ML24046A051
| ML24046A051 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  (NPF-086) |
| Issue date: | 05/07/2024 |
| From: | V Sreenivas Plant Licensing Branch 1 |
| To: | Coffey B NextEra Energy Seabrook |
| Poole J, NRR/DORL/LPLI, 415-2048 | |
| References | |
| EPID L-2023-LLA-0041 | |
| Download: ML24046A051 (1) | |
Text
May 7, 2024 Bob Coffey Executive Vice President, Nuclear and Chief Nuclear Officer Florida Power & Light Company NextEra Energy Seabrook, LLC Mail Stop: EX/JB 700 Universe Blvd.
Juno Beach, FL 33408
SUBJECT:
SEABROOK STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 174 RE: TO UPDATE THE PERIOD OF APPLICABILITY (POA) FOR THE PRESSURE-TEMPERATURE LIMITS (PTL) AND LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) CURVES (EPID L-2023-LLA-0041)
Dear Bob Coffey:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 174 to Renewed Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1. This amendment consists of changes to the technical specifications (TSs) in response to your application dated May 11, 2023, which supersedes the original submittal dated March 15, 2023, and replaces it in its entirety.
The amendment revises the applicability of the figures in the TSs for the reactor coolant system pressure-temperature limits. This modifies the period of applicability (POA) specified in the pressure-temperature limits curves of TS Figure 3.4-2, Reactor Coolant System Heatup Limitations - Applicable to 55 effective full-power years (EFPY), and Figure 3.4-3, Reactor Coolant System Cooldown Limitations - Applicable to 55 EFPY, and in Figure 3.4-4, Maximum Allowable power operated relief valve Setpoints for Cold Overpressure Protection System, and conforming changes to the TS Index by updating from 55 EFPY to 52.6 EFPY.
A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-443
Enclosures:
- 1. Amendment No. 174 to NPF-86
DOCKET NO. 50-443 SEABROOK STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 174 Renewed Facility Operating License No. NPF-86
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by NextEra Energy Seabrook, LLC, et al.
(the licensee), dated March 15, 2023, as supplemented by letter(s) dated May 11, 2023, 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.
- NextEra Energy Seabrook, LLC, is authorized to act as agent for the: Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, and Taunton Municipal Lighting Plant (collectively, with NextEra Energy Seabrook, LLC, licensees) and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
- 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-86 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 174, are incorporated into the Renewed Facility Operating License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: May 7, 2024 HIPOLITO GONZALEZ Digitally signed by HIPOLITO GONZALEZ Date: 2024.05.07 13:29:11 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 174 SEABROOK STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following page of Renewed Facility Operating License No. NPF-86 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3
3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Insert v
v vi vi 3/4 4-23 3/4 4-23 3/4 4-24 3/4 4-24 3/4 4-30 3/4 4-30 Amendment No. 174 (3)
NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)
NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility authorized herein.
(7)
DELETED C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions 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 NextEra Energy Seabrook, LLC, is authorized to operate the facility at reactor core power levels not in excess of 3648 megawatts thermal (100% of rated power).
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 174, are incorporated into the Renewed Facility Operating License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
INDEX v
02/28/2017 05/07/2024 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..................................................................
3/4 4-1 Hot Standby.............................................................................................
3/4 4-2 Hot Shutdown..........................................................................................
3/4 4-4 Cold Shutdown - Loops Filled.................................................................
3/4 4-6 Cold Shutdown - Loops Not Filled..........................................................
3/4 4-7 3/4.4.2 SAFETY VALVES Shutdown.................................................................................................
3/4 4-8 Operating.................................................................................................
3/4 4-9 3/4.4.3 PRESSURIZER.......................................................................................
3/4 4-10 3/4.4.4 RELIEF VALVES.....................................................................................
3/4 4-11 3/4.4.5 STEAM GENERATORS..........................................................................
3/4 4-13 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems....................................................................
3/4 4-14 Operational Leakage................................................................................
3/4 4-15 3/4.4.7 (THIS SPECIFICATION NUMBER IS NOT USED).................................
3/4 4-18 3/4.4.8 SPECIFIC ACTIVITY...............................................................................
3/4 4-19 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
> 1Ci/gram DOSE EQUIVALENT I-131........................................
3/4 4-20 TABLE 4.4-3 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS.......................................................................................
3/4 4-21 3/4.4.9 PRESSURE/TEMPERATURE LIMITS General....................................................................................................
3/4 4-22 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 52.6 EFPY....................................................
3/4 4-23
INDEX vi 02/28/2017 05/07/2024 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 52.6 EFPY....................................................
3/4 4-24 Pressurizer...............................................................................................
3/4 4-25 Overpressure Protection Systems...........................................................
3/4 4-26 FIGURE 3.4-4 MAXIMUM ALLOWABLE PORV SETPOINTS FOR COLD OVERPRESSURE PROTECTION SYSTEM..................................
3/4 4-30 3/4.4.10 DELETED................................................................................................
3/4 4-31 3/4.4.11 REACTOR COOLANT SYSTEM VENTS................................................
3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Hot Standby, Startup, and Power Operation...........................................
3/4 5-1 Shutdown.................................................................................................
3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350ºF..
3/4 5-4 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350ºF....................................
3/4 5-8 ECCS SUBSYSTEMS - Tavg Equal To or Less Than 200ºF....................
3/4 5-10 3/4.5.4 REFUELING WATER STORAGE TANK.................................................
3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity...............................................................................
3/4 6-1 Containment Leakage..............................................................................
3/4 6-2 Containment Air-Locks.............................................................................
3/4 6-7 Internal Pressure......................................................................................
3/4 6-9 Air Temperature.......................................................................................
3/4 6-10 Containment Vessel Structural Integrity..................................................
3/4 6-11 Containment Ventilation System..............................................................
3/4 6-12 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System.....................................................................
3/4 6-14 Spray Additive System.............................................................................
3/4 6-15 3/4.6.3 CONTAINMENT ISOLATION VALVES...................................................
3/4 6-16 3/4.6.4 COMBUSTIBLE GAS CONTROL (THIS SPECIFICATION NUMBER IS NOT USED).................................
3/4 6-18 (THIS SPECIFICATION NUMBER IS NOT USED).................................
3/4 6-19 Hydrogen Mixing System.........................................................................
3/4 6-20
SEABROOK - UNIT 1 3/4 4-23 Amendment No. 19, 89, 115, 135, 151 174 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate Rl808-I without using surveillance data, Position 1.1 LIMITI:NG ART VALUES AT 52. 6 EFPY:
l/4T. 1 I 7°F (A.xial Flaw) 3/4T, J05°F (A.xial Flaw)
Curves applicable for the first 52.6 EFPY and contain margins for possible instrument errors 2500 -r----------....-------------------,
2250 *
- 2000 1750
- c 0
~ 1500
~
"O...
t, 1250 I
0 II)
S2 1000 Q. -
Gl...
- J Ill Ill
~
- 0.
II)
~
750 500 l
- o l o*l 0
unaccep1able 01>ercnlon Healup rale below 12o*F shal not exceed 20°F/hr Critic:41 Limit Hea1u o Curve Compo~h* H.111up Curv*
lndle.111*d RCS Tampar.111u,.
T ~ 120'f 200"F > T > 120' F T ?:200'f 100"Flhr Criticality limit baiutd on Boltup inservice hydro1tatic test Temperature temperature (19 7°F) fOt the
- 60-F Hrvlco period upto S2.6 EFPY
.. I... I... ' ;... ' I.... '.. '. I ' '. ' I.. '. I.... I. '.. I ' ***
50 100 150 200 250 300 350 400 450 500 550 RCS Temperature (D~g. F, 10 Deg. F per division)
- Curve is Applicable for RCS Vacuum fill.
FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEA TUP LIMITATIONS - APPLICABLE UP TO 52. 6 EFPY
SEABROOK - UNIT 1 3/4 4-24 Amendment No. 19, 89, 115, 135, 151 174 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate RI S08-1 without using surveillance data, Position 1.1 LIM ITDIG ART VALUES AT 52.6 EFPY:
l/4T.
I l 7°F (Axial Flaw) 3/4T, 105°F (Axial Flaw)
Curves applicable for the first 52.6 EFPY and contain margins for possible instrument errors C
.Q II) i5...
Cl)
Cl.
Cl
<ii ll 0
It) ci U) 0..
G)..
- i en i...
0..
en (J
~
2500,-----------------------------.
2250 2000 1750 1500 1250 1000 750 500 250 Unacceptab5e 0
ration Acceptable Oooration Steady-State cu,ve
{0-Fhlr)
Composltt Cooldown Curvt lndlcJted RCS Boftup
~-i---1 TemPorau.n*
= 60"F Tem1141r;i1urt T < 150"F T ~ 1so*F M~1.lffl4.lm AJlow.lble Cooldown R;ilt
- zo*F1hr 100'FJhr o*
-~
- .<I,, -+ < i ii.<< O If i I I I' Ii TI t 1-.....-.-+.............-+-,,-....-r-\--,,.........,,.........
0 50 100 150 200 250 300 350 400 450 500 550 RCS Temperature (Deg. F, 10 Deg. F per division)
- Cuive is Applicable for RCS Vacuum fill.
FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN I.IMITATIONS -APPLICABLE UP TO 52.6 EFPY
SEABROOK - UNIT 1 3/4 4-30 Amendment No. 89, 115, 116, 135, 151 174 8!,Q
.ilOO 7!0 8
Vi 700
~
~O!.,
w U) 5 00 0..
- e
- J
= SY.>
~
500
- oo VALID FOR THE FIRST 52.6 EFPY, MAXIMUM SETPOINT ACCOUNTS FOR INSTRUMENT UNCERTAINTIES Ts 75 0°F, P = 520.0 PSIG; 75.0°F < r ~ 12s.o*F, P = 1 s*(T-75 o, + 520.0 PStG; 125.0'F <Ts. 160 0°F, P = 3. 71.(T-1250) + 595.G PSJG;
. *F <
C t
I-*- ----*-*** 1 -.- -i---
f
1----- --*------ *
+**-**-* -----
1 l I
- D 10D 1$()
200 300 INDICA ll:D RCS 'l'EMPERA'TURE (' F)
FIGURE 3..4-4 MAXIMUM ALLOWABLE PORV SETPOINTS FOR COl.D OVERPRESSURE
~OTECTION SYSTEM
- Note that above the enable temperature the PORV setpoints will not restrict plant neatup and cooldown operations since COMS Is not required to be armed at temperatures higher than 225"F. Hence the PORV setpoint values ramp up to the nominal setpolnt value of 2385 pslg Is not 5hown.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 174 TO FACILITY OPERATING LICENSE NO. NPF-86 NEXTERA ENERGY SEABROOK, LLC SEABROOK STATION, UNIT NO. 1 DOCKET NO. 50-443
1.0 INTRODUCTION
By letter dated May 11, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23131A115), NextEra Energy Seabrook, LLC (NextEra, the licensee) submitted Revision 1 to license amendment request (LAR) 23-01 to revise the applicability of the figures in the Technical Specifications (TSs) for the reactor coolant system (RCS) pressure-temperature (P-T) limits at the Seabrook Station, Unit 1 (Seabrook). The proposed amendment revises the applicability of the TS figures from 55 effective full-power years (EFPY) to 52.6 EFPY. Revision 1 to LAR 23-01 supersedes the original submittal for LAR 23-01 dated March 15, 2023 (ML23074A176), and replaces it in its entirety.
By letter dated May 4, 2023 (ML23117A365) the U.S. Nuclear Regulatory Commission (NRC) requested supplemental information from the licensee. In its response, NextEra submitted Revision 1, which supersedes and replaces in its entirety the original application dated March 15, 2023. Revision 1 renders the requested supplemental information no longer necessary since the superseding amendment request does not include a request to relocate the period of applicability (POA) from TS to licensee control. The requested amendment follows the latest reactor vessel peak fluence projections and the updated POA based on surveillance capsule dosimetry obtained at 26.46 EFPY. Data and information from WCAP-18607-NP, Analysis of Capsule X from the NextEra Energy Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program, dated September 30, 2021 (ML21277A388), includes the latest reactor vessel peak neutron fluence projections and updated period of applicability, and were incorporated into Revision 1. Because the change in the period of applicability resulting from the capsule X analysis results in a lower, non-conservative period of applicability in comparison to the currently approved period of applicability, the licensee concluded that authorization to update the POA in TS Figure 3.4-2 (Reactor Coolant System Heatup Limitations), Figure 3.4-3 (Reactor Coolant System Cooldown Limitations) and Figure 3.4-4 (Maximum Allowable power operated relief valve (PORV) Setpoints for Cold Overpressure Protection System) was necessary to correct the non-conservative TS.
2.0 REGULATORY EVALUATION
2.1 Applicable Regulations Under 10 CFR 50.92(a), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations.
Accordingly, for this LAR, the NRC staff must conclude that there is reasonable assurance that the proposed changes to the technical specifications do not endanger public health and safety.
The provisions of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36, Technical Specifications, paragraph (a)(1), require that each operating license application for a production or utilization facility include proposed TSs and a summary statement of the bases for such specifications. The provisions of 10 CFR) Part 50.36(c) require, in part, that TSs include the following categories related to facility operation: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.
The provisions of 10 CFR Part 50.60, Acceptance criteria for fracture prevention measures for light-water nuclear power reactors for normal operation, require that all light-water nuclear power reactors meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary (RCPB) set forth in 10 CFR Part 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements.
The provisions of 10 CFR Part 50, Appendix G, establish fracture toughness requirements to maintain the integrity of the RCPB in nuclear power plants. P-T limit requirements for the reactor pressure vessel (RPV) are established in paragraph IV.A.2 and Table 1 of this rule.
Paragraph IV.A.2 and Table 1 specify that P-T limit curves and minimum temperature requirements for the RPV are defined by the operating condition (i.e., pressure testing or normal operation, including anticipated operational occurrences), the RPV pressure, whether or not fuel is in the RPV, and whether the core is critical. In Table 1, the RPV pressure is defined as a percentage of the preservice system hydrostatic test pressure. The requirements for both the RPV P-T limit curves and the minimum RPV temperature must be met for all normal operating and pressure test conditions.
Additionally, 10 CFR Part 50, Appendix G, requires that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of the P-T limits and that the P-T limits be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials. Appendix H of 10 CFR Part 50 further requires a material surveillance program to monitor fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from exposure of these materials to neutron irradiation and the thermal environment.
2.2 Applicable Guidance Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ML003740284), describes procedures for calculating the adjusted nil-ductility transition reference temperature adjusted RTNDT (ART) due to neutron irradiation on RPVs.
RG 1.190, Revision 2, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001 (ML010890301), provides guidance on methods for the neutron fluence calculation. Fluence calculations are acceptable if they are performed with NRC-approved methods or with methods which are otherwise shown to conform with the guidance in RG 1.190 consistent with general design criteria 14, 30, and 31.
Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014 (ML14149A165), provides evaluation guidance for P/T curves and pressure and temperature limits reports, including the consideration of neutron fluence and structural discontinuities in the development of P/T curves.
3.0 TECHNICAL EVALUATION
3.1 Background
The ASME Code,Section XI, Appendix G methodology for generating P-T limit curves is based upon the principles of linear elastic fracture mechanics. The basic parameter of this methodology is the stress intensity factor, KI, which is a function of the stress state in the component and flaw configuration. The ASME Code,Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on these stress intensities for hydrostatic and pressure testing limits. The ASME Code,Section XI, Appendix G specifies that the P-T limits be generated by postulating a flaw with a depth that is equal to 1/4 of the RPV shell thickness and a length equal to 1.5 times the RPV section thickness. The critical locations in the RPV shell thickness (T) for calculating heat-up and cool-down P-T limit curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.
The P-T limit curve calculations are based, in part, on the reference nil-ductility temperature (RTNDT) for the material, as specified in the ASME Code,Section XI, Appendix G. The RTNDT is the critical parameter for determining the critical or reference stress intensity factor (fracture toughness, KIC) for the material. As required by 10 CFR Part 50, Appendix G, RTNDT values for materials in the RPV beltline region shall be adjusted to account for the effects of neutron radiation. Regulatory Guide 1.99, Revision 2, contains methodologies for calculating the ART due to neutron irradiation. The ART is defined as the sum of the initial (unirradiated) reference nil-ductility temperature (initial RTNDT), the mean value of the shift in reference temperature caused by irradiation (RTNDT), and a margin term. The RTNDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from the tables in RG 1.99, Revision. 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the postulated flaw depths described above. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the neutron fluence and the calculational procedures.
To satisfy the requirements of 10 CFR Part 50, Appendix G, methods for determining neutron fluence are necessary to estimate the fracture toughness of the RPV materials. Appendix H, Reactor Vessel Material Surveillance Program Requirements, of 10 CFR Part 50, requires the installation of surveillance capsules, including material test specimens and flux dosimeters, to monitor changes in fracture toughness.
3.2 Proposed Changes Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.9.1 requires that all times:
The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 100°F in any 1-hour period,
- b. A maximum cooldown of 100°F in any 1-hour period, and
- c. A maximum temperature change of less than or equal to 1 0°F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
TS LCO 3.4.9.3. requires in part:
The following Overpressure Protection Systems shall be OPERABLE:
- a.
In MODE 4 when the temperature of any RCS cold leg is less than or equal to 225°F; and in MODE 5 and MODE 6 with all Safety Injection pumps inoperable at least one of the following groups of two overpressure protection devices shall be OPERABLE when the RCS is not depressurized with an RCS vent area of greater than or equal to 1.58 square inches:
- 1) Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig +0, -3 %; or
- 2) Two power-operated relief valves (PORVs) with lift setpoints that vary with RCS temperature which do not exceed the limit established in Figure 3.4-4, or
The licensees LAR dated May 11, 2023, proposes the following changes to TS figures 3.4-2, 3.4-3, and 3.4-4, as used by and referenced in the LCOs above. The changes are to make the figures cover 52.6 effective full power years (EFPY) instead of 55 EFPY. In those three figures, each instance of 55 EFPY will be changed to 52.6 EFPY.
No other changes are proposed to the curves and values in the figures. The proposed changes are analyzed below.
3.3 P-T Limits and Low Temperature Overpressure Protection (LTOP) System Setpoint By letter dated November 2, 2015 (ML15096A255), the NRC issued Seabrook Amendment No. 151, which approved the current Seabrook P-T limit and LTOP curves in TS Figure 3.4-2, Figure 3.4-3, and Figure 3.4-4 for 55 EFPY of operation. The amendment included the P-T limit curves of Seabrook TS Figure 3.4-2 and Figure 3.4-3 and LTOP curve in Figure 3.4-4 that are currently approved for 55 EFPY.
In WCAP-18607-NP, NextEra submitted a summary technical report for surveillance capsule X from the Seabrook Reactor Vessel Radiation Surveillance Program, as required by Appendix H to 10 CFR Part 50. Information from WCAP-18607-NP, including the latest reactor vessel peak neutron fluence projections, mechanical test results, and an updated period of applicability, have been incorporated into Revision 1 of LAR 23-01. The capsule X specimens were tested at 26.52 EFPY after receiving a neutron fluence of 6.03 x 1019 n/cm2. The results of the capsule X analysis determined that the period of applicability in TS Figures 3.4-2, 3.4-3, and 3.4-4 needed revision from 55 EFPY to 52.6 EFPY.
The NRC staff confirmed the nil-ductility transition temperature shifts (RTNDT), chemistry factors, and Charpy upper-shelf energy values for the reactor vessel materials for capsule X contained in WCAP-18067-NP, Revision 0. The staff also confirmed that the adjusted RTNDT values for the Seabrook reactor vessel materials in Appendix D of WCAP-18607-NP, Revision 0, based on the surveillance capsule data, are within RG 1.99, Revision 2 projections and are credible according to the criteria in Section B of RG 1.99, Revision 2.
WCAP-18607-NP, Revision 0 analyzed the capsule X dosimetry which resulted in updated neutron fluence projections, including an updated peak clad/base metal fluence at 55 EFPY of 3.19 x1019 n/cm2. This is higher than the neutron fluence value of 3.05 x1019 n/cm2 which corresponds to the currently approved 55 EFPY P-T limit curves, as documented in WCAP-17441-NP. As a result, using the same 1/4T and 3/4T adjusted RTNDT values of 117°F & 105°F, respectively, and maximum reactor vessel neutron fluence for the limiting material (intermediate plate R1808-1), the neutron fluence and adjusted RTNDT values are now reached at 52.6 EFPY based upon the updated neutron fluence projections. Using the same limiting material properties for the P-T and LTOP limits, the staff find that the currently approved P-T and LTOP curves remain applicable, however, with the period of applicability is reduced from 55 EFPY to 52.6 EFPY.
3.4 Neutron Fluence Calculations The licensee provided WCAP-18607-NP, which included a neutron fluence evaluation based on capsule X dosimetry. Therein, the licensee utilized WCAP-18124-NP-A, Fluence Determination with RAPTOR-M3G and FERRET, an NRC approved Topical Report for fluence determination that is consistent with the guidance provided in RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. The licensees fluence evaluation appropriately applies the methodology including a benchmark against plant-specific measurement data. The licensee determined the calculational uncertainty to be 13%, which is less than the 20% uncertainty limit specified in RG 1.190. Therefore, the NRC staff has determined that the licensees application of WCAP-18124-NP-A is acceptable and that the fluence values calculated for various reactor pressure vessel components are reasonable estimations.
The licensee projected the peak clade/base metal neutron fluence at 55 EFPY to be 3.19 x1019 n/cm2. This is higher than the previous projection of 3.05 x1019 n/cm2. To maintain use of the current P-T and LTOP curves in the licensees TS, the period of applicability must be reduced from 55 EFPY. Based the licensees calculations, a reduction in the period of applicability to 52.6 EFPY will render the P-T and LTOP curves conservative once again. The NRC staff finds that the current P-T and LTOP curves remain applicable up to 52.6 EFPY given the calculated increase in peak neutron fluence.
3.5 Evaluation of Changes to Technical Specifications The RCS is a fission product barrier that is assumed to remain intact in the UFSAR accident analyses for several events. There are limits on the RCS pressure and RCS rates of temperature change to prevent excessive stresses to the RCS components. This helps to ensure that integrity of the RCS will be maintained. The TS contain LCOs which are operating requirements that are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The LCO definition and additional requirements for LCOs are contained in 10 CFR 50.36(c)(2)(i) and LCOs are required to be established per criteria in 10 CFR 50.36(c)(2)(ii). The proposed change revises figures that are referenced by LCOs 3.4.9.1 and 3.4.9.3. LCO 3.4.9.1 requires that RCS pressure and temperature, and changes to those parameters, be limited to prevent excessive stress and potential failure of the RCS. LCO 3.4.9.3 establishes minimum requirements to prevent overpressurization of the RCS at reduced temperatures. TS figures 3.4-2 and 3.4-3 (for LCO 3.4.9.1) provide RCS heatup and cooldown limitations for various pressures and temperature combinations. TS figure 3.4-4 (for LCO 3.4.9.3) provides maximum pressure settings for PORV for various pressure and temperature combinations. For TS 3.4.9.3, the PORV setpoints in the figure (and other equipment and operating requirements) are imposed to assure that excessive pressurization of the RCS will not occur at reduced temperatures during lower modes of operation. The NRC staff previously found that both of the LCOs, as implemented using the information on the curves, met the requirements set forth in the regulations.
The TS figures containing P-T limit curves and the PORV setpoints are proposed to be changed, based on testing and analysis of a surveillance capsule, as discussed in this safety evaluation. The analysis determined that the POA of the curves, as expressed in EFPY should be decreased because the current POA was found to be non-conservative. No changes to the P-T curves or PORV setpoints in the figures are proposed. The NRC staff reviewed the licensees analysis and confirmed its conclusions. The proposed change would reduce the length of operating time, in EFPY, that the curves may be used to ensure that the RCS is being operated under conditions that meet the LCO regulatory requirements. The updated POA is reflected on the proposed figures. No other changes are proposed. Operation within the restrictions in the figures will provide adequate assurance that the RPV will not fail during the applicable analyzed events. Plant operation within the P-T limits and PORV setpoint limits for the updated POA will continue to meet the regulatory requirements for LCOs on equipment needed to ensure safe operation of the facility. Therefore, the NRC staff finds that the proposed updated figures are acceptable.
The licensee provided markups of the TS that are consistent with the technical requirements of the plant design basis, the associated analyses, and the regulations. The TS changes require that the plant be operated within the bounds of the analysis by imposing appropriate limitations and conditions as required by 10 CFR 50.36.
4.0 TECHNICAL CONCLUSION Upon review of the information submitted by the licensee, the NRC staff has determined that the proposed change to the period of applicability from 55 EFPY to 52.6 EFPY for the Seabrook P-T limit and LTOP curves, based on the results of capsule X, is acceptable. Based on the evaluation above, the NRC staff concludes that the Seabrook P-T limit curves are applicable through 52.6 EFPY.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the New Hampshire and Massachusetts State officials were notified of the proposed issuance of the amendment on March 6, 2024. The State officials had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 [and changes surveillance requirements]. The NRC staff finds 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, and there has been no public comment on such finding (July 11, 2023; 88 FR 44162). Accordingly, the amendment meets 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 be prepared in connection with the issuance of the amendment.
7.0 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 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.
Principal Contributors: S. Smith, NRR B. Wise, NRR C. Fairbanks, NRR Date: May 7, 2024
- via memorandum OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DNRL/NVIB/BC(A)
NAME VSreenivas KZeleznock JTsao DATE 02/14/2024 02/16/2024 12/04/2023 OFFICE NRR/DSS/SNFB/BC NRR/DSS/STSB/BC(A)
OGC - NLO NAME SKrepel SMehta IMurphy DATE 02/07/2024 02/13/2024 03/11/2024 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME HGonzalez VSreenivas DATE 05/01/2024 05/07/2024