ML20148M359
| ML20148M359 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/08/2020 |
| From: | Vaughn Thomas Plant Licensing Branch II |
| To: | Stoddard D Virginia Electric & Power Co (VEPCO) |
| Thomas V | |
| References | |
| EPID L-2019-LLA-0206 | |
| Download: ML20148M359 (23) | |
Text
December 8, 2020 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, VA 23060-6711
SUBJECT:
SURRY POWER STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENT NOS. 302 AND 302 TO REVISE TECHNICAL SPECIFICATIONS FIGURE 3.1-1, SURRY UNITS 1 AND 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS, AND FIGURE 3.1-2, SURRY UNITS 1 AND 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS (EPID L-2019-LLA-0206)
Dear Mr. Stoddard:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 302 to Renewed Facility Operating License No. DPR-32 and Amendment No. 302 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station (Surry), Units 1 and 2, respectively.
The amendments revise the Technical Specifications (TS) in response to your application dated September 19, 2019 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML19269B775), as supplemented by letter dated January 28, 2020 (ADAMS Accession No. ML20034D727).
The amendments revise TS Figure 3.1-1, Surry Units 1 and 2 Reactor Coolant System Heatup Limitations, and Figure 3.1-2, Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations, for Surry Units 1 and 2 to update the cumulative core burnup applicability limit and to revise and relocate the limiting material property basis from the TS figures to the TS basis.
D.
A copy of the related safety evaluation is also enclosed. The Commissions biweekly Federal Register notice will include the notice of issuance.
Sincerely,
/RA/
Vaughn V. Thomas, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281
Enclosures:
- 1. Amendment No. 302 to DPR-32
- 2. Amendment No. 302 to DPR-37
- 3. Safety Evaluation cc: Listserv VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 302 Renewed License No. DPR-32
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated September 19, 2019, as supplemented by a letter dated January 28, 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 Specification as indicated in the attachment to this license amendment, and Figures 3.1-1 and 3.1-2 of the Renewed Facility Operating License No. DPR-32 are hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications Contained in Appendix A, as revised through Amendment No. 302 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective as of the date of issuance of the subsequent renewed license and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License and the Technical Specifications Date of Issuance: December 8, 2020 Michael T.
Markley Digitally signed by Michael T. Markley Date: 2020.12.08 12:48:46 -05'00' VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 302 Renewed License No. DPR-37
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated September 19, 2019, as supplemented by a letter dated January 28, 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 Specification as indicated in the attachment to this license amendment, and Figures 3.1-1 and 3.1-2 of the Renewed Facility Operating License No. DPR-32 are hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications Contained in Appendix A, as revised through Amendment No. 302 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective as of the date of issuance of the subsequent renewed license and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License and the Technical Specifications Date of Issuance: December 8, 2020 Michael T.
Markley Digitally signed by Michael T. Markley Date: 2020.12.08 12:49:20 -05'00'
ATTACHMENT TO AMENDMENT NO. 302 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 302 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contained marginal lines indicating the areas of change.
Renewed Facility Operating License No. DPR-32 REMOVE INSERT 3
3 Renewed Facility Operating License No. DPR-37 REMOVE INSERT 3
3 Technical Spectifications REMOVE INSERT 3.1-9 3.1-9 3.1-10 3.1-10 3.1-11 3.1-11 3.1-11a 3.1-12 3.1-12 Figure 3.1-1 Figure 3.1-1 Figure 3.1-2 Figure 3.1-2 Surry - Unit 1 Renewed License No. DPR-32 Amendment No. 302
- 3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 302 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.
E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 I.
Fire Protection The licensee shall implement and maintain in effect the provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report and as approved in the SER dated September 19, 1979, (and Supplements dated May 29, 1980, October 9, 1980, December 18, 1980, February 13, 1981, December 4, 1981, April 27, 1982, November 18, 1982, January 17, 1984, February 25, 1988, and Surry - Unit 2 Renewed License No. DPR-37 Amendment No. 302 E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such by product and special nuclear materials as may be produced by the operation of the facility.
- 3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power Levels not in excess of 2587 megawatts (thermal)
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 302 are hereby incorporated in this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.
C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.
D. Records The licensee shall keep facility operating records in accordance with the Requirements of the Technical Specifications.
E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227
Heatup and cooldown limit curves are calculated using a bounding value of the nil-ductility reference temperature, RTNDT, at the end of 68 Effective Full Power Years (EFPY) for Units 1 and 2. The heatup and cooldown limit curves were calculated using the most limiting value of RTNDT (228.4°F) which occurred at the 1/4-T, 0° azimuthal location in the Unit 1 intermediate-to-lower shell circumferential weld. The limiting RTNDT at the 1/4-T location in the core region is greater than the RTNDT of the limiting unirradiated material. This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.
The reactor vessel materials have been tested to determine their initial RTNDT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the copper and nickel content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials. The heatup and cooldown limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNDT at the end of 68 EFPY for Units 1 and 2 (as well as adjustments for location of the pressure sensing instrument).
Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cooldown curves must be recalculated when the RTNDT determined from the surveillance capsule exceeds the calculated RTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 68 EFPY for Units 1 and 2 prior to a scheduled refueling outage.
TS 3.1-9 Amendment Nos. 302 and 302
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.
The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of one and one half T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure. To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, RTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.
The approach for calculating the allowable limit curves for various heatup and cooldown rates in the 1986 Edition of the ASME Code specifies that the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The KIR curve is given by the equation:
KIR = 26.78 + 1.223 exp [0.0145(T - RTNDT + 160)]
(1) where KIR is the reference stress intensity factor as a function of the metal temperature T and the metal nil ductility reference temperature RTNDT. Thus, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:
C KIM + KIt < KIR (1) where, KIM is the stress intensity factor caused by membrance (pressure) stress.
TS 3.1-10 Amendment Nos. 302 and 302
TS 3.1-11 KIt is the stress intensity factor caused by the thermal gradients KIR is provided by the code as a function of temperature relative to the RTNDT of the material.
C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.
At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, KIt, for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60F per hour. The cooldown limit curves of Figure 3.1-2 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The cooldown limit curves are valid for cooldown rates up to 100F/hr.
The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 68 EFPY for Units 1 and 2. The adjusted reference temperature was calculated using materials properties data from the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) documented in the most recent revision to BAW-1543 and reactor vessel neutron fluence data obtained from plant-specific analyses.
Amendment Nos. 302 and 302
TS 3.1-11a The technical basis for the data points and the associated RTNDT values used to generate the heatup and cooldown curves is provided in WCAP-14177 (Reference 2) and were determined to be applicable to the 48 EFPY period of extended operation under first license renewal. The associated RTNDT values used to calculate the heatup and cooldown curves provided in WCAP-14177 (Revision 2) are based upon the Surry Unit 1 Intermediate to Lower Shell Circ Weld:
1/4-T, 228.4°F and 3/4-T, 189.5°F The heatup and cooldown curves for operation through 48 EFPY were based upon the Klr methodology. These heatup and cooldown curves were subsequently evaluated using the Klc methodology for Subsequent License Renewal (SLR) at 68 EFPY in WCAP-18243-NP (Reference 3).
The limiting reactor vessel materials at 68 EFPY were determined to be the Surry Unit 1 Lower Shell Longitudinal Weld L2 at 1/4-T and the Surry Unit 2 Intermediate to Lower Shell Circumferential Weld at 3/4-T. The associated RTNDT values calculated at 68 EFPY are:
1/4-T, 219.4°F and 3/4-T, 179.8°F The data points and the associated RTNDT values used to generate the heatup and cooldown curves in TS Figures 3.1-1 and 3.1-2, respectively, are conservative based upon use of the Klc methodology. Therefore, the heatup and cooldown curves did not require revision as a result of SLR. However, the fluence applicability is updated from 48 EFPY to 68 EFPY.
Amendment Nos. 302 and 302
TS 3.1-12 Amendment Nos. 302 and 302 The reactor boltup temperature is defined in 10 CFR 50, Appendix G as The highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload. The reactor vessel may be bolted up at a temperature greater than the initial RTNDT of the material stressed by the boltup (e.g., the vessel flange). As noted on Figures 3.1-1 and 3.1-2, the limiting boltup temperature is 10°F. An administrative minimum boltup temperature limit greater than 10°F is imposed in station procedures to ensure the Reactor Coolant System temperatures are sufficiently high to prevent damage to the reactor vessel closure head/vessel flange during the removal or installation of reactor vessel head bolts. The limiting boltup temperature and the administrative minimum boltup temperature limit are in effect when the reactor vessel head bolts are under tension.
References (1)
UFSAR, Section 4.1, Design Bases (2)
WCAP-14177, Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, (October 1994)
(3)
WCAP-18243, Rev. 2, Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, (July 2018)
Figure 3.1-1 Figure 3.1-1 : Surry Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60°F/hr) Applicable for 68 EFPY Surry Units 1 and 2 Reactor Coolant System Heatup Limitations Limiting Boltup Temperature Surry 1 Initial RTNDT Closure Flange Region: 10oF Amendment Nos. 302 and 302
Figure 3.1-2 Figure 3.1-2 : Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for 68 EFPY Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations Limiting Boltup Temperature Surry 1 Initial RTNDT Closure Flange Region: 10oF Amendment Nos. 302 and 302 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR THE UPDATE OF THE REACTOR COOLANT SYSTEM HEATUP AND THE PRESSURE-TEMPERATURE LIMITATIONS FIGURES AMENDMENT NO. 302 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 302 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY DOMINION ENERGY VIRGINIA SURRY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-280 AND 50-281
1.0 INTRODUCTION
By letter dated September 19, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19269B775), supplemented by letter dated January 28, 2020 (ADAMS Accession No. ML20034D727), Virginia Electric and Power Company (Dominion Energy Virginia, the licensee) submitted a license amendment request to revise Technical Specifications (TS) for the Surry Power Station (Surry), Units 1 and 2.
This amendment revises TS Figure 3.1-1, Surry Units 1 and 2 Reactor Coolant System Heatup Limitations, and Figure 3.1-2, Surry Units 1 and 2 Reactor Coolant System Cooldown Limitations, to update the cumulative core burnup applicability limit from 48 to 68 effective full-power years (EFPY) and to revise and relocate the limiting material property basis from the TS figures to the TS basis.
The licensee proposed that the 48-EFPY pressure-temperature (P-T) limits were acceptable for operation out to 68 EFPY. The licensees analysis providing the technical basis for its proposal appears in WCAP-18242, Revision 2, Surry Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal, issued July 2018, and WCAP-18243, Revision 3, Surry Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, issued January 2019 (ADAMS Accession No. ML20034D727).
The supplement dated January 28, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed determination that the amendment involves no significant hazards consideration, as published in the Federal Register on November 19, 2019 (84 FR 63901).
2.0 REGULATORY EVALUATION
The NRC has established requirements in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic licensing of production and utilization facilities, 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: Appendix G, Fracture Toughness Requirements, to 10 CFR Part 50; Appendix H, Reactor Vessel Material Surveillance Program Requirements, to 10 CFR Part 50; Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, issued May 1988; RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, issued March 2001; Generic Letter (GL) 92-01, Revision 1, Reactor Vessel Structural Integrity, dated March 6, 1992; GL 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity, dated May 19, 1995; and NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:
LWR Edition (the SRP), Section 5.3.2, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock.
Appendix G to 10 CFR Part 50 requires that reactor pressure vessel (RPV) beltline materials must have upper-shelf energy (USE) of no less than 75 foot-pounds (ft-lb) initially and 50 ft-lb throughout the life of the vessel. It also requires that facility P-T limit curves for the RPV be at least as conservative as those obtained by applying the linear elastic fracture mechanics 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). For RPV beltline materials having USE below 50 ft-lb, Appendix K, Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels, to Section XI of the ASME Code provides an equivalent margins analysis methodology to demonstrate that adequate fracture toughness is maintained. Appendix G to 10 CFR Part 50 establishes methodologies for determining the increase in transition temperature and the decrease in USE resulting from neutron radiation. GL 92-01, Revision 1, requested that licensees submit the RPV data for their plants to the NRC staff for review, and GL 92-01, Revision 1, Supplement 1, requested that licensees provide and assess data from other licensees that could affect their requirements related to facility RPV material surveillance programs. RG 1.99, Revision 2, contains guidance for RPV embrittlement integrity evaluations. SRP Section 5.3.2 provides an acceptable method for determining the P-T limit curves for ferritic materials in the beltline of the RPV based on the methodology in ASME Code,Section XI, Appendix G.
The most recent version of Appendix G to Section XI of the ASME Code, which has been endorsed in 10 CFR 50.55a, Codes and standards, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2013 Edition of the ASME Code. In addition to use of reference stress intensity factor KIc as fracture toughness and use of a circumferential flaw for circumferential welds, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20 percent of the preservice hydrostatic test pressure.
RG 1.190 describes the methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to the general design criteria (GDC) in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50. In consideration of the guidance in RG 1.190, GDC 14, Reactor coolant pressure boundary, GDC 30, Quality of reactor coolant pressure boundary, and GDC 31, Fracture prevention of reactor coolant pressure boundary, are applicable. GDC 14 requires the design, fabrication, erection, and testing of the RCPB so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. GDC 30 requires, among other things, that components comprising the RCPB be designed, fabricated, erected, and tested to the highest quality standards practical. GDC 31 pertains to the design of the RCPB, stating the following:
The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.
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), clarifies that P-T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may define P-T curves that are more limiting because the consideration of stress levels from structural discontinuities (such as RPV inlet and outlet nozzles) may produce a lower allowable pressure. RIS 2014-11 also clarifies that the beltline definition in 10 CFR Part 50, Appendix G, is applicable to all reactor vessel ferritic materials with projected neutron fluence values greater than 1x1017 neutrons per square centimeter (n/cm2) (E>1 megaelectron volts (MeV)), and this fluence threshold remains applicable for the design life, as well as throughout the licensed operating period.
3.0 TECHNICAL EVALUATION
3.1 Licensees Evaluation The current P-T limit curves for Surry Units 1 and 2 are valid for 48 EFPY. These curves are based on the limiting beltline material properties and the crack arrest toughness (KIa). The licensee documented its evaluation of the P-T limits for 68 EFPY in WCAP-18242, Revision 2, and WCAP-18243, Revision 3. In those documents, the licensee reevaluated the limiting beltline material using fracture toughness (KIc), updated fluence values for 68 EFPY, revised chemistry factor values, and updated initial reference temperature (RTNDT) values.
WCAP-18242-NP, Revision 2, evaluates the updated surface fluence for the Surry Unit 1 and 2 vessels. The licensee applied RG 1.99 to calculate the adjusted reference temperature (ART) values for the reactor beltline materials at 68 EFPY. The licensee compared the updated ART values to those used to develop the 48-EFPY P-T limit curves in Table 4.2.4-9 of the licensees subsequent license renewal application (SLRA) (ADAMS Accession No. ML18291A828). This reevaluation suggested that the current 48-EFPY P-T limit curves are acceptable for ensuring structural integrity of the RPV through 68 EFPY. Thus, the licensee proposed to update TS Figures 3.1-1 and 3.1-2 to revise the cumulative core burnup applicability limit from 48 to 68 EFPY, without changing the P-T limit curves themselves. The licensee also performed a fracture mechanics analysis of the RPV nozzles and concluded that the RPV shell materials remained limiting throughout 68 EFPY. The licensees SLRA contains detailed supporting information for this proposed change, including the following:
neutron fluence projections (Section 4.2.1)
USE (Section 4.2.2) pressurized thermal shock (Section 4.2.3)
ART (Section 4.2.4)
P-T limits (Section 4.2.5) low-temperature overpressure protection analyses (Section 4.2.6) 3.2 NRC Staffs Evaluation The licensee concluded that the 48-EFPY P-T limits are acceptable through 68 EFPY. The licensees analysis for 68 EFPY replaced the KIa curve for the toughness of the RPV materials with the KIc curve. The KIa curve is an older, more conservative method than the KIc curve. Use of the KIc curve is consistent with Appendix G to ASME Code,Section XI. The NRC staff has reviewed and accepted the KIC methodology as a conservative representation of the toughness of certain pressure vessel steels as part of incorporation by reference of the ASME Code into 10 CFR 50.55a. Therefore, the staff finds reevaluation of the P-T limits using KIc in place of KIa to be acceptable.
to the submittal shows the licensees limiting ART values: 219.4 degrees Fahrenheit (F) for the lower shell longitudinal weld L2 at the quarter-thickness (1/4T) for Unit 1 and 179.8 degrees F for the intermediate-to-lower-shell circumferential weld at 3/4T for Unit 2.
Both are based on the surveillance data. The applicant documented the calculation details in WCAP-18243, Revision 3. The P-T limit curve is based on the limiting material (i.e., the material with the highest ART value). The staff performed an independent evaluation of the licensees analysis of ART values, as described next.
The staff reviewed the licensees analysis on ART by confirming the licensees 68-EFPY ART calculations at the 1/4T location presented in WCAP-18243, Revision 3, Tables 5-3 through 5-7.
The staff verified that the licensee applied the methodology of RG 1.99, Revision 2, in determining the 68-EFPY ART values at the 1/4T location.
The staffs confirmatory evaluations included verifying the RTNDT(U) and RTNDT values, the margins due to uncertainties in both RTNDT(U) and RTNDT, and attenuation of the 68-EFPY fluence values to the 1/4T location. The staff noted that if attenuated fluence values are less than 1x1017 n/cm2 (E>1 MeV), the licensee set the RTNDT and the corresponding term to account for uncertainties due to RTNDT,, to zero, consistent with the fluence threshold established in Appendix H to 10 CFR Part 50 for monitoring changes in the fracture toughness properties of ferritic materials.
The NRC staff verified that the licensee correctly applied the master curve methodology in BAW-2308, Revisions 1-A and 2-A, Initial RTNDT of Linde 80 Weld Materials, (ADAMS Accession No. ML022200555) in determining the RTNDT(U) (and corresponding margin term for RTNDT(U)) of RPV beltline shell welds made of Linde 80 flux metal (Heat #s 299L44, 72445, 8T1762, and 8T1554). By letter dated June 27, 2007 (ADAMS Accession No. ML071160287),
the staff approved the exemption for the licensees use of the master curve methodology for Surry Units 1 and 2. For the other RPV beltline and extended beltline components, the NRC staff verified that the licensee applied acceptable methodologiesParagraph NB-2331 of Section III of the ASME Code; Branch Technical Position 5-3, Fracture Toughness Requirements; BWRVIP-173-A; Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials, (ADAMS Accession No. ML12083A268) and fabrication information from Supplement 1 of BAW-2313, Revision 7, B&W Fabricated Reactor Vessel Materials and Surveillance Data Information,for determining RTNDT(U), as the NRC staff observed in the in-office audit report (ADAMS Accession No. ML19128A079) for SLRA Section 4.2.4. The NRC staff noted that even though the RPV nozzle-to-shell welds are made of Linde 80 flux metal (Heat #s 8T1762 or 8T1554), the licensee did not apply the master curve methodology in BAW-2308, Revisions 1-A and 2-A, in determining the RTNDT(U) for these welds, but instead it applied the fabrication information from Supplement 1 of BAW-2313, Revision 7.
The NRC staff verified that the licensee applied the methodology in Position 2.1 of RG 1.99, Revision 2, to determine RTNDT for base and weld materials in the RPV beltline that had available credible surveillance capsule data as identified in WCAP-18243, Tables 5-3 through 5-7. Furthermore, the NRC staff verified that the licensee included surveillance capsule data from the appropriate non-Surry sources for weld heats that are irradiated in non-Surry plants, as the staff observed in the in-office audit report for SLRA Section 4.2.4.
The NRC staff verified that SLRA Table 4.2.4-9 shows the limiting materials. For Surry Unit 1, the limiting material is the lower shell axial weld L2 (Heat #299L44), with a 68-EFPY ART value of 219.4 degrees F at 1/4T and the intermediate-to-lower-shell circumferential weld with a 68-EFPY ART value of 173.6 degrees F at 3/4T. For Surry Unit 2, the limiting material is the intermediate-to-lower-shell circumferential weld (Heat #0227) at both 1/4T and 3/4T, with 68-EFPY ART values of 210.7 degrees F and 179.8 degrees F, respectively. Therefore, in the TS basis revision, the licensee proposed to cite the limiting values of 219.4 degrees F and 179.8 degrees F. For these limiting materials, the staff verified that the licensee correctly applied the guidance in Position 2.1 of RG 1.99, Revision 2, for using credible surveillance data in calculating the 68-EFPY ART values. Therefore, based on the review above, the staff finds that the licensees projections of limiting material ART values are acceptable.
In addition to the review of the licensees reevaluation of ART, the NRC staff performed an independent calculation of the P-T limit curves in the current TS. The NRC staff included relevant input from WCAP-18243, including end-of-life surface fluence, chemistry factors, and vessel geometry parameters. The NRC staff drew on thermal stress intensity factors for heatup and cooldown transients tabulated in ORNL/NRC/LTR-03/03, Tabulation of Thermally-Induced Stress Intensity Factors (KIT) and Crack Tip Temperatures for Generating Pressure-Temperature Curves per ASME Section XIAppendix G, issued March 2003 (ADAMS Accession No. ML110070355). Other calculation steps followed by the NRC staff are consistent with ASME Code,Section XI, Nonmandatory Appendix G. Finally, the NRC staff independently confirmed that the RPV nozzle P-T limit curves are bounded by the RPV P-T limit curves, as discussed in RIS 2014-11. As a result of these P-T limit confirmatory calculations and the NRC staffs evaluation of the limiting ART calculation, the NRC staff concludes that the licensees proposed P-T limit curves for 68 EFPY are acceptable.
4.0 TECHNICAL CONCLUSION The NRC staff concludes that the proposed changes to TS Figures 3.1-1 and 3.1-2 meet the requirements of Appendix G to 10 CFR Part 50. The NRC staff performed independent evaluations and verified that the 68-EFPY ART values were bounded by the 48-EFPY ART values. As an additional step, the NRC staff performed confirmatory calculations showing that the 68-EPFY P-T limit curves, using inputs discussed in the licensees submittal, were bounded by the 48-EFPY P-T limit curves. Thus, the NRC staff considers the licensees proposed change to TS Figures 3.1-1 and 3.1-2 to reflect the increased cumulative core burnup applicability limit RCS P-T limits, low-temperature overpressure protection setpoints, and T-Enable values from 48-to 68-EFPY cumulative core burnup applicability limit to be acceptable.
The staff also finds that the proposed change will not impact the licensees continued compliance with regulatory requirements and guidance described in Section 2.0 of this safety evaluation.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the NRC notified an official from the Virginia Division of Radiological Health of the proposed issuance of the amendment. On May 18, 2020, the State official confirmed that the Commonwealth of Virginia had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 previously issued a proposed finding that the amendments involve no significant hazards consideration, published in the Federal Register on November 19, 2019 (84 FR 63901),
and the agency has received no public comments on this finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Under 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
Michael Benson Date: December 8, 2020
- via e-mail OFFICE NRR/DORL/ LPL2-1/PM*
NRR/DORL/LPL2-1/LA*
NRR/DNRL/NVIB/BC*
NAME VThomas KGoldstein HGonzalez DATE 05/26/2020 12/03/2020 04/08/2020 OFFICE OGC*
NRR/DORL/LPL2-1/BC*
NRR/DORL/LPL2-1/PM*
NAME KGamin MMarkley VThomas DATE 06/02/2020 12/08/2020 12/08/2020