain (Dominion Energy Nuclear Connecticut, Inc.) to USNRC, "Millstone Power Station Unit 3 Proposed License Amendment Request Measurement Uncertainty Recapture Power Uprate," dated November 19, 2020 (ADAMS Accession Number ML20324A703).

8. Pressurized Water Reactor Owners Group Report, PWROG-15109-NP-A, Rev. 0, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020, (ADAMS Accession Number ML20024E573).

9. WCAP-16629-NP, Revision 0, Analysis of Capsule W from the Dominion Nuclear Connecticut Millstone Unit 3 Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Corp., September 2006 (ADAMS Accession Number ML062850221).

Serial No.22-361 Docket No. 50-423 Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGES Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.22-361 Docket No. 50-423 , Page 1 of 4

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Serial No.22-361 Docket No. 50-423 , Page 3 of 4

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Serial No.22-361 Docket No. 50-423 Attachment 3 EVALUATION OF ADJUSTED REFERENCE TEMPERATURES AND REFERENCE TEMPERATURE SHIFTS Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.22-361 Docket No. 50-423 File No.: 1901402.301 Attachment 3, Page 1 of 30 Project No.: 1901402 13 Structural Integrity Quality Program Type: ~ Nuclear D Commercial Associates, Inc.

CALCULATION PACKAGE PROJECT NAME:

1901402 Update of Millstone Unit 3 P-T Curves CONTRACT NO.:

4000020831 CLIENT: PLANT:

Dominion Energy, Inc. Millstone Unit 3 CALCULATION TITLE:

Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts Preparer(s) &

Document Affected Project Manager Approval Revision Description Checker(s)

Revision Pages Signature & Date Si natures & Date 0 All Initial Issue Approved By: Prepared By:

October 21, 2022 Checked By:

October 21, 2022

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 2 of 30 Table of Contents

1.0 INTRODUCTION

................................................. ...... .. ................ .............. ........ ... ....... 4 2.0 METHODOLOGY ..... ... .. ... ..................... .............. ... ......... ........................ .... ... ..... .. .. .. .. 4 3.0 DESIGN INPUTS .... .. .. .. .................................... .. .. ... .. ....................... .. ...... .. ................ 6 3.1 32 EFPY ART Tables .......................... .. .. .... ........... .. ........................ ............... 9 3.2 Credible Surveillance Data ...................... .. ........................................ ... ........... 9 3.3 Fluence Values ................................................................................ .............. 12 3.4 Extended Beltline ......... ... ... ... ............ ...... ... .. ................................... .... ........ .. 13 3.5 Pressurized Thermal Shock ................ .. .. .. ....... .......... ....... ....... .... ................. 16 4.0 ASSUMPTIONS .... .... .. .. ........................................... ............................ ...... .... .... .... ... 17 5.0 CALCULATIONS ..... .... .. ..... ............................ .. ...... .. ................................................. 18 5.1 RT NDT Values .. ..... ................................. .. .. ..................................... .. .. ... ......... 18 5.2 Pressurized Thermal Shock ................ ....... .. .... .. .................. .. ...... ........ .. .. ..... 19

6.0 CONCLUSION

S ........................................................................................................ 23

7.0 REFERENCES

... ..... ........... ......... ..... .... ....... ...... .... ........ ........ .... .. ... ........... .. .. ... ......... 24 APPENDIX A SUPPORTING FILES .................................. .... .. .............................. .. ............ A-1 APPENDIX B CMTR INFORMATION [19] .................... ...... .. ............................... .. .. .. .......... 8-1 List of Tables Table 1: MPS3 Reactor Vessel Beltline Chemistry and Initial RTNDT Values [6] .......... ...... .. .. 8 Table 2: Docketed Beltline Chemistries and Initial RT NDT Values from Millstone LRA [15] ... . 9 Table 3: Credible Surveillance Data for Plate B-9805-1 in MPS3 [12] .... .. ........... .. .............. 10 Table 4: MPS3 54 EFPY OT Fluence Values [13] ........... .. .. .. .................................... ............ 14 Table 5: Docketed Chemistry Values for MPS3 ............ ...... ...................................... ........... 17 Table 6: MPS3 1/4T ART Values for 54 EFPY .............. ..... ................................... .... .. .......... 20 Table 7: MPS3 3/4T ART Values for 54 EFPY ............ ........ ........................................ .. .... ... 21 Table 8: MPS3 RT PTs Values for 54 EFPY .. .......................... .. ............. ........... ............... ...... 22 List of Figures Figure 1: Millstone Unit 3 RPV Weld Designations ....... .. ...................................... ................. 6 File No. : 1901402.301 Page 2 of 25 Revision : 0 F0306-01R4 tJ Structural Integrity Associales, Inc. info@structint.com m 1-877-4S1-POWER. structint.com (ffll

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 3 of 30 Figure 2: Millstone Unit 3 Weld Alignment [14] .............. ...... ................................ .. ................ 7 Figure 3: Millstone Unit 3 Reactor Vessels Nozzles and Supports [14] ................................. 8 Figure 4: Millstone Unit 3 Plate Designations from FSAR [11] ..................................... ........ 11 Figure 5: Millstone Unit 3 RPV Rollout Map .................... ......................................... ............ 12 Figure 6: Plate B-9805-2 Initial RT NDT and Chemistry Values ............................. .. .. .... .. ....... 18 Figure 7: Plate B-9805-3 Initial RT NDT and Chemistry Values .................................. ............ 18 Figure 8: Plate B-9820-1 Initial RT NDT and Chemistry Values ..................... ............ .. ............ 19 Figure 9: CMTR Information for Plate 9820-1 ................. .. ................................................... B-2 Figure 10: CMTR Information for B-9820-2 .............. .... ... ... .... ... .... .......... .. ............ ... .. .... ...... B-3 Figure 11: CMTR Information for Plate B-9820-3 ............... ................................................ B-4 File No. : 1901402.301 Page 3 of 25 Revision : 0 F0306-01 R4 SJ Structural Integrity Associates, Inc.* ' info@structint.com m 1-877-4S1-POWER e structint.com @l

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 4 of 30

1.0 INTRODUCTION

Radiation embrittlement of reactor pressure vessel (RPV) materials causes a decrease in fracture toughness. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2 (RG 1.99) [1]

describes general procedures to evaluate the effects of neutron irradiation embrittlement on the alloy steel used in RPVs. To perform this evaluation, RG1 .99 requires calculation of Adjusted Reference Temperature (ART) and Reference Temperature Shift (LlRT Nor) values. The ART values are then used to determine the local fracture toughness of the RPV wall and pressure-temperature limits, according to ASME Code,Section XI, Non-mandatory Appendix G [2] evaluations. (It is noted that the 32 EFPY P-T Limit Curves were assessed to the 1995 Edition of the Code.)

Under the current purchase order 70379609 dated December 22, 2020, Structural Integrity Associates (SI) is performing an update of the Millstone Power Station Unit 3 (MPS3) pressure-temperature (P-T) limit curves for 54 EFPY. The projected fluences for the MPS3 vessel for 54 EFPY are available, and the effects of neutron fluence on embrittlement of the vessel beltline materials (including credible surveillance capsule test results) must be factored into the operating P-T limits for extended plant operation. Calculation of ART and LlRT NDT values have to be developed for MPS3 plates, welds, and nozzles in the RPV beltline.

Calculations have been based on updated fluence calculations provided, including the increase in neutron flux due to a Measurement Uncertainty Recapture (MUR) Uprate. The ART and LlRT NDT values were calculated at 54 effective full power years (EFPY). The reported values for 54 EFPY are intended to be applicable through the end of MPS3's current extended operating period (i.e., 60 years).

The purpose of this calculation is to develop 1/4T and 3/4T ART and LlRT NDT values for each MPS3 RPV ferritic material at the projected fluence levels for 60 years (54 EFPY) with updated fluence values

[13] and incorporation of MPS3 credible surveillance data.

The bounding ART values will be utilized in generation of the 54 EFPY Pressure-Temperature (P-T)

Limit Curves. The current P-T Limit Curves are effective for 32 EFPY [3][4]. The ART values were 124.8°F and 107.0°F for the 1/4T and 3/4T locations, respectively, for input into 32 EFPY curves.

2.0 METHODOLOGY When surveillance data are limited or not available, RG1 .99 [1] specifies that ART is calculated with the following equation:

ART = Initial RTNDT + !iRTNDT + Margin (1)

The "Initial RT Nor" term refers to the reference temperature of nil ductility transition for the non-irradiated material.

The reference temperature shift, LlRT NDT, is defined in RG1 .99 [1] as the shift in the reference temperature resulting from neutron irradiation. LlRT NDT is calculated from the product of the chemistry factor (CF) and fluence factor (FF) as follows:

Af?..1/4DT =CF* FF (2)

The CF is a function of the weight percent copper (Cu) and weight percent nickel (Ni) of the weld and base metal (plate or forging) materials. Tables 1 and 2 of RG1 .99 [1] provide the standard CF values used in this calculation.

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 5 of 30 The FF is based on the accumulated fast neutron exposure (E > 1 MeV) and is typically corrected by the thickness at the location of interest. The FF can be read directly from Figure 1 of RG1 .99, or calculated using the following equation [1] :

FF= f0.2&-0.lOlog(f) (

3)

Due to attenuation effects, the fluence decreases with distance into the RPV wall. The fluence values at 1/4T and 3/4T are utilized from Reference [13] where plant-specific material damage assessment is used to estimate the damage fluence throughout the RPV wall in accordance with the following formula:

f = Jsurf F

• e-0.24X (4)

Where: f = fast neutron fluence (10 19 n/cm 2 , E > 1 MeV) at depth x fsurf = fast neutron fluence at the RPV inside OT surface (i.e., at base metal I cladding interface, same units as f)

X = depth into the RPV wall from the inside surface (inches)

For ASME Code,Section XI, non-mandatory Appendix G [2] evaluations, the "x" value is taken at one-quarter and three-quarters of the base metal thickness (1/4T and 3/4T). The fast neutron fluence can be attenuated through the stainless steel cladding on the inside surface of the RPV. By design, however, the cladding is treated purely as a lining, and not as a load-bearing member. Thus, for the purposes of this evaluation, the inside surface neutron fluence is considered to be at the base metal / cladding interface.

Margin (M), a conservative term defined in RG1 .99 [1], accounts for uncertainty in the initial reference temperature and for variance in LlRTNDT, The margin is calculated using the following formula:

Margin= 2 * .JCY/ + CY,/ (5)

Where: o1 = the standard deviation for the Initial RT Nor (°F)

Of',. = the standard deviation for LlRT NDT (°F)

RG 1.99 [1] states that the standard value of Of',. is 28°F for welds and 17°F for base metal (plates or forgings), and O!',. need not exceed 0.5 times the mean reference temperature shift (0.5* LlRT Nor).

The 01 term, which is related to the uncertainty in the precision of the Initial RT NDT, is applied for values that are determined by measurement and also when generic or default values are used. For MPS3 components where a 01 value is not explicitly identified, 01 is assumed to be equal to 0°F.

When surveillance data exist containing an identical match for the heat number of the vessel beltline material being evaluated, a separate procedure is used to evaluate the ART. This procedure first determines the credibility of the data and, using best estimate chemistry values, calculates a fitted CF.

If the surveillance data is credible, the 0 6 factor of the margin equation (Equation 4 in RG1 .99) may be cut in half, as specified in RG1 .99 [1]. (For the current evaluation, there are credible surveillance data that were utilized.)

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 6 of 30 3.0 DESIGN INPUTS A drawing of the MPS3 reactor vessel showing weld designations is given in Figure 1 [14], and the vessel beltline dimensions are:

• Inside Radius (IR) to clad wetted surface= 86.656 in. [3]

• Clad thickness = 0.156 in. [3]

• Base metal thickness = 8.625 in. [3]

The azimuthal locations of the vessel shell course welds and the inlet and outlet nozzles in the upper shell course are shown in Figure 1, Figure 2, and Figure 3[14][11 ]. The plate designations and weld orientations have been shown in Figure 4.

II 10 r

C 7

C 8

r IIUC.\

I SECTION b-s' i

A W- }!:.~E::f!~:::wrt.gs I j 1- Htl-ffi-- IC,, Q

,311-1r- nn . r .-MI *r" o-t l"fr * .. - l h N

\llf.itQ!llo:ffl'-"--- 4.(VIU\.Q,,91 u:::*c.~ .~~'-:'~~~

fl EACTO'l VESSE'I

.!!2!ll I, ~O"IL.IHC,,Yl~eE!lTf"C:llrJi IU.l'HO"l'Cl:0lll(ICII.L,1,*.

2. u~~~l~~~IJJ..:lfl.lC.l'tol. l Figure 1: Millstone Unit 3 RPV Weld Designations A summary of the MPS3 vessel beltline materials is shown in Table 1 [6]. However, some differences were identified in the reported initial RT NOT values for the beltline plates. The docketed chemistry and initial RT NOT values for the MPS3 vessel beltline plates from the MPS3 License Renewal Application (LRA) are shown in Table 2 [15].

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 7 of 30 Items of note in the LRA table evaluation relative to the current analysis of record ART evaluation follows:

• An interpolation of Charpy testing to find the temperature at which 50 ft-lbs would have been achieved was utilized in the prior ART analysis of record. Testing of Charpy specimens is often performed at 10°F intervals, which provides a conservative estimate for the temperature at which 50 ft-lbs is achieved. The docketed LRA values were utilized for the current ART evaluation. This interpolation has been re-evaluated in the current analysis of record ART evaluation (see Section 5).

• Updated chemistry values have been identified due to rounding in prior calculations and is described in Section 4. *

, ... _, .... .., ....n,1,* . .....

'. . ." . *~~-=

~. ~ -* -. ..-*.....:::* . .......

, OIi

--=+-- --~--:r.-

1 - ... -

. - - : ,_ :: + :*:-

• H

~ *~

• • • ** .- 1

--:~~~ ........... .......,.....*

~:5:$1~~;:::~ iiii~~i

• - ct--~..,._~

1 l *llll*U ** - ll'I.H I *

• flll *

• U

• IOIU-.. U

• *

• n u , M I * *

• ll'fll

:::::::::::: :'.! :: ~, ----*-cAD*-----1

==~~m ~.=~ :~ :: ~~~:~~~:~c:J - I Figure 2: Millstone Unit 3 Weld Alignment [14)

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 8 of 30 r

• n* tc. A 111r*1

• v .. ,. .. .

Ir

'i~ ,;*J., ~? **;~:*~

Figure 3: Millstone Unit 3 Reactor Vessels Nozzles and Supports [14]

Table 1: MPS3 Reactor Vessel Beltline Chemistry and Initial RT NDT Values [6]

Reactor Vessel Material and Cu Ni Initial Identification Number Wt.% Wt.% RTNoT (°F)lbl Intermediate Shell Plate B9805-11' 1 0.05 0.63 60 Intermediate Shell Plate B9805-2 0.05 0.64 10 Intermediate Shell Plate B9805-3 0.05 0.65 0 Lower Shell Plate B9820-1 0.08 0.63 10 Lower Shell Plate B9820-2 0.07 0.6 40 Lower Shell Plate B9820-3 0.06 0.61 20 Intermediate Shell Longitudinal Weld 0.05 0.05 -50 Seams 101-124 A,B,C1'*d)

Intermediate to Lower Shell Girth 0.05 0.05 -50 Weld Seam 101-171(,.d)

Lower Shell Longitudinal Weld 0.05 0.05 -50 Seams 101-142 A,B,C(,.d)

Notes:

(b) Initial RTNDT ('F) are all measured values .

(c) Surveillance Capsule data Is available for Intermediate Shell Plate B9805-1 and all beltline region reactor vessel welds (Intermediate Shell Longitudinal Weld Seams 101-124 A-C , Intermediate to Lower Shell Girth Weld Seam 101-171 and Lower Shell Longitudina l Weld Seams 101-142 A-C).

(d) All beltllne region reactor vessel welds (Intermediate Shell Longitudinal Weld Seams 101-124 A-C, Intermediate to Lower Shell Girth Weld Seam 101-171 and Lower Shell Longitudinal Weld Seams 101-142 A-C) were fabricated with the same w eld wire heat (Heat# 4P6052).

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 9 of 30 Table 2: Docketed Beltline Chemistries and Initial RTNOT Values from Millstone LRA [15]

Table 4.2-2 MIiistone Unit 3

• RTPT& Values at 64 EFPY Chemical Material Description Composition Initial Chemistry Inner Margin &RTPTs RTpra RTNDT Factor Surface Reactor 'F 'F Fluence

'F 'F 'F Vessel Matl. Heat Cu NI E19 n/cm2 Beltllne Type

!dent. Number Wt.'/, Wt,'/,

Region Location Intermediate SA-5338 89805-1 C4039-2 0.05 0.64 80.0 1 31 .02 3.31 34.0 40.7 134.7 Shell Cl.1 Intermediate SA-533B 89805-2 C4068-1 0.05 0.64 6.2 1 3t.o2 3.31 34.0 40.7 80.9 Shell Cl.1 Intermediate SA-5338 89805-3 C4028-1 0.05 0.65 -3.3 1 3t.o2 3.31 34.0 40.7 71 .4 Shell Cl.1 Lower Shell SA-5338 89820-1 88981-1 0.08 0.83 7.0 1 5t.02 3.31 34.0 87.0 108.0 Cl.1 SA-5338 Lower Shell 89820-2 01242-2 0.07 0.60 38.8 44.02 3.31 34.0 57.8 130.6 Cl. 1 SA-5338 Lower Shell 89820-3 01242-1 0.06 0.61 18.6 37.02 3.31 34.0 48.6 101.2 Cl.1 All Welds 4P6052 Linde 0091 0.05 0.05 -50.0 1 31 .72 3.31 58.0 47.7 47.7

1. Measure Value

2. Regulalory Gulde 1.99, Revision 2, Position 1 3.1 32 EFPY ART Tables The adjusted reference temperature (ART) values were calculated at the 1/4T and 3/4T locations through the depth of the vessel for determining the 32 EFPY P-T limit curves. Those ART values were calculated using the best-estimate Cu and Ni values, initial RT NOT values, and projected fluence values in the vessel at each of the locations for the corresponding vessel beltline materials [1 0].

The 32 EFPY P-T limit curves were determined by using the limiting vessel ART values [3] as calculated per the Reg. Guide 1.99, Rev. 2 equations [1]. The highest vessel ART values occur in the plate B-9805-1 (plate heat number C4039-2) with ART values of 124.8°F and 107.0°F respectively for 1/4T and 3/4T.

The official docketed values in the LRA utilize an interpolation to achieve the 50 ft-lb limit from the available data. From the Certified Material Test Reports, the testing had been performed at 10°F intervals. As officially docketed information, the LRA initial RT NOT values have been utilized as inputs to the 32 EFPY P-T Limit Curve Calculation.

For information cited as part of 32 EFPY P-T Limit Curve calculation, basis material document information and calculations were available for review and confirmation [9][1 0].

3.2 Credible Surveillance Data Per Reg. Guide 1.99 Rev. 2 Position 2.1, surveillance data from Capsule Wallows fitting of the capsule test data to reduce the chemistry factor of Intermediate Shell Plate B-9805-1 from 31.0°F to 26.7°F.

When using Regulatory Position 2.1 for credible surveillance data, the margin term, Ot,, may be cut in half. This adjustment due to credible data removes B-9805-1 as the limiting beltline material for MPS3

[12]. The specific data have been shown in Table 3.

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 10 of 30 Table 3: Credible Surveillance Data for Plate B-9805-1 in MPS3 [12]

Table D-t Calculation of Chemistry Factors using MIiistone Unit 3 Surveillance Capsule Data Material Capsule Capsule r*> Ff-(l) ARTNDT(J) FF*ARTr;oT fpl lnlennedinte u 0.400 0.746 2S.9 19.322 O.S57 Shell Plate X 1.98 1.19 25.8 30.612 1.41 8980S-1 (Longit11di11al) w 3.16 1.30 27.8 36.223 1.70 u 0.400 0.746 27.9 20.814 0.5S7 lnlermediale X 1.98 1.19 25.6 30.374 1.41 Shell Pfale B9805-1 w 3.16 1.30 44.6 S8.114 1.70 (Tra11n*erse)

SUM: 19S.46 7.324 CFe9so,.2 =I:(FF. RTNOT) + re FF 2) = (19S.46) + (7.324) =26.7°F Notes:

I. f= flucncc. Units ore x I019 n/cm1 (E > 1.0 MeVJ. Sec Tobie: 6.2-3.

2. FF., fluencc: factor= i<021

• o.l 'lo; fl.

3. ART,.0 r values ore the me11surcd 30 ft-lb shift values. Sec AppendiK C herein, Units arc 1°F].

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 11 of 30

-FOR INFORMATION ONLY -

Revision 33.02-03/31/21 !vIPS-3 FSAR 5.3-26 FIGURE 5.3-1 IDENTITICATION AND LOCATION OF BEL1LINE REGION i\L<\.TERIAL FOR THE REACTOR VESSEL 90' 89805-1 ,,,. 89805-2

f Ill r

01 uJ O' - . - 180' I-

<(

i5 w

(Z'.

,...w

~ 101-124C 101-124B B9805-3 w 270° 0,:

C G

101-171 go*

...-- 101-142A w

I

(/) 180" 0,:

w

§ 101-142C

• ---~ 89820-2 I

270' n-.c- l!lftr:-,J, tcn c, r11Jl'lt::i !n l!l t t ~ tl'o"l t: \'ff':l t n ,; r ln t 3 M m:s1 u d l!'t r t m r.Km Ult t.Dm,.Y.cn V.r.l h tht t- Md:c:17, .,tr: l::.!1 d ~ .t- :A~ . Cvt.-i :t:-c rcrmi:~ Ji-e !nti:rl!cn~ IJTG Ji1:

ltt rt:,,Jt c* ~ t CYt:i t l'll"l;;'lt~ ll ttl t OAR lr..¥.tu'; t f'--C':) t ! l:-ttn t:>,b-,~ '.0 Ul t tlRO:..

Figure 4: Millstone Unit 3 Plate Designations from FSAR [11]

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 12 of 30 Platt B940-4-1 Plate B9404*2 Plate B9404-3 40 103*121 (Heat No. 4PB052) 80 .

Plate = Plate ~

Plate !1z

(/)

B9605-1 (Hut No. C4039*2)

~

~ .,,

z f'

IS B9605*2 (Heat No. C4088*1) z f'

& ,,l!l B9605*3 (Heat No. C4028*1)

~

n

~ t 0,

~ 120 0

(.)

!! ~ tl

~ 101-171 (Heat No. 4P8052)

~ 160

~

200 240 2s1 _ _ _ _ _ _ _ _ _ _ _ _ _ _,___ _.__ _ _ _ _ _ _ _..___ _

0 90 180 270 360 AZIMUTHAL DEGREES Figure 5: Millstone Unit 3 RPV Rollout Map It is noted that the projected 54 EFPY values for fluence noted in the LRA table (Table 4) are conservatively high and were not utilized relative to the updated fluence evaluation docketed for the MUR uprate [13].

The MPS3 vessel is comprised of three shells, joined through vertical and circumferential welding to create continuously fused material. These welds were considered for neutron embrittlement resulting in increased ART values, but the plates sufficiently bound the weld material.

3.3 Fluence Values 54 EFPY surface fluence values for the beltline and extended beltline materials were assessed for the Measurement Uncertainty Recapture (MUR) uprate with 2% power increase that had been analyzed for implementation. The limiting MUR 54 EFPY surface fluence value of 2.72 x 10 19 n/cm 2 was utilized as input (Table 4, 30 degree lower shell plates B-9820-1, B-9820-2, B-9820-3 and lower shell 30 degree long welds 101-142B and 101-142C), as it is the most recently docketed analytical value. Surface fluence values have been provided at different azimuthal locations in the polar direction (Table 4).

Limiting (highest) values have been utilized for the 54 EFPY ART ana lysis for the MPS3 RPV.

The surface fluence values in Table 4 were attenuated using the RG 1.99 [1] equation 3 to determine the 1/4T and 3/4T ART values . The extended beltline materials (e.g., Inlet/Outlet Nozzles & Welds, Nozzle Shell Plates, and Intermediate Shell to Circumferential Welds) were much more limited in fluence (with magnitudes in the 10 17 n/cm 2 range versus 10 19 n/cm2 of the beltline materials) and therefore the fluence had significantly reduced effect on the 54 EFPY ART values relative to the beltline.

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 13 of 30 3.4 Extended Beltline Components beyond the beltline (including nozzles and vessel plates far from the core mid-line) require an assessment to determine whether irradiation embrittlement will have a large effect on the ART and the associated P-T Limit Curves . Nozzles have large stress concentrators at the corners which may be limiting for fracture mechanics analysis.

The Pressurized Water Reactor Owners Group (PWROG) issued a report (PWROG-15109-NP-A) [16]

to address the RPV Nozzles potentially in the Extended Beltline for Appendix G Evaluations. The report provides substantiation such that plant-specific nozzle analysis for evaluation is not required for generation of P-T Limit Curves , with certain criteria.

The maximum fluence for the MPS3 Inlet and Outlet Nozzles (as documented in the MUR submittal

[13]) for 54 EFPY is 1.29 x 10 17 n/cm 2 . The criterion for non-evaluation is a maximum permissible surface fluence of 4.28 x 10 17 n/cm 2 [16J.

Because the fluence on the extended beltline nozzles and welds is below 4.28 x 10 17 n/cm 2 , the MPS3 extended beltline requires no further analysis or evaluation, and is bounded by the results of PWROG-15109-NP-A [16J.

The nozzle shell and welds are bounded by the assessment of limiting beltline material B-9820-2.

Many factors bound the assessment as follows:

• Irradiation embrittlement is limited to ferritic materials exposed to fluence greater than 1x10 17 n/cm 2 (E > 1 MeV), so all plates and welds below this fluence exposure have their initial RT NOT values equivalent to their ART values, so no calculation of shift is necessary.

• The thermal stress intensity factor applicable to the various plate and weld materials is equivalent because the applied heatup and cooldown rates equally affect all components in the cylindrical portion of the RPV.

• The membrane (pressure) stress for other plates in the cylindrical portion of the RPV are equivalent because of consistent thickness of the plate materials and nearly equivalent as-built diameters for the shell segments.

Potential stress concentrators from rapid changes in geometry that are applicable to the nozzle sections are not applicable to the cylindrical shell plates, as the transitions are smooth without sharp points or edges .

Based on the discussion above, beltline material B-9820-2 bounds the material assessment for limiting ART value for the development of 54 EFPY P-T Limit Curves, even with the inclusion and consideration of extended beltline materials. The determination of stress intensity factors (including membrane and thermal stress intensity factors) for this material is bounding for the RPV and RCPB system.

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 14 of 30 Table 4: MPS3 54 EFPY OT Fluence Values [13]

Neutron Fluence (nfcm 2, E >1.0 MeV)

RV Material Azimuth (Deg.) (Beltllne and Extended Beltllne) 54 EFPY 60 EFPY OuUet Nozzles and Nozzle Welds <1 .00E+17 <1.00E+17 Inlet Nozzles and Nozzle Welds <1 .00E+17 <1 .00E+17 Nozzle Shell Plates (B9804-1 ,

5.02E+17 5.63E+17 B9804-2, 89804-3)

Nozzle Shell O Degree Long. Weld 5.02E+17 5.63E+17 (101-122A)

Int. Shell to Nozzle Shell Gire. Weld 5.02E+17 5.63E+17 (103-121)

Int. Shell Plates (B9805-1 , B9805-2, 1.61E+19 1.79E+19 B9805-3) 0 Int. Shell O Degree Long. Weld 1.61E+19 1.79E+19 (101-124A)

Lower Shell to Int. Shell Gire. Weld 1.61E+19 1.79E+19 (101-171)

Lower Shell Plates (B9820-1, 1.64E+19 1.82E+19 B9820-2, B9820-3)

Lower Shell O Degree Long . Weld 1.64E+19 1.82E+19 (101-142A)

Lower Head to Lower Shell Cire.

<1.00E+17 <1.00E+17

• Weld (101-141)

Neutron Fluence (nfcm 2, E >1.0 MeV)

RV Material Azimuth (Deg.) (Beltllne and Extended Beltline) 54 EFPY 60 EFPY OuUet Nozzles and Nozzle Welds <1 .00E+17 <1.00E+17 Inlet Nozzles and Nozzle Welds 1.13E+17 1.26E+17 Nozzle Shell Plates 7.10E+17 7.95E+17 (B9804-1, B9804-2, B9804-3)

Int. Shell to Nozzle Shell Gire. Weld 7.10E+17 7.95E+17 (103-121) 15 Int. Shell Plates 2.32E+19 2.56E+19 (B9805-1, B9805-2, B9805-3)

Lower Shell to Int. Shell Gire. Weld 2.32E+19 2.56E+19 (101-171)

Lower Shell Plates 2.35E+19 2.61E+19 (B9820-1, B9820-2, 89820-3)

Lower Head to Lower Shell Cire.

<1.00E+17 <1.00E+17 Weld (101-141)

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 15 of 30 Outlet Nozzles and Nozzle Welds <1.00E+17 <1.00E+17 Inlet Nozzles and Nozzle Welds 1.16E+17 1.30E+17 Nozzle Shell Plates 7.40E+17 8.27E+17 (B9804-1 , B9804-2, B9804-3)

Int. Shell to Nozzle Shell Circ. Weld 7.40E+17 8.27E+17 (103-121)

Int. Shell Plates 2.54E+19 2.80E+19 45 (B 9805-1 , B9805-2, B9805-3)

Lower Shell to Int. Shell Circ. Weld 2.54E+19 2.80E+19 (101-171)

Lower Shell Plates 2.58E+19 2.85E+19 (B9820-1, B9820-2, B9820-3)

Lower Head to Lower Shell Circ.

<1 .00E+17 <1 .00E+17 Weld (10 1-141)

Neutron Fluence (n/cm 2, E >1.0 MeV)

RV Material Azimuth (Deg.) (Beltllne and Extended Beltllne) 54 EFPY 60 EFPY Outlet Nozzles and Nozzle Welds <1.00E+17 <1.00E+17 Inlet Nozzles and Nozzle Welds 1.29E+17 1.45E+1 7 Nozzle Shell Plates 8.14E+17 9.11 E+17 (B9804-1, B9804-2, B9804-3)

Nozzle Shell 30 Degree Long. Welds 8.14E+17 9.1 1E+17 (101 -1 22B, 101 -1 22C)

Int. Shell to Nozzle Shell Circ. Weld 8.14E+17 9.11E+17 (103-121)

Int. Shell Plates 2.68E+19 2.96E+19 (B9805-1 , 89805-2, 89805-3) 30 Int. Shell 30 Degree Long. Welds 2.68E+19 2.96E+19 (101-1248, 101 -124C)

Lower Shell to Int Shell Circ. Weld 2.68E+19 2.96E+19 (101-171)

Lower Shell Plates 2.72E+19 3.02E+19 (B9820-1, 89820-2, B9820-3)

Lower Shell 30 Degree Long . Welds 2.72E+19 3.02E+19 (101-142B, 101-1 42C)

Lower Head to Lower Shell Circ.

<1 .00E+17 < 1.00E+17 Weld (101-141)

File No.: 1901402.301 Page 15 of 25 Revision: 0 F0306-01R4 lJ Structural Integrity Associates, Inc.* info@structint.com B 1-877-451-POWER '9 slrudint.com ~

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 16 of 30 3.5 Pressurized Thermal Shock PWR RPVs must ensure protection against Pressurized Thermal Shock (PTS), as defined by 10 CFR 50.61. A pressurized thermal shock event is an event causing severe overcooling with or followed by significant pressure in the RPV. This shock has the potential to induce non-ductile failure of the material by means of overcoming the critical fracture toughness criterion while at a sufficiently low temperature to enable brittle fracture of low alloy steel.

The potential for this phenomenon must be addressed and resolved. The criteria for passing this evaluation is an RT Prs of 270°F for plates, forgings, and weld material, and 300°F for circumferential welds . The pressurized thermal shock assessment has been made presently without need to assess through the alternate requirements of 10 CFR 50.61a.

File No.: 1901402.301 Page 16 of 25 Revision: 0 F0306-01R4

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 17 of 30 4.0 ASSUMPTIONS The assumptions made to define the evaluation approach and perform the analysis are summarized in the following list.

1. According to RG1 .99, the 01 term is equal to the standard deviation of the Initial RTNoT when that quantity is estimated from physical measurements [1]. Values calculated by this method include substantial conservatism, rendering it unnecessary to create additional conservatism via the 01 term. Consequently, for MPS3 ART calculations, 01, is set equal to zero.

2. Because of rounding at an early stage of calculation, the values provided in Table 5 are conservatively high for the purposes of ART calculation . The CMTRs for the following plate materials have been included in Appendix Bas Figure 9, Figure 10, and Figure 11, respectively.

a. For plate B-9820-1, rather than the values of 0.08% and 0.63% for the copper and nickel, respectively, the un-rounded values of 0.075% and 0.625% were utilized [9].

b. For plate B-9820-2, rather than the values of 0.07% and 0.60% for the copper and nickel, respectively, the un-rounded values of 0.065% and 0.60% were utilized [9].

c. For plate B-9820-3, rather than the values of 0.06% and 0.61 % for the copper and nickel, respectively, the un-rounded values of 0.060% and 0.605% were utilized [9].

Table 5: Docketed Chemistry Values for MPS3

.llllll*!on* 3 Chaml*!!l Data I -

Cu NI .Rafare nca Ma ncu Ma nNI

ce NPSD-1038 Final Report - -

• P1oa2

- 0.047 0.049

-dated June 1987

-0.0&

Beat E1Um1ta Data . 0.0&

B110&* 1 , 0.050 0.670 .WCAP-10732 (pg A*2) o.Q<la t 0.620 :WCAP-10732 (pg A-2) 0.060 I 0.620 l,1C_R _ _

0.0(! _ _O_,M__

0.05 1 0,620 .WCAP-10732 (pg A,2)

B1105*2

- 0.050 . 0.650 .MCR 0.08 o.,. -

B9805-3 0,040 o.oeo

~

0.620 0.880

.WCAP-107~2 ll!!I A,2)

MCR

+

I 0.05 .

a.es BH20,1 j 0.070 1

0 .620 WCAP-10732 (pgA-2)

- - t

• B8820-2 t 0.080 0.060

. 0.630 fMCR 0.600 !wcAP-10732 (pg A-2)

I O.OI 0 ,IS

- 0.070 0.800 jMCR - - - - - 0.07 0.10 B9820..:1 0.050 0.580 !wcAP-10132 CP9!-*2l -

0.070 0.830 IMCR -+ 0.01 ' o.e1 File No.: 1901402.301 Page 17 of 25 Revision : 0 F0306-01R4 SJ Structural Integrity m G Associales, lnc.e info@structint.com 1-877-4S1-POWER. slructinl.com

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 18 of 30 5.0 CALCULATIONS The methodology in Section 2.0 is used to evaluate the ART and LlRT NDT values for MPS3, based on the design inputs in Section 3.0 and consistent with the assumptions in Section 4.0 . The calculations for the 54 EFPY 1/4T and 3/4T ART have been performed in the Excel spreadsheet included as a supporting file in Appendix A. The design inputs, and resultant 1/4T and 3/4T ART values are given in Table 6 and Table 7 respectively for 54 EFPY.

The bounding 54 EFPY 1/4T ART and 3/4T values for the RPV plate B-9820-2 are 118.7°F and 107.0°F, respectively. The utilization of credible surveillance data for plate B-9805-1 (the previously 32 EFPY limiting material) reduces the ART and removes the plate as a limiting location.

5.1 RTNDT Values The interpolation utilized for the RT NDT values was confirmed to provide the initial values utilized for the 32 EFPY values for plates B-9805-2, B-9805-3, and B-9820-1, for interpolated values of 6.2°F, -3.3°F, and 7.0°F, respectively, as shown in .

Millstone 3 Plate B-9805-2, Heat C4068-1 Calculate Chcmisuy Factor IRG I 99 rev 2l Cu:= o.os Ni :* 0.64 Mean Values (see Table 8)

CF :=31 Dctennination of Initial RTndt (NB233I)

Tnc1, := --60*°F Unirradiatod Drop Weight Te*ting Dal a (C6 MCR)

Tc, := 66.IS*°F Interpolation between SS@70'F and 42@60°F Tcv-60*°F=6.2°P RTndl := 6.2* "F Figure 6: Plate B-9805-2 Initial RTNDT and Chemistry Values MHlstone 3 Plate B-9805-3, Heat C4028-t Calculate Qlcmjstry Factor CRG I 99 rev 2}

Cu := 0.05 Ni:= 0.65 Mean Value., (See Table 8)

CF:= 31 Determjnatjon or Initial RTndt /NQ2331}

T ndl := "F Uoirradiated Drop Weight Testin11 Data (CE MCR)

Tcv := 56.7*°F Interpolation between 48@50°F and 51@60°F Tcv - 60*°F = -3.3 °F RTndt := -3.3*°F Figure 7: Plate B-9805-3 Initial RTNDT and Chemistry Values File No. : 1901402.301 Page 18 of 25 Revision: 0 F0306-01 R4 SJ Structural Integrity Associates, Inc.~ info@structint.com m 1-877-4SI-POWER. struclint.com @)

Serial No.22-361 Docket No . 50-423 Attachment 3, Page 19 of 30 Millstone 3 Piute B-9820-1, Heat B8961-1 Cu := 0.08 Ni := 0.63 Mean Values (See Table 8)

CF :=51 Petenniu*liou oflulti*I RTndt INB2331)

Tndt := -50*°F Unirradiated Drop Weight Testing Data (CE MCR)

T.- := 67.0*°F Interpolation between 53@70'F and 43@60"F T0 v-60*'F = 7.0 'F RTndt := 7.0- 'F Figure 8: Plate B-9820-1 Initial RTNor and Chemistry Values 5.2 Pressurized Thermal Shock RT PTs is calculated in a similar manner as ART except that the surface fluence value is used and the requirements are prescribed in 10 CFR 50.61. The calculations for RT PTs are shown in Table 8. The bounding RT PTs value is 124.1 °F in 8-9820-2, which satisfies the 270°F criterion defined in 10 CFR 50.61 [20] .

File No. : 1901402.301 Page 19 of 25 Revision : 0 F0306-01R4 13 Structural Integrity Associates, Inc.~ info@structinl.com m 1-877-451-POWER C9 structint.com (ffl)

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 20 of 30 Table 6: MPS3 1/4T ART Values for 54 EFPY Initial 54 EFPY Component & 54 EFPY 1/4T Fluence ~RTNor Margin Shift

%Cu %Ni CF RT NOT O"; O"/l 1/4T ART ID Fluence (n/cm 2 ) Factor (OF) (OF) (OF)

(OF) (OF)

Intermediate Shell Plate B-9805-1 (1.1) 0.052 0.637 32.2 60.0 1.60 X 1019 1.129 36.36 0 17.0 34.0 70.4 130.4 19 B-9805-1 (2.1) 0.052 0.637 26.7 60.0 1.60 X 10 1.129 30.15 0 7.54 15.1 45.2 105.2 19 B-9805-2 0.050 0.635 31.0 6.2 1.60 X 10 1.129 35.01 0 17.0 34.0 69 .0 75.2 B-9805-3 0.050 0.650 31.0 -3.3 1.60 X 10 19 1.129 35.01 0 17.0 34.0 69 .0 65 .7 Lower Shell Plate B-9820-1 0.075 0.625 47.5 7.0 1.62 X 10 19 1.133 53.83 0 17.0 34.0 87.8 94.8 19 B-9820-2 0.065 0.600 40.5 38.8 1.62 X 10 1.133 45.90 0 17.0 34.0 79.9 118.7 19 B-9820-3 0.060 0.605 37.0 18.6 1.62 X 10 1.133 41 .93 0 17.0 34.0 75.9 94.5 Welds 101-124 A, B, C 0.047 0.049 31.7(b) -50.0 1.62 X 10 19 1.133 35.93 0 18.0 35.9 71.9 21.9 Notes:

(a) Regulatory positions 1.1 and 2.1 have both been evaluated, although the values associated w ith 2.1 are applicable to MPS3.

(b) Value utilized from Reference [9], which is conservative relative to calculated Chemistry Factor of 30.7°F.

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Serial No .22-361 Docket No. 50-423 Attachment 3, Page 21 of 30 Table 7: MPS3 3/4T ART Values for 54 EFPY Initial 54 EFPY Component & 54 EFPY 3/4T Fluence 8RTNDT Margin Shift

%Cu %Ni CF RTNDT a; Or,. 3/4T ART ID Fluence (n/cm 2) Factor (OF) (OF) (OF)

(OF) (OF)

Intermediate Shell Plate B-9805-1 (1.1) 0.052 0.637 32.2 60.0 5.67 X 10 18 0.841 22.47 0 13.5 27.1 54.2 114.2 B-9805-1 (2.1) 0.052 0.637 26.7 60.0 5.67 X 10 18 0.841 22.47 0 5.62 11.2 33.7 93.7 B-9805-2 0.050 0.635 31.0 6.2 5.67 X 10 18 0.841 26.09 0 13.0 26.1 52.2 58.4 18 B-9805-3 0.050 0.65 31.0 -3.3 5.67 X 10 0.841 26.09 0 13.0 26.1 52.2 48.9 Lower Shell Plate B-9820-1 0.075 0.625 47.5 7.0 5.76 X 10 18 0.846 40.16 0 17.0 34.0 74.2 81.2 18 B-9820-2 0.065 0.600 40.5 38.8 5.76 X 10 0.846 34.25 0 17.0 34.0 68.2 107.0 18 B-9820-3 0.060 0.605 37.0 18.6 5.76 X 10 0.846 31.29 0 15.6 31.3 62.6 81.2 Welds 101-124 A, B, C 0.047 0.049 31.7(b) -50.0 5.76 X 1018 0.846 26.80 0 13.4 26.8 53.6 3.6 Notes:

(a) Regulatory positions 1.1 and 2.1 have both been evaluated, although the values associated with 2.1 are applicable to MPS3.

(b) Value utilized from Reference [9], which is conservative relative to calculated Chemistry Factor of 30.7°F File No.: 1901402.301 Page 21 of 25 Revision: 0 F0306-01R4 si,,o1,,,1 l'1egrify Associates, Inc. i nfo@structi nt.com m 1-877-45!-POWER e structi nt.com @l

Serial No.22-361 Docket No. 50-423 Attachment 3, Page 22 of 30 Table 8: MPS3 RTPTS Values for 54 EFPY 54 EFPY ID Fluence ~RTNDT 54 EFPY Initial Margin Shift Component & ID %Cu %Ni CF Fluence O; 04 (OF) RTpts RTNDT Factor (OF) (OF)

(n/cm 2

)

(OF)

Intermediate Shell Plate B-9805-1 (1.1) 0.052 0.637 32.2 60.0 2.68 X 10 19 1.263 40.682 0 17 34 70.4 134.7 B-9805-1 (2.1) 0.052 0.637 26.7 60.0 2.68 X 10 19 1.263 33.733 0 8.4 16.9 45.2 110.6 B-9805-2 0.050 0.635 31.0 6.2 2.68 X 10 19 1.263 39.166 0 17 34 69.0 79.4 B-9805-3 0.050 0.650 31.0 -3.3 2.68 X 10 19 1.263 39.166 0 17 34 69.0 69.9 Lower Shell Plate B-9820-1 0.075 0.625 47 .5 7.0 2.72 X 10 19 1.267 60.185 0 17 34 87.8 101.2 B-9820-2 0.065 0.600 40.5 38.8 2.72 X 10 19 1.267 51.316 0 17 34 79.9 124.1 B-9820-3 0.060 0.605 37.0 18.6 2.72 X 10 19 1.267 46.881 0 17 34 75.9 99.5 Welds 101-124 A, B, C 0.047 0.049 31.7(b) -50.0 2.72 X 10 19 1.267 40.1655 0 20.1 40.2 71.9 30.3 Notes:

(a) Regulatory positions 1.1 and 2.1 have both been evaluated, although the values associated with 2.1 are applicable to MPS3.

(b) Value utilized from Reference [9], which is conservative relative to calculated Chemistry Factor of 30.7°F File No.: 1901402.301 Page 22 of 25 Revision : 0 F0306-01R4

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 23 of 30

6.0 CONCLUSION

S SI has gathered and reviewed the available unirradiated measured test data, best-estimate chemistries and initial RT NDT values, fluence values, and the corresponding docketed and reported values that were used to establish the existing 32 EFPY P-T limit curves. From this review, the basis for existing 32 EFPY P-T limits was the projected ART value at the 1/4T and 3/4T locations for the limiting intermediate shell plate B-9805-1. The values were 124.8°F and 107.0°F for the 1/4T and 3/4T locations, respectively.

The additional design input information provided to SI contains more of the measured chemistries and test data than were used to establish initial RT NDT values for the current beltline materials and for the extended beltline materials, including the plates and welds. Fluence projections for 54 EFPY were provided, the surveillance capsule data for the matching vessel shell plate B-9805-1 was reviewed for credibility, and the resulting projections of the 54 EFPY ART values were calculated.

The bounding 54 EFPY 1/4T ART and 3/4T values for the RPV plate B-9820-2 are 118.7°F and 107.0°F, respectively. The utilization of credible surveillance data for plate B9805-1 (the previously 32 EFPY limiting material) reduces the ART and removes plate B-9805-1 as a limiting location.

Because of changes in the limiting beltline plate due to the credit for the surveillance data, the predicted ART value at the 3/4T location for the new limiting material (lower shell plate B9820-2) for 54 EFPY is equivalent to the value used for the P-T limit curves for 32 EFPY. This information will flow downstream to the P-T Curve calculation.

The bounding RT PTs value is 124.1 °F in B-9820-2, which satisfies the 270°F criterion defined in 10 CFR 50.61 [20].

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 24 of 30

7.0 REFERENCES

1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1995 Edition.

3. Dominion Energy Calculation, M3-LOE-00284EM, Rev. 5, "Millstone Unit 3:

Pressure/Temperature Limits for 32 EFPY", June 29, 2001, SI File Number 1901402.201.

4. Millstone Nuclear Power Station, Unit No. 3, Technical Specifications Change Request 3 11 00, Reactor Coolant System Heatup and Cooldown Curves {ADAMS Accession No. ML011210201).

5. ASME Boiler & Pressure Vessel Code,Section XI, 1995 Edition, Rules for lnservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection Against Failure."

6. Millstone Unit 3 Plant Technical Specifications, Issuance of Amendment Re: Reactor Coolant System Heatup and Cooldown Curves {TAC No. 1785)(ADAMS Accession No. ML012060343).

7. Cold Overpressure Protection System for Measurement Uncertainty Recapture (including Fluence Values), Engineering Technical Evaluation, Document No. ETE-MP-2017-1162, Rev. 0, March 8, 2018, SI File Number 1901402.201.

8. Millstone 3: PORV Setpoint Curves for the Cold Overpressure System for 32 EFPY, Calculation No. 94-ENG-01042C3, Rev. 05, Change No. 01, August 28, 2008, SI File Number 1901402.201.

9. Calculation of Initial Properties for the MP2 and MPS3 Reactor Vessels, Calculation No. 95-SDS-1007MG, Rev. 7, May 4, 2005, SI File Number 1901402.201.

10. Calculation of Adjusted Reference Temperatures for the MP2 and MP3 Reactor Vessels, Calculation No. 95-SDS-1008MG, Rev. 5, May 4, 2005, SI File Number 1901402.201 .

11. MPS3 FSAR Chapter 5, Figure 5.3-1, Identification and Location of Beltline Region Material for the Reactor Vessel, RV Beltline Rollout, Revision 33.02 - March 31, 2021, SI File Number 1901402.201.

12. WCAP-16629-NP, Revision 0, Analysis of Capsule W from the Dominion Nuclear Connecticut Millstone Unit 3 Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Corp.,

September 2006.

13. Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station Unit 3 Proposed License Amendment Request Measurement Uncertainty Recapture Power Uprate, Submittal to U.S .

Nuclear Regulatory Commission, November 19, 2020 (ADAMS Accession No. ML20324A703).

14. Millstone Nuclear Power Station Unit 3 Reactor Vessel - Zone 001 Weld Designations, Vessel Drawing No. 25212- 20900, Sheet 1 of 3 and Sheet 2 of 3, March 8, 1983, SI File Number 1901402.201.

File No.: 1901402.301 Page 24 of 25 Revision: 0 F0306-01 R4

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 25 of 30

15. Millstone Power Station Unit 3, Application for Renewed Operating License, Technical and Administrative Information through Appendix D, January 20, 2004, (ADAMS Accession No. ML040260103, ML040400289) .

16. Pressurized Water Reactor Owners Group Report, PWROG-15109-NP-A, Rev. 0, "PWR Pressure Vessel Nozzle Appendix G Evaluation", January 2020, (ADAMS Accession Number ML20024E573).

17. NRC Technical Report Letter, TLR-RES/DE/CIB-2013-01, "Evaluation of the Beltline Region for Nuclear Reactor Pressure Vessels", November 14, 2014, (ADAMS Accession Number ML14318A177).

18. Dominion Document, ETE-MP-2017-1162, Rev. 0, March 8, 2018, SI File Number 1901402.201.

19. E-mail from T. Steahr (Dominion) to D. Denis (SI), "1901402.301.RB-ART Cale", September 7, 2022, SI File Number 1901402.201.

20. United States Code of Federal Regulations, 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressured Thermal Shock Events", Updated November 3, 2021.

File No.: 1901402.301 Page 25 of 25 Revision : 0 F0306-01R4

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 26 of 30 APPENDIX A SUPPORTING FILES Date Comment 1.1901402.301.R0.xlsx 10/21/2022 Excel file contains the detailed ART calculations File No.: 1901402.301 Page A-1 of A-1 Revision: 0 F0306-01R4 I)

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 27 of 30 APPENDIX B CMTR INFORMATION [19]

File No.: 1901402.301 Page 8-1 of 8-4 Revision : 0 F0306-01 R4 SJ Structural Integrity Associates, Inc.* info@structinl.com m 1-877-4S1-POWER G structint.com @)

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Figure 9: CMTR Information for Plate 9820-1 File No.: 1901402.301 Revision : 0 Page B-2 of B-4 F0306-01R4 tr Structural Integrity Associates, Inc.

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Serial No.22-361 Docket No. 50-423 JUN 15 *se 15I4e FROM ~LUIO SYSTMS EN~N~~N~

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r-t Figure 10: CMTR Information for 8-9820-2 File No. : 1901402.301 Page B-3 of B-4 Revision : 0 F0306-01R4 Slm,Jt]

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Serial No.22-361 Docket No. 50-423 Attachment 3, Page 30 of 30

!UH 15 '92 15143 FROM FLUID SYSTMS EHGNERNG PAGE.0137

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Figure 11: CMTR Information for Plate 8-9820-3 File No.: 1901402.301 Page B-4 of B-4 Revision: 0 F0306-01R4 e

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Serial No.22-361 Docket No. 50-423 Attachment 4 PROJECTED UPPER SHELF ENERGY VALUES FOR 54 EFPY Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

Serial No.22-361 Docket No. 50-423 Attachment 4, Page 1 of 1 Projected Upper Shelf Energy Values for 54 EFPY