ML21033B007

Issuance of Amendment No, 183 to Correct Non-Conservative Technical Specification Related to Reactor Coolant System Pressure/Temperature Limits

ML21033B007
Person / Time
Site:	Harris  (NPF-063)
Issue date:	03/22/2021
From:	Michael Mahoney Plant Licensing Branch II
To:	Maza K Duke Energy Progress
Mahoney M, NRR/DORL/LPL2-2, 415-3867
References
EPID L 2020 LLA 0111
Download: ML21033B007 (20)

Text

March 22, 2021 Ms. Kim Maza Site Vice President Shearon Harris Nuclear Power Plant Mail Code NHP01 5413 Shearon Harris Road New Hill, NC 27562-9300

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 183 TO CORRECT NON-CONSERVATIVE TECHNICAL SPECIFICATION RELATED TO REACTOR COOLANT SYSTEM PRESSURE/TEMPERATURE LIMITS (EPID L-2020-LLA-0111)

Dear Ms. Maza:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 183 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). This amendment is in response to your application dated May 12, 2020, as supplemented by letter dated October 1, 2020.

The amendment revises Technical Specification (TS) 3/4.4.9, Pressure/Temperature Limits -

Reactor Coolant System, to reflect an update to the pressure and temperature limit curves in TS Figures 3.4-2, Reactor Coolant System Cooldown Limitations and 3.4-3, Reactor Coolant System Heatup Limitations. The amendment also reflects that the revised pressure and temperature limit curves in TS Figures 3.4-2 and 3.4-3 will be applicable until 55 effective full power years (EFPY) and revises TS Figure 3.4-4, Maximum Allowed [Power Operated Relief Valve] PORV Setpoint for the Low Temperature Overpressure Protection System, to reflect that the setpoint values are based on 55 EFPY reactor vessel data.

A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions regular monthly Federal Register notice.

Sincerely,

/RA/

Michael Mahoney, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. Amendment No. 183 to NPF-63

2. Safety Evaluation cc: Listserv

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 183 Renewed License No. NPF-63

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated May 12, 2020, as supplemented by letter dated October 1, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-63 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 183, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility License No. NPF-63 and Technical Specifications Date of Issuance: March 22, 2021 Undine S.

Shoop Digitally signed by Undine S. Shoop Date: 2021.03.22 14:22:27 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 183 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following page of the Renewed Facility Operating License with the revised page.

The revised page is identified by amendment number and contains a marginal line indicating the area of change:

Remove Insert Page 4 Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change:

Remove Insert 3/4 4-35 3/4 4-35 3/4 4-36 3/4 4-36 3/4 4-41 3/4 4-41 Renewed License No. NPF-63 Amendment No. 183 C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1)

Maximum Power Level Duke Energy Progress, LLC, is authorized to operate the facility at reactor Core power levels not in excess of 2948 megawatts thermal (100 percent rated core power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 183, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Antitrust Conditions Duke Energy Progress, LLC. shall comply with the antitrust conditions delineated in Appendix C to this license.

(4)

Initial Startup Test Program (Section 14)1 Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

(5)

Steam Generator Tube Rupture (Section 15.6.3)

Prior to startup following the first refueling outage, Carolina Power & Light Company* shall submit for NRC review and receive approval if a steam generator tube rupture analysis, including the assumed operator actions, which demonstrates that the consequences of the design basis steam generator tube rupture event for the Shearon Harris Nuclear Power Plant are less than the acceptance criteria specified in the Standard Review Plan, NUREG-0800, at 15.6.3 Subparts II (1) and (2) for calculated doses from radiological releases. In preparing their analysis Carolina Power &

Light Company* will not assume that operators will complete corrective actions within the first thirty minutes after a steam generator tube rupture.

1The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

• On April 29, 2013, the name of Carolina Power & Light Company (CP&L) was changed to Duke Energy Progress, Inc. On August 1, 2015, the name Duke Energy Progress, Inc. was changed to Duke Energy Progress, LLC.

FIGURE 3.4-2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 55 EFPY SHEARON HARRIS - UNIT 1 3/4 4-35 Amendment No. 183 2400 -

MATERIAL PROPERTY BASES:

Controlling Material -

Plate 84197-2 Copper Content -

0.09%

2200 -

Nickel Content -

0.50%

Regulatory Guide -

1.99 Rev. 2 RT NOT lnitial-91°F RT NOT at 1/4 T -

212°F RT NOT at 3/4 T -

198°F 2000 -

Pressure-Temperature limits have NOT been adjusted for instrument errors. Tilese errors are controlled by the Technical Specification Equipment list Program, Plant 1800 1600

~

ti)

C.

~ 1400 0::

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ti) ti) w 1200 0::

C.

C w

~ 1000 u

C z 800 600 400 200 0

50 Procedure PLP-106.

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~II' i....-

-~

100 150 j

I

~o FJH, I

I I

I I

/

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200 IS H I

j I ' I I

I J

I 250 300 350 INDICATED TEMPERATURE - °F 400 450 500

FIGURE 3.4-3 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 55 EFPY SHEARON HARRIS - UNIT 1 3/4 4-36 Amendment No. 183 2400 2200 2000 1800

~ 1600 ti)

a.

I

~1400 ti) ti) w o::: 1200

a.

C w 1-51000 0 z 800 MATERIAL PROPERTY BASES:

Controlling Material -

Plate 84197-2 Copper Content -

0.09%

Nickel Content -

0.500/4 Regulatcxy Guide-1.99 Rev. 2 RT NOT Initial -

91 'F RT NOT at 1/4 T-212"F RT NOT at 3/4 T -

198' F Pressure-Temperature limits have NOT been adjusted for instrument errors.

These errors are controlled by the Technical Specification Equipment List Program, Plant Procedure PLP-106.

H Above 350'F, pressure limit exceeds I

2485 psig setpoint of pressurizer safety relief valves (Specification 3.42.1 ). I 50°F/HR

/

/

./

V

/

600 400 200 0

50 100

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150 200 I

I I

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ISLH /

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I J

I I

I I

I J

I I

I I

V I

I 250 300 INDICATED TEMPERATURE- °F 350 400

RCS TEMP (°F)

LOW PORV* (psig)

HIGH PORV* (psig) 90 400 410 250 400 410 325 440 450

• VALUES BASED ON 55 EFPY REACTOR VESSEL DATA INSTRUMENT ERRORS ARE CONTROLLED BY THE TECHNICAL SPECIFICATION EQUIPMENT LIST PROGRAM, PLANT PROCEDURE PLP-106.

FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM SHEARON HARRIS - UNIT 1 3/4 4-41 Amendment No. 183 300 400 500 0

100 200 300 400 PORV SETPOINT (PSIG)

MEASURED RCS TEMPERATURE (°F)

LOW PORV HIGH PORV

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 183 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By application dated May 12, 2020 (Reference 1), as supplemented by letter dated October 1, 2020 (Reference 2), Duke Energy Progress, LLC (the licensee), requested changes to the technical specifications (TSs) for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris or HNP). The license amendment request (LAR) proposes changes to Technical Specification (TS) 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System, to reflect an update to the pressure and temperature limit curves in TS Figures 3.4-2, Reactor Coolant System Cooldown Limitations and 3.4-3, Reactor Coolant System Heatup Limitations.

Additionally, the licensees proposed change reflects that the revised pressure and temperature limit curves in TS Figures 3.4-2 and 3.4-3 will be applicable until 55 effective full power years (EFPY). Lastly, the proposed change revises TS Figure 3.4-4, Maximum Allowed [Power Operated Relief Valve] PORV Setpoint for the Low Temperature Overpressure Protection System, to reflect that the setpoint values are based on 55 EFPY reactor vessel data.

The licensee stated that the purpose for the proposed revisions is because results of the testing of surveillance capsule Z, documented in Framatome, Inc. report ANP-3798NP, Analysis of Capsule Z, Duke Energy Shearon Harris Nuclear Power Plant, Reactor Vessel Material Surveillance Program, Revision 0 (Reference 3), revealed that the existing Pressure/Temperature (P/T) limits are non-conservative. The licensee also stated that operation within the P/T limits are maintained through administrative controls consistent with U.S. Nuclear Regulatory Commission (NRC) Administrative Letter 98-10, Dispositioning of Technical Specifications That Are Insufficient to Assure Pant Safety, (Reference 4).

The supplement, dated October 1, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs initial proposed no significant hazards consideration determination, as published in the Federal Register on August 25, 2020 (85 FR 52370).

2.0 REGULATORY EVALUATION

2.1 System Descriptions In its letter dated May 12, 2020, Section 2.1 System Design and Operation, the licensee states, in part:

All components of the HNP 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. HNP is required to limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation.

The HNP TS contain 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 typical use of the curves is operational guidance 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.

Operating limits are established that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).

2.2 Description of Changes The licensee proposed to revise HNP Technical Specifications, as follows:

Current TS Figure 3.4-2, Reactor Coolant System Cooldown Limitations - Applicable to Up to 36 EFPY, is revised as follows:

The existing RCS cooldown limitations curves are superseded entirely by new curves applicable up to 55 EFPY.

The cooldown limitations curve for a rate of 100 °F/HR [degrees Fahrenheit per hour] is removed.

The RTNDT [reference nil-ductility temperature] at 1/4 T [location] value of 191°F is revised to 212°F in the Material Property Bases.

The RTNDT at 3/4 T [location] value of 179°F is revised to 198°F in the Material Property Bases.

The title of Figure 3.4-2 is revised to state Reactor Coolant System Cooldown Limitations - Applicable Up to 55 EFPY.

Current TS Figure 3.4-3, Reactor Coolant System Heatup Limitations - Applicable Up to 36 EFPY, is revised as follows:

The existing RCS heatup limitations curves are superseded entirely by new curves applicable up to 55 EFPY.

The RTNDT at 1/4 T [location] value of 191°F is revised to 212°F in the Material Property Bases.

The RTNDT at 3/4 T [location] value of 179°F is revised to 198°F in the Material Property Bases.

The title of Figure 3.4-3 is revised to state Reactor Coolant System Heatup Limitations -

Applicable Up to 55 EFPY.

Current TS Figure 3.4-4, Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System, is revised as follows:

The current note *VALUES BASED ON 36 EFPY REACTOR VESSEL DATA is revised to state *VALUES BASED ON 55 EFPY REACTOR VESSEL DATA.

2.3 Applicable Regulatory Requirements and Guidance Regulations The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. The NRC staff evaluates the acceptability of a facilitys proposed P/T limits based on the following NRC regulations and guidance:

The regulation in 10 CFR 50.36, Technical Specifications, requires that TSs include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

The regulation in 10 CFR 50.36(c)(2), Limiting conditions for operation, states that [l]imiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

10 CFR 50.60, Acceptance criteria for fracture prevention measures for light water nuclear power reactors for normal operation, imposes fracture toughness and material embrittlement surveillance program requirements set forth in Appendices G and H to 10 CFR Part 50.

Appendix G, Fracture Toughness Requirements, to 10 CFR Part 50 requires, in part, that facility P/T limits for the Reactor Pressure Vessel (RPV) be at least as conservative as those obtained by applying the linear elastic fracture mechanics (LEFM) methodology of Appendix G, Fracture Toughness Criteria for Protection Against Failure, to Section XI of the American Society for Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

Appendix H, Reactor Vessel Material Surveillance Program Requirements, to 10 CFR Part 50 establishes requirements for a facilitys surveillance program for monitoring RPV embrittlement due to neutron irradiation.

Regulatory Guidance Generic Letter (GL) 92-01, Reactor Vessel Structural Integrity, Revision 1 and GL 92-01, Supplement 1 (Reference 5).

Section 5.3.2, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP) (Reference 6).

Regulatory Guide (RG) 1.99, Radiation Embrittlement of Reactor Vessel Materials; Revision 2, contains guidance for RPV embrittlement integrity evaluations (Reference 7).

RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (Reference 8).

Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, (Reference 9).

3.0 TECHNICAL EVALUATION

The licensee developed P/T limits for HNP, Unit 1, applicable to 55 EFPY, which are proposed to replace the P/T limits in current TS Figures 3.4-2 and 3.4-3 in TS 3/4.4.9, as shown in of the licensees May 12, 2020, application. The licensee stated that the purpose for the proposed 55 EFPY P/T limits is because results of the testing of surveillance capsule Z, documented in Framatome report ANP-3798NP, showed that the current limiting adjusted reference temperature (ART) values for 36 EFPY increased. Framatome report ANP-3798NP provided results of the examination of capsule Z as part of the licensees Reactor Vessel Surveillance Program, consistent with 10 CFR 50, Appendix H requirements. Details of the development of the 55 EFPY P/T limits are in Section 3 of the enclosure to the licensees May 12, 2020, letter. The licensee stated that the proposed 55 EFPY P/T limits are based on the plane strain fracture toughness (KIC) methodology from Appendix G to Section XI of the ASME Code and are consistent with the requirements of 10 CFR Part 50, Appendix G.

The licensee also evaluated the current temperature enable and PORV pressure settings of the low-temperature overpressure protection (LTOP) system to demonstrate that the settings remain acceptable up to 55 EFPY.

3.1 Adjusted Reference Temperature Calculations Section 1.1.5 of the HNP, Unit 1, Updated Final Safety Analysis Report (UFSAR) (Reference 10), states that the unit has undergone a Measurement Uncertainty Recapture (MUR) power uprate in 2012 that increased the reactor core thermal output from 2900 MWt to 2948 MWt. The licensee did not state in the May 12, 2021, letter if the ART calculations included the effect of the MUR uprate in 2012. In response to the NRC staff request for additional information (Reference 11), the licensee confirmed that the effect of the 2012 MUR had been included in the proposed 55 EFPY P/T limits and referred to Chapter 1.1.5 of the HNP, Unit 1, UFSAR to explain that no other uprates have occurred since the 2012 MUR. The NRC staff finds the licensees response acceptable.

The last two rows of Table 3 of the enclosure to the licensees May 12, 2020, letter list the fluence values at the [one-quarter thickness] 1/4T Maximum Peak Location (4.04E+19

[neutrons per square centimeter] n/cm2) and [three-quarter thickness] 3/4T at Maximum Peak Location (9.99E+18 n/cm2), but do not state which component these represent. In response to the NRC staffs request for additional information, the licensee clarified that these two fluence values were for the intermediate shell plate, as indicated and noted in Table 6-3 of ANP-3798NP. The NRC staffs calculated fluence values at the 1/4T and 3/4T locations for the intermediate shell plate heat plate, number B4197-2 are 4.32E+19 n/cm2 and 1.70E+19 n/cm2, respectively, based on the thickness of 7.75 inches given in Table 1 of the enclosure to the licensees May 12, 2020, letter. The NRC staff verified that the limiting ART values in Table 1 of the enclosure to the licensees May 12, 2020, letter are based on the fluence values of 4.32E+19 n/cm2 and 1.70E+19 n/cm2 for the intermediate shell slate heat plate, number B4197-2, not on the slightly lower fluence values in Table 6-3 of ANP-3798NP. Therefore, the NRC staff finds the licensees clarification acceptable.

The licensee considered RPV materials within and outside of the traditional beltline expected to receive a neutron fluence value greater than 1E+17 n/cm2, as listed on page 5 of the enclosure to the licensees May 12, 2020, letter. The licensee evaluated the ART values for these materials and determined that intermediate shell plate, heat number B4197-2 remains to be the controlling materials. The licensee included this limiting ART value for B4197-2 in Table 1 of the enclosure to the licensees May 12, 2020, letter along with other RPV materials with high ART values.

The NRC staff verified these ART values using the neutron fluence values in Table 3 of the enclosure to the licensees May 12, 2020, letter and the RPV toughness properties in Tables 5.3.1-2 and 5.3.1-7 of the HNP, Unit 1 UFSAR (Reference 12), which include copper and nickel content, and initial nil-ductility reference temperature (RTNDT) values. The NRC staff finds the ART values in Table 1 of the enclosure to the licensees May 12, 2020, letter acceptable.

For the intermediate shell plate, heat number B4197-2, the ART values at the 1/4T and 3/4T locations are 212°F and 198°F, respectively. These represent an ART shift from the 36 EFPY values of 212°F - 191°F = 21°F at the 1/4T location, and 198°F - 179°F = 19°F at the 3/4T location.

For B4197-2 and intermediate shell-to-lower shell weld AB, heat number 5P6771, the licensee incorporated surveillance capsule data from Capsules U, V, X, and Z. The chemistry factor (CF) calculations per Position 2.1 of RG 1.99, Revision 2, resulting from the surveillance capsule data are given in Appendix F of the capsule Z report, ANP-3798NP. The NRC staff verified the CF calculations and finds them acceptable. Credibility of surveillance data must be demonstrated before using Position 2.1 of RG 1.99, Revision 2. This credibility evaluation is also in Appendix F of ANP-3798NP. The Appendix F evaluation concluded that the surveillance data for both B4197-2 and 5P6771 are not credible.

NRC document Generic Letter 92-01 and RPV Integrity Assessment (Reference 13), cited in Appendix F of ANP-3798NP, provides guidance on the use of non-credible surveillance data.

The evaluation using this guidance can be found in Appendix F, Section F.2 of ANP-3798NP.

This guidance states that even though surveillance data are shown to be non-credible, they may be used if the Table CF (i.e., the CF value resulting from the averaged chemistries of the surveillance data, not the CF value determined from the best fit line slope) is non-conservative.

The NRC staff reviewed this evaluation and found it is consistent with the methodology in Case 3: Non-Credible Surveillance Data and Table CF is Non-Conservative, of the guidance document Generic Letter 92-01 and RPV Integrity Assessment. Therefore, the NRC staff finds the CF values based on surveillance data in Appendix F of ANP-3798NP, acceptable.

3.2 Pressure/Temperature Limits The NRC staff evaluated the licensee's proposed 55 EFPY RCS P/T limits for cooldown, heatup, and inservice leak and hydrostatic testing shown in replacement TS Figures 3.4-2 and 3.4-3 of of the licensees May 12, 2020, letter. The NRC staff notes that the proposed revised TS Figure 3.4-3 did not show the criticality limits required by 2.c and 2.d of Table 1 of Appendix G to 10 CFR Part 50. In the supplemental October 1, 2020, letter, the licensee explained the licensing history of why the licensee does not consider criticality limits in TS 3/4.4.9. The licensee cited the NRCs safety evaluation (SE) supporting the issuance of Amendment No. 19 to HNP, Unit 1 operating license (Reference 14). In this SE, the NRC staff concluded that TS 3.4.9.1 and 3.4.9.2, which referenced TS Figure 3.4-3 that contained the criticality limits, were bounded by TS 3.1.1.4, and therefore determined that removal of the criticality limits imposed by TS 3.4.9.1 and 3.4.9.2 was acceptable. At the time of issuance of Amendment No. 19, TS 3.1.1.4 required that the reactor coolant system must be at a temperature of at least 551 degrees Fahrenheit (°F) before achieving criticality, with the exception of special tests during, which the minimum temperature must be at least 541°F. The NRC staff verified that the current TS 3.1.1.4 continues to impose the minimum temperature of 551°F before achieving criticality with the noted exception. The NRC staff also verified that TS 3.4.9.1 and 3.4.9.2 that refer to the revised Figure 3.4-3 applicable to 55 EFPY would be bounded by TS 3.1.1.4. Therefore, the NRC staff determined that not including the criticality limits in the 55 EFPY proposed changes to TS Figure 3.4-3 is acceptable.

Using the limiting ART values, the NRC staff independently calculated P/T limits based on the governing equation in Appendix G to Section XI of the ASME Code, but with the thermal stress intensity factor computed from a one-dimensional thermal stress analysis across a vessel wall.

The NRC staff compared its calculated values with those in the 55 EFPY P/T limits data in Tables 4 through 6 of the enclosure to the licensees May 12, 2020, letter, and determined that the licensees proposed 55 EFPY P/T limits are consistent with the methodologies in Appendix G to Section XI of the ASME Code. For higher temperatures (greater than about 250°F), the NRC staff noted that the licensees allowable pressure values are relatively lower, and therefore conservative with respect to the NRC staffs calculations. The NRC staff verified that the proposed 55 EFPY P/T limits in Figures 3.4-2 and 3.4-3 are based on the values in Tables 4 through 6 of the enclosure to the licensees May 12, 2020, letter. In addition, the NRC staff notes in the markup of TS Bases (Attachment 2 of the licensees May 12, 2020, letter), that the limiting RTNDT in the flange regions is 0°F and that therefore the minimum temperature required by Table 1 of Appendix G to 10 CFR Part 50 is 120°F, and that the proposed 55 EFPY P/T limits satisfy the minimum temperature limit.

The allowable pressure for the proposed 55 EFPY heatup P/T limits in the low temperature range (less than or equal to 130°F) is 621 pound per square inch gauge (psig), which is higher than the allowable pressure of approximately 560 psig in the previous (36 EFPY) heatup P/T limits, as shown in the TS 3/4.4.9 markup of Figure 3.4-3 in Attachment 1 to the licensees May 12, 2020, letter. The pressure value of 621 psig is 20 percent of the preservice hydrostatic test pressure, which is the maximum pressure allowed in 2.a of Table 1 of Appendix G to 10 CFR Part 50 when the core is not critical. The NRC staff notes that the allowable pressure should have decreased with increasing EFPY, but observed that in the low temperature range, the LTOP system would be enabled, and therefore, the peak pressure would be 559 psig (see discussion on LTOP Considerations in section 3.3 of this safety evaluation). The NRC staff determined that the higher allowable pressure in the low temperature range (less than or equal to 130°F) in the proposed 55 EFPY heatup P/T limits would have a negligible impact on plant operation.

There is no change in the way the licensee is handling instrumentation errors. The licensee is controlling instrumentation errors through Plant Procedure PLP-106 and thus there was no need to adjust the proposed 55 EFPY RCS P/T limits for instrumentation errors.

Based on the evaluation above, the NRC staff concludes that the licensee's proposed 55 EFPY RCS P/T limits for heatup, cooldown, and inservice leak and hydrostatic testing, are acceptable.

In Section 3 of the enclosure to the May 12, 2020, letter, the licensee addressed the potential of other ferritic materials not in the traditional RPV beltline to be more limiting per RIS 2014-11.

The licensee evaluated P/T limits at two structural discontinuities: the lower shell-to-torus shell weld and the upper shell-to-intermediate shell weld. The licensee stated that the P/T limits for these two discontinuities were not more limiting than the P/T limits for the traditional RPV beltline. The NRC staff, therefore, finds the licensees consideration of these two structural discontinuities acceptable.

The licensee did not evaluate the P/T limits for the RPV inlet and outlet nozzle. The NRC staff noted that the neutron fluence values at inlet and outlet nozzle lower welds of the RPV are 3.87E+17 n/cm2 and 1.86E+17 n/cm2, respectively, per Table 2 of the enclosure to the licensees May 12, 2020, letter. NRC-approved topical report, Pressurized Water Reactor Owners Group (PWROG), PWROG-15109-NP-A, PWR Pressure Vessel Nozzle Appendix G Evaluation (Reference 15), generically determined that if the maximum nozzle fluence is less than 4.28E+17 n/cm2, the nozzle P/T limits will not be limiting, relative to the P/T limits of the RPV beltline. Since the maximum neutron fluence value in the nozzles is 3.87E+17 n/cm2 which is less than 4.28E+17 n/cm2, the NRC staff finds that RPV nozzle P/T limits will not be limiting relative to the P/T limits in the traditional RPV beltline base metal and weld materials.

3.3 Low Temperature Overpressure Protection Considerations The NRC staff reviewed the licensees discussion of the acceptability up to 55 EFPY of the current temperature enable and PORV pressure settings in Figure 3.4-4 of TS 3/4.4.9. Using the limiting 1/4T ART value of 212°F at 55 EFPY, the NRC staff recalculated the required temperature enable value per G-2215 of ASME Code,Section XI, and verified that the current temperature enable value of 325°F is still valid. The licensee calculated a peak pressure 559 psig through a mass and heat input transient analysis with a single PORV opening, assuming the existing PORV setpoints in Figure 3.4-4 of TS 3/4.4.9. The NRC staff verified that this peak pressure does not exceed the allowable pressures of the proposed 55 EFPY P/T limits at low temperatures. Therefore, the NRC staff concludes that no change is necessary to the current temperature enable and PORV settings in Figure 3.4-4 of TS 3/4.4.9.

3.4 Technical Evaluation Summary Based on the above, the NRC staff finds that the proposed 55 EFPY RCS P/T limits and LTOP setpoints for the HNP, Unit 1, are consistent with Appendix G to Section XI of the ASME Code and continue to satisfy the requirements of Appendix G to 10 CFR Part 50. Additionally, the NRC staff finds that the revised TS Figures 3.4-2 and 3.4-3 continue to provide the lowest requisite functional capability or performance level required for safe operation, and therefore continue to meet the requirements of 10 CFR 50.36(c)(2). The NRC staff concludes that incorporating the proposed 55 EFPY RCS P/T limits and LTOP setpoints into the HNP, Unit 1, TS 3/4.4.9, is, therefore, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on January 6, 2021 (ADAMS Accession No. ML21006A404). The State of North Carolina official responded on January 25, 2021, with no comments (ADAMS Accession No. ML21025A292).

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration (85 FR 52370, dated August 25, 2020), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need to be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1.

Maza, Kim E., Duke Energy Progress, LLC., letter to U.S. Nuclear Regulatory Commission (NRC), License Amendment Request to Correct Non-Conservative Technical Specification 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System, May 12, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20134H888).

2.

Maza, Kim E., Duke Energy Progress, LLC., letter to NRC, Response to Request for Additional Information Regarding License Amendment Request to Correct Non-Conservative Technical Specification Pressure/Temperature Limits - Reactor Coolant System, October 1, 2020 (ADAMS Accession No. ML20275A241).

3.

Riley, Kevin, P., Duke Energy, letter to NRC, Submittal of the Summary Technical Report for the Reactor Pressure Vessel Surveillance Program Capsule Z, which includes, as an enclosure, Framatome [report] ANP-3798NP Revision 0, Analysis of Capsule Z, Duke Energy Shearon Harris Nuclear Power Plant, Reactor Vessel Material Surveillance Program, October 23, 2019 (ADAMS Accession No. ML19296C841).

4.

NRC Administrative Letter 98-10, Dispositioning of Technical Specifications that Are Insufficient to Assure Pant Safety, dated December 1998 (ADAMS Legacy Accession No. 9812280273), document is also located at the NRCs public website (https://www.nrc.gov/reading-rm/doc-collections/gen-comm/admin-letters/1998/al98010.html).

5.

NRC Generic Letter (GL) 92-01, Reactor Vessel Structural Integrity, Revision 1, dated March 6, 1992 (ADAMS Accession No. ML031070438) and Supplement 1, dated May 19, 1995 (ADAMS Accession No. ML031070449).

6.

NRC NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (SRP), Section 5.3.2, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, March 2007 (ADAMS Accession No. ML070380185).

7.

NRC Regulatory Guide (RG) 1.99, Radiation Embrittlement of Reactor Vessel Material, Revision 2, dated May 1988 (ADAMS Accession No. ML003740284).

8.

NRC RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Revision 0, dated March 2001 (ADAMS Accession No. ML010890301).

9.

NRC Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014 (ADAMS Accession No. ML14149A165).

10.

Chapter 1, Introduction and General Description of Plant, of Shearon Harris Nuclear Plant, Unit 1, Updated Final Safety Analysis Report (Amendment 63) (ADAMS Accession No. ML20147A018, part of ADAMS Package No. ML20147A016).

11.

Electronic correspondence from Mahoney, Michael, NRC to Zaremba, Art, Duke Energy, Request for Additional Information - Shearon Harris Nuclear Power Plant, Unit 1 -

Revise TS 3/4.4.9, Pressure/Temperature Limits - Reactor Coolant System (EPID L-2020-LLA-0111), dated September 3, 2020 (ADAMS Accession No. ML20247J559).

12.

Tables 5.3.1-2, Reactor Vessel Toughness Properties, and 5.3.1-7, Reactor Vessel Beltline Region Weld Metal, of Chapter 5, Reactor Coolant System and Connected Systems, of Shearon Harris Nuclear Plant, Unit 1, Updated Final Safety Analysis Report (Amendment 63) (ADAMS Accession No. ML20147A022, part of ADAMS Package No. ML20147A016).

13.

NRC document, Generic Letter 92-01 and RPV Integrity Assessment, dated February 12, 1998 (ADAMS Accession No. ML110070570).

14.

Becker, Richard A., NRC, letter to Eury, Lynn W. Carolina Power & Light Company, Issuance of Amendment No. 19 to Facility Operating License No. NPF Shearon Harris Nuclear Power Plant, Unit 1, Regarding Pressure Temperature Limits Relating to Generic Letter 88-11 (TAC No. 71500), dated May 31, 1990 (ADAMS Accession No. ML020560285).

15.

NRC-approved topical report from the Pressurized Water Reactor Owners Group (PWROG), PWROG-15109-NP-A, PWR Pressure Vessel Nozzle Appendix G Evaluation, Revision 0, dated January 2020 (ADAMS Accession No. ML20024E573).

Principal Contributors: D. Dijamco, NRR M. Mahoney, NRR Date of Issuance: March 22, 2021

ML21033B007 *by memorandum OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DNRL/NVIB/BC*

DSS/STSB/BC NAME MMahoney BAbeywickrama GCheruvenki (for HGonzalez)

VCusumano DATE 02/16/2021 02/11/2021 11/14/2020 01/28/2021 OFFICE DSS/SNSB/BC OCG - NLO DORL/LPL2-2/BC DORL/LPL2-2/PM NAME SKrepel TJones UShoop MMahoney DATE 02/02/2021 03/09/2021 03/22/2021 03/22/2021