ML14058B029
| ML14058B029 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/09/2014 |
| From: | Beltz T Plant Licensing Branch III |
| To: | Mccartney E Point Beach |
| Beltz T | |
| References | |
| TAC MF1943, TAC MF1944 | |
| Download: ML14058B029 (24) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Eric McCartney Site Vice President NextEra Energy Point Beach, LLC 6610 Nuclear Road Two Rivers, WI 54241 May 9, 2014
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT REGARDING THE USE OF OPTIMIZED ZIRLO ' FUEL ROD CLADDING MATERIAL (TAC NOS. MF1943 AND MF1944)
Dear Mr. McCartney:
The U.S. Nuclear Regulatory Commission (NRC) has issued Amendment Nos. 249 and 253 to renewed Facility Operating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2 (Point Beach), respectively. This amendment consists of changes to the facility technical specifications (TS) in response to your application dated June 4, 2013.
The amendment modifies the Point Beach TSs to allow the use of Optimized ZIRLO' as an approved fuel rod cladding material. This change is consistent with the NRC's allowed use of Optimized ZIRLO' fuel cladding material as documented in the safety evaluation (SE) included in Addendum 1-A to Westinghouse topical report, WCAP-12610-P-A and CENPD-404-P-A, "Optimized ZIRLO'." Your request for exemption from specific requirements of Title 10 of the Code of Federal Regulation (1 0 CFR), Section 50.46, and paragraph I.A.5 of Appendix K to 10 CFR Part 50 is addressed under separate correspondence (Agencywide Documents Access and Management System Accession No. ML14058B059).
A copy of our Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Docket Nos. 50-266 and 50-301
Enclosures:
- 1. Amendment No. 249 to DPR-24
- 2. Amendment No. 253 to DPR-27
- 3. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely,
~~
Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 249 License No. DPR-24
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by NextEra Energy Point Beach, LLC (the licensee) dated June 4, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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 4.B of the Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 249, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License No. DPR-24 Date of Issuance: May 9, 2014
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 253 License No. DPR-27
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by NextEra Energy Point Beach, LLC (the licensee) dated June 4, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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 4.B of the Renewed Facility Operating License No. DPR-27 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 253, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days.
FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License No. DPR-27 Date of Issuance: May 9, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 249 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AND LICENSE AMENDMENT NO. 253 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 Replace the following pages of Renewed Facility Operating License Nos. DPR-24 and DPR-27, and Technical Specifications with the attached revised page. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Renewed Facility Operating License REMOVE INSERT 3
3 Technical Specifications REMOVE 4.0-1 5.6-4 5.6-5 5.6-6 5.6-7 INSERT 4.0-1 5.6-4 5.6-5 5.6-6 5.6-7 D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach 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 E Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.
- 4.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 1 0 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 to 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 Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 249, are hereby incorporated in the renewed operating license.
Next Era Energy Point Beach shall operate the facility in accordance with Technical Specifications.
C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.
Renewed License No. DPR-24 Amendment No. 249 C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use at any time any byproduct. source, and special nuclear material as sealed neutron sources for reactor startup, sealed source for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use in amounts as required any byproduct, source of special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.
- 4.
This renewed operating 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 to 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 Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 253, are hereby incorporated in the renewed operating license.
NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.
C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21. 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.
Renewed License No. DPR-27 Amendment No. 253
Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Point Beach Nuclear Plant is located on property owned by NextEra Energy Point Beach at a site on the shore of Lake Michigan, approximately 30 miles southeast of the city of Green Bay. The minimum distance from the reactor containment center line to the site exclusion boundary as defined in 10 CFR 100.3 is 1200 meters.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 121 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy-4, ZIRLO, or Optimized ZIRLO' fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods or vacancies for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2 Rod Cluster Control (RCC) Assemblies Point Beach The reactor core shall contain 33 RCC assemblies. The control material shall be silver indium cadmium alloy clad with stainless steel as approved by the NRC.
4.0-1 Unit 1 - Amendment No. 249 Unit 2 -Amendment No. 253
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.4 CORE OPERATING LIMITS REPORT (COLR) (continued)
Point Beach (4)
WCAP-14787, Rev 3, "Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodology for Point Beach Units 1 & 2 Power Uprate (1775 MWt Core Power with Feedwater Venturis, or 1800 MWt Core Power with LEFM on Feedwater Header)"
(5)
WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using The NOTRUMP Code," August 1985.
(6)
WCAP-1 0054-P-A, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:
Safety Injection into the Broken Loop and COSI Condensation Model," Addendum 2, Revision 1, July 1997.
(7)
WCAP-87 45-P-A, "Design Bases for the Thermal Overpower
~T and Thermal Overtemperature ~T Trip Functions,"
September 1986.
(8)
DELETED (9)
WCAP-1 0924-P-A, "Large Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addenda, December 1988. (cores not containing 422 V+ fuel)
(10)
WCAP-10924-P-A, "LBLOCA Best Estimate Methodology:
Model Description and Validation: Model Revisions," Volume 1, Addendum 4, August 1990. (cores not containing 422 V+ fuel)
(11)
Caldon, Inc., Engineering Report-BOP, "TOPICAL REPORT:
Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM,/ System,"
Revision 0, March 1997.
(12)
Caldon, Inc., Engineering Report-160P, "Supplement to Topical Report R-80P: Basis for a Power Uprate With the LEFM,(rM System," Revision 0, May 2000.
( 13)
WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005.
(14)
WCAP-16259-P-A, "Westinghouse Methodology for Application of 3-D Transient Neutronics to Non-LOCA Accident Analysis,"
August 2006.
(15)
WCAP-8403 (nonproprietary), "Power Distribution Control and Load Following Procedures, "Westinghouse Electric Corporation," September 1974.
(16)
NS-TMA-2198, Westinghouse to NRC Letter, Attachment "Operation and Safety Analysis Aspects of Improved Load Follow Package," January 31, 1980.
(17)
NS-CE-687, Westinghouse to NRC Letter, "Power Distribution Control Analysis," July 16, 1975.
5.6-4 Unit 1 - Amendment No. 249 Unit 2 - Amendment No. 253
Reporting Requirements 5.6 5.6 Reporting Requirements (18)
(19)
WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," Apri11995.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'," July 2006.
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient *analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC 5.6.5 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
Point Beach
- a.
RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, hydrostatic testing, L TOP enabling, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
(1) LCO 3.4.3, "RCS Pressure and Temperature (PIT) Limits" (2) LCO 3.4.6, "RCS Loops-MODE 4" (3) LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled
(4) LCO 3.4.10, "Pressurizer Safety Valves" (5) LCO 3.4.12, "Low Temperature Overpressure Protection (L TOP)"
- b.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC Letters dated October 6, 2000, July 23, 2001, and October 18, 2007.
- c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6-5 Unit 1 - Amendment No. 249 Unit 2 - Amendment No. 253
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.6 PAM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 Tendon Surveillance Report Abnormal conditions observed during testing will be evaluated to determine the effect of such conditions on containment structural integrity. This evaluation should be completed within 30 days of the identification of the condition. Any condition which is determined in this evaluation to have a significant adverse effect on containment structural integrity will be considered an abnormal degradation of the containment structure.
Any abnormal degradation of the containment structure identified during the engineering evaluation of abnormal conditions shall be reported to the Nuclear Regulatory Commission pursuant to the requirements of 10 CFR 50.4 within thirty days of that determination. Other conditions that indicate possible effects on the integrity of two or more tendons shall be reportable in the same manner. Such reports shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedure and the corrective action taken.
5.6.8 Steam Generator Tube Inspection Report Point Beach A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG,
- b.
Active degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications, 5.6-6 Unit 1 - Amendment No. 229 Unit 2 -Amendment No. 234
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 Steam Generator Tube Inspection Report (continued)
Point Beach
- e.
Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f.
Total number and percentage of tubes plugged to date,
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing, and
- h.
The effective plugging percentage for all plugging in each SG.
- i.
Following completion of an inspection performed in Unit 1 Refueling Outage 31 (and any inspections performed in the subsequent operating cycle}, the number of indications and location, size, orientation, whether initiated on primary or secondary side for each service-induced flaw within the thickness of the tubesheet, and the total of the circumferential components and any circumferential overlap below 17 inches from the top of the tubesheet as determined in accordance with TS 5.5.8,
- j.
Following completion of an inspection performed in Unit 1 Refueling Outage 31 (and any inspections performed in the subsequent operating cycle}, the primary to secondary LEAKAGE rate observed in each steam generator (if it is not practical to assign leakage to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report, and
- k.
Following completion of an inspection performed in Unit 1 Refueling Outage 31 (and any inspections performed in the subsequent operating cycle), the calculated accident leakage rate from the portion of the tube below 17 inches from the top of the tubesheet for the most limiting accident in the most limiting steam generator.
5.6-7 Unit 1 - Amendment No. 229 Unit 2 - Amendment No. 234
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 249 and 253 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301
1.0 INTRODUCTION
By letter dated June 4, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13155A239), NextEra Energy Point Beach, LLC (NextEra or the licensee) submitted a license amendment request (LAR) for technical specification (TS) revisions and an exemption for the Point Beach Nuclear Plant (Point Beach), Units 1 and 2.
The proposed changes to TS 4.2.1, "Fuel Assemblies," and TS 5.6.4, "Core Operating Limits Report (COLR)," will add Optimized ZIRLO' as an acceptable fuel rod cladding material. The licensee also requested an exemption from the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.46, "Acceptance Criteria for Emergency Core Cooling Systems [ECCS] for Light-Water Nuclear Power Reactors," and Appendix K to 10 CFR Part 50, "ECCS Evaluation Models," to allow the use of fuel rods clad with Optimized ZIRLO' alloy for future reload applications. The U.S. Nuclear Regulatory Commission (NRC) staff addresses this exemption request in a separate document, with work performed under TAC Nos. MF1945 and MF1946 (ADAMS Accession No. ML14058B059).
The Optimized ZIRLO' cladding, manufactured by Westinghouse Electric Company (Westinghouse) is a new version of the ZIRLO' material and was approved in a topical report (TR) Addendum 1-A to WCAP-12610-P-A and CENPD-404-P-A, entitled "Optimized ZIRLO',"
for Westinghouse and Combustion Engineering (CE) fuel designs (ADAMS Accession Nos. ML051670403 and ML062080569). The fuel rod burnup limits were approved to a peak rod average of 62,000 megawatt-days per metric ton of uranium (MWd/MTU) for Westinghouse fuel, and 60,000 MWd/MTU forCE fuel. However, the NRC staff requires that licensees using Optimized ZIRLO' comply with the conditions and limitations listed in the NRC staff safety evaluation (SE) dated June 10, 2005 (ADAMS Accession No. ML051670408).
2.0 REGULATORY EVALUATION
The regulations in 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," allow a licensee to apply to amend the license or permit. The regulations in 10 CFR 50.92, "Issuance of Amendment," specify that the NRC staff will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate in determining whether an amendment will be issued to the applicant. The licensee requests a license amendment to add Optimized ZIRLO' as an acceptable fuel rod cladding material in the TS.
By letter dated June 10, 2005, the NRC staff issued an SE approving Addendum 1 to Westinghouse TR, WCAP-12610-P-A and CENPD-404-P-A, "Optimized ZIRLO'," wherein the NRC staff approved the use of Optimized ZIRLO' as an acceptable fuel cladding material.
This approval of Optimized ZIRLO' is based on:
- 1) similarities to standard ZIRLO',
- 2) demonstrated material performance, and
- 3) a commitment to provide irradiated data and validate fuel performance models ahead of burnups achieved in batch application.
The NRC staff's SE for Optimized ZIRLO' includes 10 conditions and limitations for its use, which will be discussed in the staff's Technical Evaluation.
The NRC staff's SE for Optimized ZIRLO' refers to Standard Review Plan (SRP), Section 4.2, "Fuel System Design," in its regulatory evaluation. Both the SE and SRP Section 4.2 refer to the General Design Criteria (GDC), contained in Appendix A of 10 CFR 50, as a regulatory basis for fuel system safety review. These GDCs were published in 1971, after Point Beach, Units 1 and 2, were licensed in 1967 and 1968, respectively. As such, both plants are not licensed to the Appendix A GDCs, but are licensed to their own plant-specific GDCs. The plant-specific GDCs are listed in Section 1.3 of the Point Beach Final Safety Analysis Report.
These plant-specific GDCs are basically equivalent to those published in Appendix A. In particular, Point Beach GDC-6 contains the same language as Appendix A, GDC-10, which is cited in the regulatory evaluation of the NRC staff's SE for Optimized ZIRLO' as a basis for the specified acceptable fuel design limits. There are also parallel GDCs discussing classifications of events and emergency core cooling system criteria. Because of the parallels between the Point Beach GDCs and 10 CFR 50, Appendix A, GDCs, the NRC staff determines that the conclusions of theSE for Optimized ZIRLO', including the conditions and limitations, are applicable to Point Beach, Units 1 and 2.
The NRC staff also considered the following regulatory requirements:
Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.36, "Technical specifications," in which the Commission established its regulatory requirements related to the contents of the TS. Specifically, 10 CFR 50.36(a)(1) states that "Each applicant for a license authorizing operation of a production or utilization facility shall include in its application proposed technical specifications in accordance with the requirements of this section."
Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power sources,"
which establishes standards for the calculation of emergency core cooling system (ECCS) performance and acceptance criteria for that calculated performance.
Paragraph (a)(1 )(i) of 10 CFR 50.46 states the each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an ECCS that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. Optimized ZIRLO' is not a zirconium alloy specified in this regulation as a fuel cladding material.
Appendix K to 10 CFR Part 50, "ECCS Evaluation Models," which establishes required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a LOCA.
Paragraph I.A.5 of Appendix K requires use of the Baker-Just equation to predict the rates of energy release, hydrogen concentration, and cladding oxidation for the metal-water reaction. The Baker-Just equation assumes the use of a zirconium alloy different than Optimized ZIRLO'.
3.0 TECHNICAL EVALUATION
3.1 Conditions and Limitations In the letter dated June 10, 2005, theSE approving the use of Optimized ZIRLO', (ADAMS Accession No. ML051670408) the NRC staff concluded that:
Based upon demonstrated material performance in Addendum 1 and in response to [request for additional information... ] and the irradiated database, the NRC staff has approved Optimized ZIRLOTM for full batch implementation.
In the SE conclusion, the NRC staff stated:
The NRC staff reviewed the effects of Optimized ZIRLO' using the appropriate fuel design requirements of [Standard Review Plan] SRP 4.2 and 10 CFR Part 50, Appendix A, General Design Criteria and found that the TR provided reasonable assurance that under both normal and accident conditions, Westinghouse and CE fuel assembly designs implementing Optimized ZIRLO' fuel cladding would be able to safely operate and comply with NRC regulations.
The NRC staff's SE also stated that licensees referencing Addendum 1 to the TR to implement Optimized Zl RLO ' must ensure compliance with 1 0 conditions, as specified in the SE.
NextEra documented its compliance with the applicable conditions and limitations in its LAR dated June 4, 2013 (ADAMS Accession No. ML13155A239), and committed to ensuring compliance with them for future fuel reloads.
Condition 1 Until rulemaking to 10 CFR Part 50 addressing Optimized ZIRLO' has been completed, implementation of Optimized ZIRLO' fuel clad requires an exemption from 10 CFR 50.46 and 1 0 CFR Part 50 Appendix K.
The licensee stated that the request for exemption from 10 CFR 50.46 and 10 CFR Part 50 Appendix K was included as an attachment to the LAR. The NRC staff finds this condition is met. As stated in Section 1.0, "Introduction," the exemption request is addressed by separate correspondence.
Condition 2 The fuel rod burnup limit for this approval remains at currently established limits:
62 GWd/MTU for Westinghouse fuel designs and 60 GWd/MTU for CE fuel designs.
The licensee stated that the maximum fuel rod burn up limit for Westinghouse fuel designs, including Optimized ZIRLO', continues to be 62 GWd/MTU. The NRC staff finds this condition is met.
Condition 3 The maximum fuel rod waterside corrosion, as predicted by the best-estimate model, will [satisfy proprietary limits included in TR and proprietary version of SE]
of hydrides for all locations of the fuel rod.
The licensee stated that the maximum fuel rod corrosion using the Optimized ZIRLOTM cladding will be confirmed to meet the design limit as part of the core reload design process. The core reload process is part of the current licensing basis, and provides verification that cycle-specific operating limits and conditions of the approved analytical methodologies referenced in COLR TS 5.6.4.b continue to be met. Based on the above, the NRC staff finds that this condition is met by the licensee's stated action and no additional requirements need to be imposed.
Condition 4 All the conditions listed in previous NRC SE approvals for methodologies used for standard ZIRLO' and Zircalo~-4 fuel analysis will continue to be met, except that the use of Optimized ZIRLOT cladding in addition to standard ZIRLO' and Zircaloy-4 cladding is now approved.
The licensee stated that the fuel analysis of Optimized ZIRLO' fuel rod cladding will continue to meet all conditions associated with approved methods, and will be confirmed as part of the normal reload design process. The core reload process is part of the current licensing basis, and provides verification that cycle-specific operating limits and conditions of the approved analytical methodologies referenced in COLR TS 5.6.4.b continue to be met. Based on the above, the NRC staff finds that this condition is met by the licensee's stated action and no additional requirement needs to be imposed.
Condition 5 All methodologies will be used only within the range for which ZIRLO' and Optimized ZIRLO' data were acceptable and for which the verifications discussed in Addendum 1 and responses to RAis were performed.
The licensee stated that the application of Optimized ZIRLO' in approved methodologies will be consistent with the approach accepted in WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, dated July 2006, and will be confirmed as part of the normal reload design process. The core reload process is part of the current licensing basis, and provides verification that cycle-specific operating limits and conditions of the approved analytical methodologies referenced in COLR TS 5.6.4.b continue to be met. Based on the above, the NRC staff finds that this condition is met by the licensee's stated action and no additional requirement needs to be imposed.
Condition 6 The licensee is required to ensure that Westinghouse has fulfilled the following commitment: Westinghouse shall provide the NRC staff with a letter(s) containing the following information (based on the schedule described in response to RAI #3):
- i.
Visual ii.
Oxidation of fuel rods iii. Profilometry iv. Fuel rod length
- v. Fuel assembly length
- b. Using the standard and Optimized ZIRLO' database including the most recent L TA data, confirm applicability with currently approved fuel performance models (e.g., measured vs. predicted).
The licensee indicated that Westinghouse provided the NRC staff a number of L TA data and models in several letters (ADAMS Accession Nos. ML070100388, ML073130560, ML080390452, ML090080381, and ML102140214). The data provided included, but was not limited to, the information described in Condition 6a.
Westinghouse will continue to evaluate the applicability and adequacy of the fuel performance models for the Optimized ZIRLO' fuel design. Continued acceptability of analytical methods described in the COLR TSs requires confirmation of the approved models' applicability up through the projected end-of-cycle burnup for the Optimized ZIRLO' fuel rods, and must be completed prior to its initial batch loading and prior to the startup of subsequent cycles. The licensee will confirm that, as higher burnups/fluences are achieved for Optimized ZIRLO TM clad fuel rods, the requirements of this condition are met as it applies to Point Beach, Units 1 and 2.
Based on the review of Westinghouse's L TA references cited by the licensee, and the fact that the core reload process is part of the current licensing basis, the NRC staff finds this condition is met by.the licensee.
Condition 7 The licensee is required to ensure that Westinghouse has fulfilled the following commitment: Westinghouse shall provide the NRC staff with a letter containing the following information (based on the schedule described in response to RAI
- 11 ):
- a. Vogtle growth and creep data summary reports
- b. Using the standard ZIRLO' and Optimized ZIRLO' database including the most recent Vogtle data, confirm applicability with currently approved fuel performance models (e.g., level of conservatism in W rod pressure analysis, measured vs. predicted, predicted minus measured vs. tensile and compressive stress).
Westinghouse provided the NRC staff a number of L TA data and models including the Vogtle results (ADAMS Accession Nos. ML070100388, ML073130560, ML080390452, ML090080381,
and ML102140214). The data provided included, but was not limited to, the information described in Condition 7a.
The licensee indicated that the data from previous cycles of operation had been evaluated and that the updated creep model has been used to predict the growth and creep in fuel rod performance. The licensee provided the favorable results to the NRC staff (ADAMS Accession No. ML102140214). Confirmation of the approved models' applicability up through the projected end-of-cycle burnup for the Optimized ZIRLO' fuel rods must be completed prior to its initial batch loading and prior to the startup of subsequent cycles. The licensee will confirm that the requirements of this condition are met as it applies to Point Beach, Units 1 and 2.
Based on the review of Westinghouse's L TA references cited by the licensee and the fact that the core reload process is, in part, required by the COLR TSs as part of the current licensing basis, the NRC staff finds this condition is met by the licensee.
Condition 8 The licensee shall account for the relative differences in unirradiated strength (YS [yield strength] and UTS [ultimate tensile strength]) between Optimized ZIRLO' and standard ZIRLO' in cladding and structural analyses until irradiated data for Optimized ZIRLO' have been collected and provided to the NRC staff.
- a. For the Westinghouse fuel design analyses:
- i.
The measured, unirradiated Optimized ZIRLO' strengths shall be used for BOL [beginning-of-life] analyses.
ii.
Between BOL up to a radiation fluence of 3.0 x 1021 n/cm2 [neutrons per square centimeter] (Energy (E) > 1 megaelectron-volt (MeV)), pseudo-irradiated Optimized ZIRLO' strength set equal to linear interpolation between the followin!l, two strength level points: At zero fluence, strength of Optimized ZIRLOT equal to measured strength of Optimized ZIRLO' and at a fluence of 3.0 x 1021 n/cm2 (E > 1 MeV), irradiated strength of standard ZIRLO' at the fluence of 3.0 x 1021 n/cm2 (E > 1MeV) minus 3 ksi.
iii. During subsequent irradiation from 3.0 x 1021 n/cm2 up to 12 x 1021 n/cm2, the differences in strength (the difference at a fluence of 3 x 1021 n/cm2 due to tin content) shall be decreased linearly such that the pseudo-irradiated Optimized ZIRLO' strengths will saturate at the same properties as standard ZIRLO' at 12 x 1021 n/cm2.
- b. For the CE fuel design analyses, the measured, unirradiated Optimized ZIRLO' strengths shall be used for all fluence levels (consistent with previously approved methods).
The licensee stated that the relative differences in unirradiated strength between Optimized ZIRLO' and standard ZIRLO' in cladding and structural analyses will be accounted for until irradiation data for Optimized ZIRLO' is provided to the NRC staff. Analysis of Optimized ZIRLO' clad fuel rods will use the yield strength (YS) and ultimate tensile strength (UTS) as modified per Conditions 8.a.i, 8.a.ii, and 8.a.iii until such time that irradiation data for Optimized ZIRLO' strengths are collected and provided to the NRC staff. Until the values are accepted by the NRC staff, the licensee will confirm that the requirements of this condition are met as applied to Point Beach, Units 1 and 2. The licensee stated that Condition 8.b is not applicable because Point Beach, Units 1 and 2, are Westinghouse fuel design plants.
Based on the licensee's statement, the NRC staff finds that this Condition 8.a is met by the licensee, and Condition 8.b does not apply.
Condition 9 As discussed in response to RAI #21, for plants introducing Optimized ZIRLO '
that are licensed with LOCBART or STRIKIN-11 and have a limiting PCT (peak cladding temperature] that occurs during blowdown or early reflood, the limiting LOCBART or STRIKIN-11 calculation will be rerun using the specified Optimized ZIRLO' material properties. Although not a condition of approval, the NRC staff strongly recommends that, for future evaluations, Westinghouse update all computer models with Optimized ZIRLO' specific material properties.
The licensee stated that Point Beach, Units 1 and 2, are not licensed with LOCBART or STRIKIN-11 loss-of-coolant accident methodology and, therefore, this condition does not apply.
Based on the licensee's statement, the NRC staff finds that this condition does not apply at Point Beach.
Condition 10 Due to the absence of high temperature oxidation data for Optimized ZIRLO',
the Westinghouse coolability limit on PCT during the locked rotor event shall be
[proprietary limits included in topical report and proprietary version of safety evaluation].
The licensee stated that the locked rotor event will be assessed against this coolability limit for the Optimized ZIRLO' fuel design as part of the normal reload design process. The core reload process is part of the current licensing basis and provides verification that cycle-specific operating limits and conditions of the approved analytical methodologies referenced in COLR TS 5.6.4.b continue to be met. Based on the above, the NRC staff finds that this condition is met by the licensee's stated action and that no additional requirement needs to be imposed.
Based on the information provided above, the NRC staff finds that the licensee response, as discussed above, meets the requirements of all applicable SE conditions and limitations for Point Beach. Therefore, the NRC staff concludes that the.Optimized ZIRLO' fuel design is acceptable for use at Point Beach, Units 1 and 2, to a peak rod average burn up limit of 62 GWd/MTU.
3.2 TS Revisions 3.2.1 TS Section 4.2.1, "Fuel Assemblies" The licensee proposes to add Optimized ZIRLO' as an acceptable fuel rod cladding material.
The new sentence is stated as follows:
Each assembly shall consist of a matrix of cylindrical Zircaloy-4, ZIRLO, or Optimized ZIRLO' clad fuel rods...
Based on the approval of Optimized ZIRLO' fuel cladding by the NRC staff, along with the licensee's documented compliance with the 10 conditions and limitations on Optimized ZIRLO' in theSE approving its use, the NRC staff finds the revision acceptable for Point Beach, Units 1 and 2.
3.2.2 Section 5.6.4, "Core Operating Limits Report (COLR)"
The licensee proposes to add two approved Westinghouse TRs to its list of approved analytical methods in TS Section 5.6.4.b. The addition of the first report, WCAP-12610-P-A, "VANTAGE+
Fuel Assembly Reference Core Report," April1995, is consistent with the transition to Westinghouse 422V+ fuel as authorized in approval of Amendment Nos. 193 (Unit 1) and 198 (Unit 2), dated February 8, 2000 (ADAMS Accession No. ML003683159). The second report, Addendum 1-A to WCAP-12610-P-A and CENPD-404-P-A, "Optimized ZIRLO'," allows the licensee to use approved analytical methods for Optimized ZIRLO'. Both reports have been reviewed and approved by the NRC staff. The reports ensure that core parameters and operating limits are established consistent with the results of safety analyses performed using NRC-approved methods. Therefore, the NRC staff finds these revisions acceptable for Point Beach, Units 1 and 2.
The licensee proposed other changes to the references in Section 5.6.4.b, including correcting the title of Reference 14 by adding the word "Accident," and correcting a typographical error in Reference 15 by changing the report number from "9403" to "8403." The NRC staff reviewed these changes and considers them to be editorial, and not technical, in nature. Therefore, the NRC staff finds these changes acceptable for Point Beach, Units 1 and 2.
3.3 Exemption to Regulations As explained in Section 2.0 above, the regulations in 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," and 10 CFR Part 50, Appendix K, "ECCS Evaluation Models," make no provisions for use of fuel rods clad in a material other than Zircaloy or ZIRLO'. Since the aforementioned regulations make no reference to Optimized ZIRLO' fuel cladding material, a plant-specific exemption is needed to support this amendment. The exemption (ADAMS Accession No. ML14058B059) is being issued separately from this SE and amendment.
3.4 Summary of Technical Evaluation The NRC staff has reviewed the licensee's LAR forTS revisions. Based on the evaluation, the NRC staff concludes that the Optimized ZIRLO' fuel design is acceptable to a peak rod average burn up limit of 62 GWd/MTU and the TS revisions are acceptable for the Point Beach Nuclear Plant, Units 1 and 2.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Wisconsin State official was notified of the proposed issuance of the amendment. The State officials provided no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or 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 NRC has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published October 29, 2013 (78 FR 64545). The amendment also makes minor editorial and corrective changes. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(1 0). 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.
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.
Principal Contributor: R. Anzalone D~e: May 9, 2014
- via memo dated December 2 2013 OFFICE LPL3-1/PM LPL3-1/LA SNPB/BC(A)
LPL3-1/BC LPL3-1 /PM NAME TBeltz MHenderson JDean
- REIIiott MYoung RCa rison TBeltz DATE 03/07/2014 03/04/2014 12/02/2014 03/10/2014 04/23/2014 05/1/2014 05/9/2014