ML13113A400
| ML13113A400 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  (DPR-042, DPR-060) |
| Issue date: | 07/02/2013 |
| From: | Thomas Wengert Plant Licensing Branch III |
| To: | Jeffery Lynch Northern States Power Co |
| Wengert T | |
| References | |
| TAC ME9254, TAC ME9255, TSTF-510 | |
| Download: ML13113A400 (39) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 2,2013 Mr. James E. Lynch Site Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089-9642
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING, PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: UNIT 2 STEAM GENERATOR REPLACEMENT AND ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE TRAVELER-51 0, REVISION 2 (TAC NOS. ME9254 AND ME9255)
Dear Mr. Lynch:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 208 to Renewed Facility Operating License No. DPR-42 and Amendment No. 195 to Renewed Facility Operating License No. DPR-60 forthe Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, respectively. The amendments consist of changes to'the Technical Specifications (TSs) in response to your application dated July 25, 2012.
The amendments revise Technical Specifications (TSs) 3.4.19 - "Steam Generator (SG) Tube Integrity," 5.5.8 - "Steam Generator (SG) Program," and 5.6.7 - "Steam Generator Tube Inspection Report" to apply the appropriate program attributes to the Unit 2 replacement steam generators that are planned for installation in fall 2013. The amendments also revise the PINGP Units 1 and 2 TSs to adopt the program improvements in Technical Specifications Task Force Traveler (TSTF) 510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
J. Lynch w.2 A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306
Enclosures:
- 1. Amendment No. 208 to DPR-42
- 2. Amendment No. 195 to DPR-60
- 3. Safety Evaluation cc w/encls: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERA riNG LICENSE Amendment No. 208 License No. DPR-42
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated July 25, 2012, 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 2.C.(2) of Renewed Facility. Operating License No. DPR-42 is hereby amended to read as follows:
- 2 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, are hereby incorporated in the renewed operating license.
NSPM shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days after reactor startup following Unit 2 steam generator replacements.
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 Renewed Facility Operating License and Technical Specifications Date of Issuance: Jgly 2, 2013
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 195 License No. DPR-60
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated July 25,2012, 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 2.C.(2) of Renewed Facility Operating License No. DPR-60 is hereby amended to read as follows:
- 2 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.195, are hereby incorporated in the renewed operating license.
NSPM shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days after reactor startup following Unit 2 steam generator replacements.
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 Renewed Facility Operating License and Technical Specifications Date of Issuance: Jilly 2, 2013
ATTACHMENT TO LICENSE AMENDMENT NOS. 208 AND 195 RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42 and DPR-60 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT DPR-42, License Page 3 DPR-42, License Page 3 DPR-60, License Page 3 DPR-60, License Page 3 Replace the following pages of the Appendix A Technical Specifications with the att~ched revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT 3.4.19-1 3.4.19-1 3.4.19-2 3.4.19-2 5.0-13 5.0-13 5.0-14 5.0-14 5.0-15 5.0-15 5.0-16 5.0-16 5.0-17 5.0-17 5.0-18 5.0-18 5.0-19 5.0-19 5.0-20 5.0-20 5.0-21 5.0-21 5.0-39 5.0-39 5.0-40 5.0-40
- 3 (3)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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; (4)
Pursuant to the Act and 1 °CFR Parts 30, 40, and 70, NSPM 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 and equipment calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6)
Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer. byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination.
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.
(3)
Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-42 Amendment No. 208
- 3 (3)
Pursuant to the Act and 10 CFR Parts 30, 46 and 70, NSPM 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; (4)
Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM 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 and equipment calibration or associated with radioactive apparatus or components;
,(S)
Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6)
Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites own,ed by NSPM for the purposes of volume reduction and decontamination. '
C.
This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Sections SO.S4 and SO.S9 ofPart SO, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 195, are hereby incorporated in the renewed operating license.
NSPM shall operate the facility in accordance with the Technical Specifications (3)
Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.SS (S1 FR 27817 and 27822) and to the authority of 10 CFR SO.90 and 10 CFR SO.S4(p). The combined set of plans, which contains Renewed Operating License No. DPR-60 Amendment No. 19S
SG Tube Integrity 3.4.19 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.19 Steam Generator (SG) Tube Integrity LCO 3.4.19 SG tube integrity shall be maintained.
All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program APPUCAB1UIY:
MODES 1, 2, 3, and 4.
ACTIONS
NOTE-------------------------------------------------
Separate Condition entry is allowed for each SG tube.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program.
A.l Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG inspection.
7 days A.2 Plug the affectt(d tube(s) in accordance with the Steam Generator Program.
Prior to entering MODE 4 following the next refueling outage or SG tube inspection Prairie Island Unit 1 Amendment No. m-208 Units 1 and 2 3.4.19-1 Unit 2 - Amendment No..f.67. 195
SG Tube Integrity 3.4.19 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and associated Completion Time of Condition A not met.
SG tube integrity not maintained.
8.1 Be in MODE 3.
AND 8.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.19.1 Verify SG tube integrity in accordance with the Steam Generator Program.
In accordance with the Steam Generator Program SR 3.4.19.2 Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program.
Prior to entering MODE 4 following an SG tube inspection Prairie Island Unit 1 - Amendment No. -l:-++ 208 Units 1 and 2 3.4.19-2 Unit 2 Amendment No. Mf 195
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Steam Generator (SG) Program A Stearn Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Stearn Generator Program shall include the following:
- a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
- b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.
- 1. Structural integrity performance criterion: All in-service steam.
generator tubes shall retain structural integrity over the full range
(
of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to 'burst or collapse.
Prairie Island Unit 1 - Amendment No. ~ +++ 208 Units 1 and 2 5.0-13 Unit 2 Amendment No..f.49 -HH-195
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued)
In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and' assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
- 2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed.l gpm per SG.
- 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.14, "RCS Operational LEAKAGE."
- c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceedIng 40% of the nominal tube wall thickness shall be plugged.
- d. Provisions for SG tube inspections. Periodic SG tube inspections shaH be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to tube sheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.l, d.2 and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be Prairie Island Unit 1 - Amendment No. +5& +++ 208 Units 1 and 2 5.0-14 Unit 2 - Amendment No. -l49.f..67. 195
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8
. Steam Generator (SG) Program (continued) performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 1. Ins'pect 100% of the tubes in each SG during the first refueling outage following SG installation.
- 2. After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, cand d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during theremainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
a)
After the first refueling outage following SG installation, inspect 100% of the tubes during the next 144 effective full power months. This constitutes the first inspection period; Prairie Island Unit 1 - Amendment No. H& H-+ 208 Units 1 and 2 5.0-15 Unit 2 - Amendment No. +/-49+6+ 195
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) b)
During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c)
During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d)
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.
- 3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive I.
information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
- e. Provisions for monitoring operational primary to secondary LEAKAGE.
Prairie Island Unit 1 - Amendment No. +§-8 t+l-208 Units 1 and 2 5.0-16 Unit 2 - Amendment No. 149 +6+ 195
5.5 5.5 Programs and Manuals and Manuals This page retained for page numbering Prairie Island Unit 1 - Amendment No. ~ ++7 208 Units I and 2 5.0-17 Unit 2 - Amendment No. -l-49 -l67-195
Programs and Manuals 5.5 5.5 Programs and Manuals (continued)
This page retained for page numbering Prairie Island.
Unit 1 - Amendment No. l§8 H+ 208 Units 1 and 2 5.0-18 Unit 2 - Amendment No. -l-49 +&7-195
Programs and Manuals 5.5 5.5 Programs and Manuals (continued)
This page retained for page numbering Prairie Island Unit 1 ~ Amendment No. M& +++ 208 Units 1 and 2 5.0-19 Unit 2 Amendment No. -l49 +61195
Programs and Manuals
.5.5 5.5 Programs and Manuals (continued)
This page retained for page numbering Prairie Island Unit 1 - Amendment No. +§.S 208 Units 1 and 2 5.0-20 Unit 2 - Amendment No. +49 W 195
5.5 5.5 Programs and Manuals HTP;ITTl" and Manuals This page retained for page numbering Prairie Island Unit 1 Amendment No. M& H-+ 208 Units 1 and 2 5.0-21 Unit 2 Amendment No. -l-49 -Ht7 195
5.6 Reporting Requirements
- 5.6 Reporting Requirements (continued) 5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entryinto 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. 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,
- e. Number of tubes plugged during the inspection outage for each degradation mechanism,
- f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and
- g. The results of condition monitoring, including the results of tube pulls and in-situ testing.
Prairie Island Unit 1 - Amendment No. l+f -l-99 208 Units 1 and 2 5.0-39 Unit 2 - Amendment No. m W 195
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.8 EM Report When a report is required by <:ondition C or I of LCO 3.3.3, "Event Monitoring (EM) 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.
Prairie Island Unit 1 Amendment No. m- -l-9-9 208 'I, Units 1 and 2 5.0-40 Unit 2 Amendment No. i:-61 +8+ 195
UNITED 5TATE5,.o~
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 208 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 195 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY - MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306
1.0 INTRODUCTION
By application dated July 25, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12207A522), Northern States Power Company, a Minnesota Corporation (the licensee), doing business as Xcel Energy, requested changes to the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2.
The proposed changes would revise TS 3.4.19 "Steam Generator (SG) Tube Integrity," TS 5.5.8 "Steam Generator (SG) Program," and TS 5.6.7 "Steam Generator Tube Inspection Report" and support the replacement of the SGs at PINGP Unit 2 during the fall 2013 outage. The licensee also proposes to revise the same TSs listed above to adopt the program improvements in Technical Specification Task Force Traveler (TSTF)-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," for PINGP, Units 1 and 2.
The licensee stated that this license amendment request (LAR) is consistent with TSTF-510, Revision 2, which was noticed as being available for use as part of the consolidated line item improvement process in the Federal Register on October 27, 2011 (76 FR 66763). Because the LAR includes TS revisions for both the SG replacement and the implementation of TSTF-51 0, this LAR is not being processed under the consolidated line item improvement process.
TSTF Travelers, such as TSTF-510, evaluate changes to the Standard Technical Specifications (STS). The STS applicable to the PINGP Nuclear Steam Supply System is NUREG-1431, "Standard Technical Specifications Westinghouse Plants." The current STS provisions related to SG programs were established in May 2005 with the NRC staffs approval of TSTF-449, Revision 4, "Steam Generator Tube Integrity" (!\\IRC Federal Register Notice of Availability
- 2 (70 FR 24126). The TSTF-449 changes to the STS incorporated a new, largely performance based approach for ensuring that the integi"ity of the SG tubes is maintained. The performance based provisions were supplemented by prescriptive provisions relating to tube inspections and tube repair limits to ensure that conditions adverse to quality are detected and corrected on a timely basis.. By letter dated March 20, 2007 (ADAMS Accession No. ML070330455), the NRC approved TSTF-449 for implementation in the PINGP TS.
After the issuance of TSTF-449, TSTF-510 was developed to reflect the industry's early implementation experience with respect to TSTF-449. TSTF-510 characterizes the changes as editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with implementing industry documents, and usability without changing the intent of the requirements. Further, according to the licensee's application, the proposed changes are an improvement to the existing SG inspection requirements and continue to provide assurance that*
the plant licensing basis will be maintained between SG inspections.
The following section details the regulatory requirements and guidance used by the NRC staff to evaluate the application.
2.0 REGULATORY EVALUATION
Steam generator tubes function as an integral part of the reactor coolant pressure boundary (RCPB) and also serve to isolate radiological fission products in the primary coolant from the secondary coolant and the environment. For the purposes of this safety evaluation, tube integrity means that the tubes are capable of performing these functions in accordance with the plant design and licensing basis.
Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements with respect to the integrity of the SG tubing. Specifically, the General Design Criteria (GDC) in Appendix A to 10 CFR Part 50 states that the RCPB shall have "an extremely low probability of abnormalleakage...and gross rupture" (GDC 14), "shall be designed with*
sufficient margin" (GDCs 15 and 31), shall be of "the highest quality standards possible" (GDC 30), and shall be designed to permit "periodic inspection and testing...to assess...
structural and leak tight integrity" (GDC 32). To this end, 10 CFR 50.55a(c) specifies that components that are part of the RCPB must meet the requirements for Class 1 components in Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code). Section 50.55a(g) further requires that components and supports that are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice examination of these components and must meet the pre-service examination requirements set forth in the editions and addenda of Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the ASME Code incorporated by reference in 10 CFR 50.55a(b) that were applied to the construction of the particular component.
Section XI requirements pertaining to inservice inspection of SG tubing are augmented by additional SG tube surveillance requirements in the TSs.
As part of the plant licensing basis, applicants for pressurized-water reactor (PWR) licenses are required to analyze the consequences of postulated design-basis accidents such as an SG tube rupture and main steam line break. These analyses consider the primary-to-secondary leakage that may occur during these events and must show that the offsite radiological consequences do
- 3 not exceed the applicable limits of the 10 CFR Part 100.11 guidelines for offsite doses (or 10 CFR 50.67, as appropriate); GDC 19 criteria for control room operator doses, or some fraction thereof as appropriate to the accident; or the NRC-approved licensing basis identified in Section.14.9 of the PINGP Updated Safety Analysis Report (USAR).
Technical Specification 5.5.8 "Steam Generator (SG) Program" for PINGP Units 1 and 2 requires that a SG Program be established and implemented to ensure that SG tube integrity is maintained. Tube integrity is maintained by meeting specified performance criteria for structural and leakage integrity consistent with the plant design and licensing bases. Technical Specification 5.5.8.a requires a condition monitoring assessment be performed during each outage, during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met. Technical Specification 5.5.8.d includes provisions regarding the scope, frequency, and methods of SG tube inspections.
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes
3.1.1 3.4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Current Limiting Condition for Operation (LCO) 3.4~19 states, in part:
All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.
Revised LCO 3.4.19 would state, in part:
All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.
Current LCO 3.4.19 ACTION CONDITION A states:
A.
One or more SG tubes satisfying the tube repair criteria and not plugged or repaired in accordance with, the Steam Generator Program.
Revised LCO 3.4.19 ACTION CONDITION A would state:
A.
One or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program.
Current LCO 3.4.19 REQUIRED ACTION A.2 states:
A.2 Plug or repair the affected tube(s) in accordance with the Steam Generator Program.
Revised LCO 3.4.19 REQUIRED ACTION A.2 would state:
A.2 Plug the affected tube(s) in accordance with the Steam Generator Program.
-4
./
Current Surveillance Requirement (SR) 3.4.19.2 states:
SR 3.4.19.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged or repaired in accordance with the Steam Generator Program.
Revised SR 3.4.19.2 would state:
SR 3.4.19.2 Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged or repaired in accordance with the Steam Generator Program.
3.1.2 5.5.8 STEAM GENERATOR (SG) PROGRAM Current TS 5.5.8 STEAM GENERATOR (SG) PROGRAM states, in part:
A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:
Revised TS 5.5.8 STEAM GENERATOR (SG) PROGRAM would state, in part:
A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:
Current TS 5.5.8.a states:
- a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.
Revised TS 5.5.8.a would state:
- a.
Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.
Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.
- 5 Current TS 5.5.8.b.1 states:
- 1.
Structural integrity performance criterion:* All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.8.c.2(c), a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated,to determine if the associated loads contribute significantly to burst or collapse.
In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads. For Unit 2, when alternate repair criteria discussed in Specification 5.5.8.c.2(c) are applied to axially oriented outside diameter stress corrosion cracking indications at the tube support plate locations, the probability that one or more of these indications in an SG will burst under postulated main steam line break '
conditions shall be less than 1 E-02.
Revised TS 5.5.8.b.1 would state:
- 1.
Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and.licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.
In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with
- 6 the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
Current TS S.S.S.b.2 states:
- 2.
Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. For Unit 1, leakage is not to exceed 1 gpm per SG. For Unit 2, leakage. from all sources, excluding the leakage attributed to the degradation associated with implementation of the voltage based repair criteria discussed in Specification S.S.S.c.2(c), is not to exceed 1 gpm per SG.
Revised TS S.S.S.b.2 would state:
- 2.
Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG.
Current TS S.S.S.c states:
- c.
Provisions for SG tube repair criteria:
- 1.
Unit 1 steam generator tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
- 2.
Unit 2 steam generator tubes found by inservice inspection to contain flaws shall be dispositioned as follows:
(a) Depth Based Criteria:
)
(1) Tubes found by inservice inspection containing a flaw in a non sleeved region with a depth equal to or exceeding SO% of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate tube repair criteria discussed in Specification S.S.S.c.2(b) or in Specification S.S.S.c.2(c). If significant general tube thinning occurs, this criterion is reduced to 40% wall penetration.
(2) Tubes found by inservice inspection containing a flaw in the pressure boundary region of any sleeve exceeding 2S% of the nominal sleeve wall thickness shall be plugged.
- 7 (3) Tubes with a flaw in a sleeve to tube joint that occurs in the original tube wall of the joint shall be ph..mged.
(b)
The following F* or EF* Alternate Repair Criteria may be applied to the hot-leg of the tubesheet as an alternative to the depth based criteria in Specification 5.5.8.c.2(a)(1):
(1) F* Criterion: If the bottom of the uppermost hardroll transition in the tubesheet is below the midplane of the tubesheet, then all flaws located below 1.07 inches from the bottom of this uppermost hard roll transition (not including eddy current uncertainty) may be allowed to remain in service provided the tube does not contain any flaws within this 1.07-inch span (not including eddy current uncertainty). This 1.07 -inch span (increased for measurement r
uncertainty) is referred to as the F* region. If flaws are contained within the F* region, the tube shall be plugged or repaired.
(2) EF* Criterion: If the bottom of the uppermost hardroll transition in the tubesheet is above the midplane of the tubesheet but at least 2.0 inches below the top of the secondary face of the tubesheet, then all flaws located below 1.67 inches from the bottom of the uppermost hard roll transition (not including eddy current uncertainty) may be allowed to remain in service provided the tube does not contain any flaws within this 1.67-inch span (not including eddy current uncertainty). This 1.67-inch span (increased for measurement uncertainty) is referred to as the EF*
region. If flaws are contained within the EF* region, the tube shall be plugged or repaired.
(c)
The following Alternate Tube Support Plate Voltage-Based Repair Criteria may be applied as an alternative to the depth based criteria in Specification 5.5.8.c.2(a)(1): For regions of the tube affected by predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of tube support plates the plugging or repair limit is as follows:
(1) If the bobbin voltage associated with the indication is less than or equal to 2.0 Volts, the indication is allowed to remain in service.
(2) If the bobbin voltage associated with the indication is greater than 2.0 Volts, the tube shall be plugged or repaired unless the voltage is less than or equal to the upper voltage repair limit (calculated according to the methodology in GL 95-05 as supplemented) and a rotating pancake coil (or comparable examination technique) does not detect a flaw. In this latter-case,
. the indication may remain in service.
- 8
- 3.
If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits in Specifications 5.5.8.c.2(c)(1) and 5.5.8.c.2(c)(2) above. The mid-cycle repair limits are determined from the following equations:
CL-l1t)
VMLRL = VMURL -
(VURL -
2.0) (
CL Where:
VURL = upper voltage repair limit VLRL = lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL =mid-cycle lower voltage repair limit based on VMURL and time into cycle
.6.t =length of time since last scheduled inspection during which VURL, and VLRL were implemented CL =cycle length (time between two scheduled steam generator inspections)
VSL = structural limit voltage Gr =average growth rate per cycle length NDE =95 percent cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20 percent has been approved by the NRC)
Implementation of these mid-cycle repair limits should follow the same approach as described in Specifications 5.5.8.c.2(c)(1) and 5.5.8.c.2(c)(2) above.
- 9 Note: The upper voltage repair limit is calculated according to the methodology in GL 9S-0S as supplemented.
Revised TS S.S.S.c would state:
- c.
Provisions for SG tube plugging criteria.
Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.
Current TS S.S.S.d states:
- d.
Provisions for SG t~be inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. In tubes repaired by sleeving,' the portion of the original tube wall between the sleeve's joints is not an area requiring re-inspection. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, and d.4 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
Revised TS S.S.S.d would state:
- d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 10 Current TS 5.5.S.d.1 states:
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
Revised TS 5.5.S.d.1 would state:
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
Current TS 5.5.S.d.2 states:
- 2.
For Unit 1 SGs, inspect 100% of the tubes at sequential periods of 144, 10S, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.
Revised TS 5.5.S.d.2 would state:
- 2.
After the first refueling outage following SG installation, inspect each SG at least every 72 effective full power months or at least every third refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, c and d below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated.
The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
a)
After the first refueling outage following SG. installation, inspect 100% of the tubes during the next 144 effective full power months.
- 11 This constitutes the first inspection period; b)
During the next 120 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; c)
During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the third inspection period; and d)
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.
Current TS 5.5.8.d.3 states:
- 3.
For Unit 2 SGs, inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.
(a) During each Unit 2 SG inspection (every 24 effective full power months (EFPM) or one refueling outage (whichever is less>>, all tubes within that SG which have had the F* or EF* criteria applied will be inspected in the F* and EF* regions of the roll expanded region. The region of these tubes below the F* and EF* regions do not need to be inspected, unless there is a sleeve (or portion of a sleeve) that extends below the F* or EF* region.
(b) Implementation of the SG tube alternate repair criteria discussed in Specification 5.5.8.c.2(c) requires a 10b percent bobbin coil inspection for hot leg and cold leg tube support plate intersections down to the lowest.
cold leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20 percent random sampling of tubes inspected over their full length:
(c) SG tube indications left in service as a result of application of the alternate repair criteria discussed in Specification 5.5.8.c.2(c) shall be inspected by bobbin coil probe every 24 EFPM or one refueling outage (whichever is less).
Revised TS 5.5.8.d.3 would be modified as follows:
The current TS 5.5.8.d.3 would be deleted in its entirety. The current TS 5.5.8.dA would be modified as noted below and renumbered as the new TS 5.5.8.d.3.
- 12 Current TS 5.5.8.d.4 states:
. 4.
If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
Revised TS 5.5.8.d.4 (renumbered as new TS 5.5.8.d.3) would state:
- 3.
If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.
Current TS 5.8.8.f states:
- f.
Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service.
For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.
- 2. For Unit 2, the following are approved repair methods:
(a) Alloy 690 tungsten inert gas welded sleeves in accordance with CEN-629-P, Revision 03-P, "Repair of Westinghouse Series 44 and 51 Steam Generator Tubes Using Leak Tight Sleeves".
(b) Hardroll expanding non-sleeved portions of tubes in the tubesheet in order to apply the F* and EF* criteria.
Revised TS 5.5.8.f would be modified as follows:
TS 5.5.8.f would be deleted in its entirety.
- 13 3.1.3 5.6.7 STEAM GENERATOR TUBE INSPECTION REPORT Current TS 5.6.7 states:
- a.
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:
- 1.
The scope of inspections performed on each SG,
- 2.
Active degradation mechanisms found,
- 3.
Nondestructive examination techniques utilized for each degradation mechanism,
- 4.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- 5.
Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,
- 6.
Total number and percentage of tubes plugged or repaired to date,
- 7.
The results of condition monitoring, including the results of tube pulls and in-situ testing,
- 8.
The effective plugging percentage for all plugging and tube repairs in each SG,
- 9.
Repair method utilized and the number of tubes repaired by each repair method, and
- 10.
The results of inspections performed under Specification 5.S.B.d.3 for all tubes that have flaws below the F* or EF* distance, and were not plugged. The report shall include: a) identification of F*
and EF* tubes; and b) location and extent of degradation.
- b.
For implementation of the alternate repair criteria discussed in Specification 5.5.8.c.2(c), notify the NRC staff prior to returning the steam generators to service should any of the following conditions arise:
- 1.
If circumferential crack-like indications are detected at the tube support plate intersections,
- 2.
If indications are identified that extend beyond the confines of the tube support plate, or
- 14
- 3.
If indications are identified at the tube support plant elevations that are attributable to primary water stress corrosion cracking.
Revised TS 5.6.7 STEAM GENERATOR TUBE INSPECTION REPORT would state:
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.
Degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of s,ervice induced indications,
- e.
Number of tubes plugged during the inspection outage for each degradation mechanism,
- f.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing.
3.2 Unit 2 SG Replacement PINGP Unit 2 currently has two Westinghouse model 51 SGs with mill-annealed Alloy 600 tubes.
The NRC staff has approved multiple amendments related to the current SGs at PINGP Unit 2, including alternate repair criteria and alternate repair methods for the SG tubes.
The two replacement SGs are AREVA model 56/19 SGs with thermally treated Alloy 690 tubes.
Other design changes have also been made to the replacement SGs, in order to minimize the potential for degradation.
The licensee is proposing to remove the TS requirements associated with the current alternate repair criteria, and is not proposing any alternate repair criteria for the replacement SGs. These requirements are contained in TS 5.5.8.b.1 (Structural integrity performance criteria),
TS 5.5.8.b.2 (Accident induced leakage performance criteria), TS 5.5.8.c.2 (Provisions for SG tube repair criteria), TS 5.5.8.d (Provisions for SG inspections), TS 5.5.8.f (Provisions for SG tube repair methods), and TS 5.6.7 (SG tube inspection report). The alternate tube repair criteria were supported by analyses that were developed for the licensee's current SGs. These analyses are not applicable to the replacement SGs, since the replacement SGs have different design features than the original SGs; therefore, the NRC staff finds these changes acceptable.
-15 The licensee is proposing to adopt inspection requirements applicable to SGs with thermally treated Alloy 690 tubes for Unit 2 (Le., the material used in their replacement SGs). The NRC staff finds these proposed changes acceptable since the licensee's replacement SGs have thermally treated Alloy 690 tubes rather than mill-annealed Alloy 600 tubes, and the proposed changes are consistent with TSTF-510, Revision 2, which the staff has already reviewed and approved. The performance-based inspection requirements in the TSs require that inspection intervals be established so as to ensure that SG tube integrity is maintained until the next SG inspection.
3.3 Adoption of TSTF-51 0 for Units 1 and 2 In adopting the changes proposed to reflect the replacement of the SGs at Unit 2, the licensee also proposed to adopt the changes specified in TSTF-510, Revision 2, for Units 1 and 2. The changes in TSTF-51 0, Revision 2, reflect the licensee's early implementation experience with their current TSs. The changes in TSTF-51 0, Revision 2, are editorial corrections, changes, and clarifications intended to improve internal consistency, consistency with implementing industry documents, and usability without changing the intent of the requirements. The proposed changes are an improvement to the existing SG inspection requirements and continue to provide assurance that the plant licensing basis will be maintained between SG inspections. The NRC staff approved TSTF-510, Revision 2, for use with the consolidated line item process on October 19, 2011 (ADAMS Accession No. ML112101604). Other than the variations/deviations discussed above (to reflect the replacement of the Unit 2 SGs), and those discussed below, the licensee is not proposing any variations or deviations from the TS changes described in the TSTF-510, Revision 2.
The PINGP TSs utilize different numbering and titles than the Standard Technical Specifications on which TSTF-510, Revision 2, was based. These differences are editorial and do not affect the applicability of TSTF-51 0, Revision 2, to the PINGP TSs. As a result, the NRC staff finds the differences between what was approved for TSTF-51 0, Revision 2, and what is being proposed, are acceptable.
The licensee also proposed one additional change that went beyond the changes made by TSTF-510, Revision 2. In Specification 5.5.S.d, the licensee replaced the words "tube repair criteria" with the words "tube plugging criteria." The replacement of "repair" with "plugging" in Specification 5.5.S.d is an administrative change that makes the wording in the specification consistent. The staff finds this change acceptable.
3.4 Summary In summary, the NRC staff finds that the proposed changes to the SG TS requirements are acceptable since the resultant TSs are consistent with TSTF-510, Revision 2, and reflect the tube material in the replacement SGs. The staff's basis for concluding TSTF-510, Revision 2, is acceptable is documented in the safety evaluation dated October 19, 2011 (ADAMS Accession No. ML112101513). Based on the above, the NRC staff concludes that the proposed changes to the SG tube integrity program TS requirements, including the associated reporting requirements, are acceptable.
- 16
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to installation or use of a facility component 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, 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 amendments involve no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on September 14,2012 (77 FR 56881). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: A. Johnson, NRRlDE Date: July 2,2013
- via memo OFFICE DE/ESGB/BC NAME LPL3-1/PM LPL3-1/LA TWengert SRohrer GKulesa*
DATE 06/17/13 06/17/13 03/18/13 NRR/LPL3-1/BC NRR/LPL3-1/PM RCarlson TWengert 07/02113 07/01/13 OGC INLO