ML14042A223
| ML14042A223 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/27/2014 |
| From: | Balwant Singal Plant Licensing Branch IV |
| To: | Flores R Luminant Generation Co |
| George A | |
| References | |
| TAC MF2651, TAC MF2652 | |
| Download: ML14042A223 (41) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Rafael Flores Senior Vice President and Chief Nuclear Officer Attention: Regulatory Affairs Luminant Generation Company LLC P.O. Box 1002 Glen Rose, TX 76043 February 27, 2014
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: ADOPTION OF TSTF-51 0, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION" (TAC NOS. MF2651 AND MF2652)
Dear Mr. Flores:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 161 to Facility Operating License No. NPF-87 and Amendment No. 161 to Facility Operating License No. NPF-89 for Comanche Peak Nuclear Power Plant, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated August 29, 2013, as supplemented by letter dated February 19, 2014.
The amendments revise Technical Specification (TS) 3.4.17, "Steam Generator (SG) Tube Integrity," TS 5.5.9, "Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program,"
and TS 5.6.9, "Unit 1 Model D76 and Unit 2 Model D5 Steam Generator Tube Inspection Report," in accordance with Technical Specification Task Force (TSTF) change traveler TSTF-510, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," as part of the consolidated line item improvement process.
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Docket Nos. 50-445 and 50-446
Enclosures:
- 1. Amendment No. 161 to NPF-87
- 2. Amendment No. 161 to NPF-89
- 3. Safety Evaluation cc w/encls: Distribution via Listserv Sincerely, betO.._;-~1~~
Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 LUMINANT GENERATION COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-445 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 161 License No. NPF-87
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Luminant Generation Company LLC dated August 29, 2013, as supplemented by letter dated February 19, 2014, 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, as amended, 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 license 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 1 0 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 Facility Operating License No. NPF-87 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 161 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Luminant Generation Company LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan as indicated in the attachment to this license amendment.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.
Attachment:
Changes to the Facility Operating License No. NPF-87 and Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 27,2014
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 LUMINANT GENERATION COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 2 DOCKET NO. 50-446 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 161 License No. NPF-89
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Luminant Generation Company LLC dated August 29, 2013, as supplemented by letter dated February 19, 2014, 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, as amended, 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 license 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 Facility Operating License No. NPF-89 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 161 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Luminant Generation Company LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.
Attachment:
Changes to the Facility Operating License No. NPF-89 and Technical Specifications FOR THE NUCLEAR REGULA TORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: February 27, 2014
ATTACHMENT TO LICENSE AMENDMENT NO. 161 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 161 TO FACILITY OPERATING LICENSE NO. NPF-89 DOCKET NOS. 50-445 AND 50-446 Replace the following pages of the Facility Operating License Nos. NPF-87 and NPF-89, and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Facility Operating License No. NPF-87 REMOVE INSERT 3
3 Facility Operating License No. NPF-89 REMOVE INSERT 3
3 Technical Specifications REMOVE INSERT 3.4-44 3.4-44 3.4-45 3.4-45 5.5-5 5.5-5 5.5-6 5.5-6 5.5-7 5.5-7 5.5-8 5.5-8 5.5-9 5.5-9 5.5-10 5.5-10 5.5-11 5.5-11 5.5-12 5.5-12 5.5-13 5.5-13 5.5-14 5.5-14 5.5-15 5.5-15 5.5-16 5.5-16 5.5-17 5.5-17 5.5-18 5.6-5 5.6-5
Unit 1 (3)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, and described in the Final Safety Analysis Report, as supplemented and amended; (4)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission=s regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission *now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Luminant Generation Company LLC is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal through Cycle 13 and 3612 megawatts thermal starting with Cycle 14 in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 161 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Luminant Generation Company LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Amendment No. 161
Unit 2 (3)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, and described in the Final Safety Analysis Report, as supplemented and amended; (4)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
Luminant Generation Company LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission=s regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Luminant Generation Company LLC is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal through Cycle 11 and 3612 megawatts thermal starting with Cycle 12 in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 161 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Luminant Generation Company LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Antitrust Conditions DELETED Amendment No. 161
3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
SG Tube Integrity 3.4.17 All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.
APPLICABILITY:
MODES 1, 2, 3, and 4 ACTIONS
N 0 T E S---------------------------------------------------------
Separate Condition entry is allowed for each SG tube.
CONDITION A. One or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program.
REQUIRED ACTION A.1 Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube
. inspection.
AND A.2 Plug the affected tube(s) in accordance with the Steam Generator Program.
COMPLETION TIME 7 days Prior to entering MODE 4 following the next refueling outage or SG tube inspection COMANCHE PEAK-UNITS 1 AND 2 3.4-44 Amendment No. 150, 156, 161
ACTIONS (continued)
CONDITION REQUIRED ACTION SG Tube Integrity 3.4.17 COMPLETION TIME B. Required Action and associated Completion Time of Condition A not met.
B.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B.2 Be in MODE 5.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.
SURVEILLANCE REQUIREMENTS SR 3.4.17.1 SR 3.4.17.2 SURVEILLANCE Verify SG tube integrity in accordance with the Steam Generator Program.
Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program.
FREQUENCY In accordance with the Steam Generator Program Prior to entering MODE 4 following a SG tube inspection COMANCHE PEAK-UNITS 1 AND 2 3.4-45 Amendment No. 150, 156, 161
5.5 Programs and Manuals (continued)
Programs and Manuals 5.5 5.5.9 Unit 1 Model 076 and Unit 2 Model 05 Steam Generator (SG) Program 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:
- 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. 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 1 gpm per SG.
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
COMANCHE PEAK-UNITS 1 AND 2 5.5-5 Amendment No. 4-&G;- 161
5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Unit 1 Model 076 and Unit 2 Model 05 Steam Generator (SG) Program (continued)
- 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.
- 1.
The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria:
- a.
For Unit 2 only, tubes with service-induced flaws located greater than 14.01 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 14.01 inches below the top of the tubesheet shall be plugged upon detection.
- d.
Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. For Unit 1, 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. For Unit 2, 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 14.01 inches below the top of the tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side 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 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.
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
- 2.
For the Unit 2 model 05 steam generators (Alloy 600 thermally treated) after the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other 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 COMANCHE PEAK-UNITS 1 AND 2 5.5-6 Amendment No. 4-SG-; 164, 168, 161
5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program (continued) inspection period as defined in a, b, and c 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 120 effective full power months. This constitutes the first inspection period;
- b.
During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and
- c.
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.
- 3.
For the Unit 1 model Delta-76 steam generators (Alloy 690 thermally treated) 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 COMANCHE PEAK-UNITS 1 AND 2 5.5-7 Amendment No. -+aG; 194, 198, 161
5.5 Programs and Manuals Programs and Manuals 5.5 5.5.9 Unit 1 Model D76 and Unit 2 Model D5 Steam Generator (SG) Program (continued) 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 1 00% of the tubes during the next 144 effective full power months. 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 1 00% 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.
- 4.
For Unit 1, 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 indications shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). For Unit 2, if crack indications are found in any SG tube from 14.01 inches below the top of the tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indications 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.
- e.
Provisions for monitoring operational primary to secondary LEAKAGE.
COMANCHEPEAK-UNITS1AND2 5.5-8 Amendment No. +W;-161
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 5.5.11 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:
- a.
Identification of a sampling schedule for the critical variables and control points for these variables;
- b.
Identification of the procedures used to measure the values of the critical variables;
- c.
Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage;
- d.
Procedures for the recording and management of data;
- e.
Procedures defining corrective actions for all off control point chemistry conditions; and
- f.
A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.
Ventilation Filter Testing Program (VFTP)
A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2 and in accordance with Regulatory Guide 1.52, Revision 2, ANSI/ASME N509-1980, ANSI/ASME N510-1980, and ASTM 03803-1989.
NOTE--------------------------------------------------
ANSI/ASME N510-1980, ANSI/ASME N509-1980, and ASTM 03803-1989 shall be used in place of ANSI 510-1975, ANSI/ASME N509-1976, and ASTM 03803-1979 respectively in complying with Regulatory Guide 1.52, Revision 2.
- a.
Demonstrate for each of the ESF systems that an in place test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1.0% for Primary Plant Ventilation System - ESF Filtration units and
< 0.05% for all other units when tested in accordance with Regulatory COMANCHE PEAK-UNITS 1 AND 2 5.5-9 Amendment No. 4W;- 161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)
Guide 1.52, Revision 2, and ANSI/ASME N510-1980 at the system flowrate specified below+/- 10%.
ESF Ventilation System Control Room Emergency filtration unit Control Room Emergency pressurization unit Primary Plant Ventilation System-ESF filtration unit Flowrate 8,000 CFM 800 CFM 15,000 CFM
- b.
Demonstrate for each of the ESF systems that an in place test of the charcoal adsorber shows a penetration and system bypass < 1.0% for Primary Plant Ventilation System-ESF Filtration units and < 0.05% for all other units when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI!ASME N510-1980 at the system flowrate specified below+/- 10%.
ESF Ventilation System Control Room Emergency filtration unit Control Room Emergency pressurization unit Primary Plant Ventilation System - ESF filtration unit Flow rate 8,000 CFM 800 CFM 15,000 CFM
- c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of
~ 30°C and greater than or equal to the relative humidity specified below.
ESF Ventilation Systems Control Room Emergency filtration unit Control Room Emergency pressurization unit Primary Plant Ventilation System-ESF filtration unit Penetration 0.5%
0.5%
2.5%
RH 70%
70%
70%
- d.
Demonstrate at least once per 18 months for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in COMANCHE PEAK-UNITS 1 AND 2 5.5-10 Amendment No. 469;- 161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 5.5.12 Ventilation Filter Testing Program (VFTP) (continued) accordance with Regulatory Guide 1.52, Revision 2, and ANSI/ASME N51 0-1980 at the system flowrate specified below+/- 10%
ESF Ventilation System Control Room Emergency filtration unit Control Room Emergency pressurization unit Primary Plant Ventilation System - ESF filtration unit.
Delta P 8.0 in WG 9.5 inWG 8.5 in WG Flowrate 8000 CFM 800 CFM 15000 CFM
- e.
Demonstrate at least once per 18 months that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ANSI/AS ME N51 0-1980.
ESF Ventilation System Control Room Emergency pressurization unit Primary Plant Ventilation System - ESF filtration unit Wattage 10 +/- 1 kW 100 +/- 5 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Gaseous Waste Processing System, the quantity of radioactivity contained in each Gas Decay Tank, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.
The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure," Revision 0, July 1981. The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures," Revision 2, July 1981.
The program shall include:
- a.
The limits for concentrations of hydrogen and oxygen in the Gaseous Waste Processing System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);
COMANCHE PEAK-UNITS 1 AND 2 5.5-11 Amendment No. 4-eG;-- 161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 5.5.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)
- b.
A surveillance program to ensure that the quantity of radioactivity contained in each Gas Decay Tank is less than the amount that would result in a whole body exposure of~ 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and
- c.
A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.
- d.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.
Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:
- a.
Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
- 1.
an API gravity or an absolute specific gravity within limits,
- 2.
a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
- 3.
a clear and bright appearance with proper color or a water and sediment content within limits.
- b.
Within 31 days following addition of the new fuel oil to the storage tanks, verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits for ASTM 2D fuel oil, and
- c.
Total particulate concentration of the fuel oil is s 1 0 mg/1 when tested every 31 days.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program.
COMANCHE PEAK-UNITS 1 AND 2 5.5-12 Amendment No. ~161
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 5.5.15 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
- a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
- 1.
a change in the TS incorporated in the license; or
- 2.
a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
- c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
- d.
Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e) and exemptions thereto.
Safety Function Determination Program (SFDP)
- a.
This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.
The SFDP shall contain the following:
- 1.
Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
- 2.
Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
- 3.
Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
- 4.
Other appropriate limitations and remedial or compensatory actions.
COMANCHE PEAK-UNITS 1 AND 2 5.5-13 Amendment No.~ 161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 5.5.16 Safety Function Determination Program (SFDP) (continued)
- b.
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
- 1.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
- 2.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
- 3.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
- c.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
Containment Leakage Rate Testing Program
- a.
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program, dated September, 1995" as modified by the following exceptions:
- 1.
The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.
- 2.
The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.
- 3.
NEI 94 1995, Section 9.2.3: The first Type A Test performed after the December 7, 1993 Type A Test (Unit 1) and the December 1, 1997 Type A Test (Unit 2) shall be performed no later than December 15, 2008 (Unit 1) and December 9, 2012 (Unit 2)."
COMANCHE PEAK-UNITS 1 AND 2 5.5-14 Amendment No. 4-&G;- 161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 5.5.17 5.5.18 Containment Leakage Rate Testing Program (continued)
- b.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48.3 psig.
- c.
The maximum allowable containment leakage rate, La, at P a* shall be 0.10%
of containment air weight per day.
- d.
Leakage rate acceptance criteria are:
- 1.
Containment leakage rate acceptance criteria is :5; 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and :5; 0. 75 La for Type A tests:
- 2.
Air lock testing acceptance criteria are:
- i.
Overall air lock leakage rate is :5; 0.05 La when tested at~ P a*
ii.
For each door, leakage rate is :5; 0.01 La when pressurized to
~ Pa.
- e.
The provision of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program, with the exception of the containment ventilation isolation valves.
- f.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Technical Requirements Manual (TRM)
The TRM contains selected requirements which do not meet the criteria for inclusion in the Technical Specification but are important to the operation of CPNPP. Much of the information in the TRM was relocated from the TS.
Changes to the TRM shall be made under appropriate administrative controls and reviews. Changes may be made to the TRM without prior NRC approval provided the changes do not require either a change to the TS or NRC approval pursuant to 10 CFR 50.59. TRM changes require approval of the Plant Manager.
Configuration Risk Management Program (CRMP)
The Configuration Risk Management Program (CRMP) provides a proceduralized risk-informed assessment to manage the risk associated with equipment inoperability.
. The program applies to technical specification structures, systems, or components for COMANCHE PEAK-UNITS 1 AND 2 5.5-15 AmendmentNo. 4-eG;-161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.18 5.5.19 5.5.20 Configuration Risk Management Program (CRMP) (continued) which a risk-informed Completion Time has been granted. The program shall include the following elements:
- a.
Provisions for the control and implementation of a Level 1, at-power, internal events PRA-informed methodology. The assessment shall be capable of evaluating the applicable plant configuration.
- b.
Provisions for performing an assessment prior to entering the LCO Action for preplanned activities.
- c.
Provisions for performing an assessment after entering the LCO Action for unplanned entry into the LCO Action.
- d.
Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in the LCO Action.
- e.
Provisions for considering other applicable risk significant contributors such as Level 2 issues, and external events, qualitatively or quantitatively.
Battery Monitoring and Maintenance Program This Program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer for the following:
- a.
Actions to restore battery cells with float voltage< 2.13 V, and
- b.
Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates.
Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Filtration System (CREFS), CRE occupants can control the reactor safety under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:
- a.
The definition of the CRE and the CRE boundary.
COMANCHE PEAK-UNITS 1 AND 2 5.5-16 Amendment No. 4-W,- 161
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.20 Control Room Envelope Habitability Program (continued)
- b.
Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
- c.
Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"
Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:
- 1.
C.- Section 4.3.2 "Periodic CRH Assessment" from NEI 99-03 Revision 1 will be used as input to a site specific Self Assessment procedure.
- 2.
C.1.2 - No peer reviews are required to be performed.
- d.
Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREFS, operating at the flow rate required by the VFTP, at a Frequency of 18 months on a STAGGERED TEST BASIS.
The results shall be trended and used as part of the 18 month assessment of the CRE boundary.
- e.
The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
- f.
The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
COMANCHE PEAK-UNITS 1 AND 2 5.5-17 Amendment No. +W;- 161
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.21 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
- a.
The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.
- b.
Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI-04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
- c.
The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
COMANCHE PEAK-UNITS 1 AND 2 5.5-18 Amendment No. 161
5.6 Reporting Requirements Reporting Requirements 5.6 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)
- 1.
WCAP-14040-NP-A; "Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
- c.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 Not used 5.6.8 PAM Report When a report is required by the required actions 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.
5.6.9 Unit 1 Model D76 and Unit 2 Model D5 Steam Generator Tube Inspection Report 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 5.5.9, 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 precentage of tubes plugged to date, and the effective plugging percentage in each steam generator,
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing,
- h.
For Unit 2, the primary to secondary leakage rate observed in each SG (if it is notpractical to assign the leakage to an individual SG, the entire primary to COMANCHE PEAK-UNITS 1 AND 2 5.6-5 Amendment No. 4-eQ, 164, 168, 161
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 161 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 161 TO FACILITY OPERATING LICENSE NO. NPF-89 LUMINANT GENERATION COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT. UNITS 1 AND 2 DOCKET NOS. 50-445 AND 50-446
1.0 INTRODUCTION
By application dated August 29, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13255A021 ), and as supplemented by letter dated February 19, 2014, Luminant Generation Company LLC (the licensee) requested changes to the Technical Specifications (TSs) for Comanche Peak Nuclear Power Plant (CPNPP), Units 1 and 2, to adopt U.S. Nuclear Regulatory Commission (NRC)-approved Revision 2 toTS Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-51 0, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection" (ADAMS Accession No. ML110610350).
The proposed changes would modify TS 3.4.17, "Steam Generator (SG) Tube Integrity,"
TS 5.5.9, "Unit 1 Model D76 and Unit 2 Model 05 Steam Generator (SG) Program," and TS 5.6.9, "Unit 1 Model 076 and Unit 2 Model 05 Steam Generator Tube Inspection Report,"
and include TS Bases changes that summarize and clarify the purpose of the TS. The specific changes concern steam generator (SG) inspection periods, and address applicable administrative changes and clarifications. In its application, the licensee confirmed applicability of TSTF-51 0, Revision 2, and the NRC staff's Model Safety Evaluation for TSTF-51 0, Revision 2 (ADAMS Accession No. ML112101513), dated October 19, 2011, as identified in the NRC Federal Register Notice of Availability, dated October 27, 2011 (76 FR 66763) to CPNPP, Units 1 and 2.
The current STS requirements in the above specifications were established in May 2005 with the NRC staff's approval of TSTF Traveler TSTF-449, Revision 4, "Steam Generator Tube Integrity" (NRC Federal Register Notice of Availability [70 FR 24126]). The TSTF-449 changes to the STS incorporated a new, largely performance-based approach for ensuring the integrity of the SG tubes is maintained. The performance-based requirements were supplemented by prescriptive requirements relating to tube inspections and tube repair limits to ensure that conditions adverse to quality are detected and corrected on a timely basis. As of September 2007, the TSTF-449, Revision 4, changes were adopted in the plant TS for all pressurized-water reactors (PWRs).
The proposed changes in TSTF-510, Revision 2, reflect licensees' early implementation experience with respect to the TSTF-449, Revision 4. 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. 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.
In its application, the licensee stated that TSTF-51 0, Revision 2, TS Section 5.5.9.d.2 regarding inspections after the first refueling outage following SG installation, has been divided in the CPNPP TS as Section 5.5.9.d.2 for the CPNPP Unit 2 model 05 steam generators and Section 5.5.9.d.3 for the CPNPP Unit 1 model Oelta-76 steam generators. Additionally, TSTF-510 TS Section 5.5.9.d.3 regarding indications of cracking in any SG tube is CPNPP TS Section 5.5.9.d.4.
The February 19, 2014, supplement did not expand the scope of the application as originally noticed, and did not change the NRC staff's initial proposed finding of no significant hazards consideration.
2.0 REGULATORY EVALUATION
The SG tubes in PWRs have a number of important safety functions. These tubes are an integral part of the Reactor Coolant Pressure Boundary (RCPB) and, as such, are relied upon to maintain primary system pressure and inventory. As part of the RCPB, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system and are relied upon to isolate the radioactive fission products in the primary coolant from the secondary system. In addition, the SG tubes are relied upon to maintain their integrity to be consistent with the containment objectives of preventing uncontrolled fission product release under conditions resulting from core damage during severe accidents.
Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.36, "Technical specifications,"
establishes the requirements related to the content of the TS. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five categories related to station operation:
(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. LCOs and accompanying action statements and SRs in the CPNPP TS relevant to SG tube integrity are in TS 3.4.17, "Steam Generator (SG) Tube Integrity,"
TS 5.5.9, "Unit 1 Model 076 and Unit 2 Model 05 Steam Generator (SG) Program," and TS 5.6.9, "Unit 1 Model 076 and Unit 2 Model 05 Steam Generator Tube Inspection Report."
The regulations in 10 CFR 50.36(c)(5) define administrative controls as "the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure the operation of the facility in a safe manner." Programs established by the licensee to operate the facility in a safe manner, including the SG Program, are listed in the administrative controls section of the TS. CPNPP's SG Program is defined in Specification 5.5.9, while the reporting requirements relating to implementation of the SG Program are described in Specification 5.6.9.
TS 5.5.9 requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. Specification 5.5.9.a requires that a condition monitoring assessment be performed during each outage in which the SG tubes are inspected, to confirm that the performance criteria are being met. SG tube integrity is maintained by meeting the performance criteria specified in TS 5.5.9.b for structural and leakage integrity, consistent with the plant design and licensing basis. Specification 5.6.9 includes provisions regarding the scope, frequency, and methods of SG tube inspections. These provisions require that the inspections be performed with the objective of detecting flaws of any type that: (1) may be present along the length of a tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet; and (2) may satisfy the applicable tube plugging criteria.
The applicable tube plugging criteria, specified in TS 5.5.9.c, are that tubes found during inservice inspections (ISis) to contain flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness shall be plugged, unless the tubes are permitted to remain in service through application of the alternate plugging criteria provided in TS 5.5.9.c.
3.0 TECHNICAL EVALUATION
3.1 Changes and Variations from TSTF-51 0, Revision 2 By letter dated August 29, 2013, the licensee, proposed the following variations from the TS changes described in TSTF-51 0, Revision 2, or the applicable parts of the NRC staff's model safety evaluation dated October 27, 2011. The letter states, The CPNPP TS utilize different numbering and titles than the Standard Technical Specifications on which TSTF-51 0 was based. Specifically, the CPNPP TS title for Section 5.5.9 is "Unit 1 Model 076 and Unit 2 Model 05 Steam Generator (SG) Program," to denote that the CPNPP SGs are different models. TSTF-510 TS Section 5.5.9.d.2 for inspections after the first refueling outage following SG installation has been divided in the CPNPP TS as Section 5.5.9.d.2 for the CPNPP Unit 2 model 05 steam generators and Section 5.5.9.d.3 for the CPNPP Unit 1 model Oelta-76 steam generators. TSTF-510 TS Section 5.5.9.d.3 is CPNPP TS Section 5.5.9.d.4 for crack indications. These differences are administrative and do not affect the applicability of TSTF-510 to the CPNPP TS.
The proposed change corrects an administrative inconsistency in TSTF-510, Paragraph d.2 of the Steam Generator Tube Inspection Program. In Section 2.0, "Proposing Change," TSTF-51 0 states that references to "tube repair criteria" in Paragraph is revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted, "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 the location and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated" (Emphasis added).
Luminant Power does not have an approved tube repair criteria. Therefore the sentence is revised to state "tube plugging criteria", consistent with TSTF-51 0, Revision 2.
3.2 Proposed TS Changes
Each proposed change along with the licensee's applicable variations to the TS is described below, followed by the NRC staff's assessment of the change.
3.2.1 TS 5.5.9: "Steam Generator (SG) Program" The last sentence of the introductory paragraph currently states:
In addition, the Steam Generator Program shall include the following provisions:
The change would delete the word "provisions" such that the sentence would state:
In addition, the Steam Generator Program shall include the following:
The basis for this change is that subsequent paragraphs in TS 5.5.9 start with "Provisions for... " and the word "provisions" in the introductory paragraph is duplicative.
The NRC staff has reviewed the proposed change and agrees that the word, "provisions" in the introductory paragraph is duplicative. The NRC staff concludes that the change is editorial in nature and, therefore, is acceptable.
3.2.2 Paragraph 5.5.9.b.1: "Structural integrity performance criterion" The first sentence of paragraph 5.5.9.b.1 currently states:
All in-service SG 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.
The revised first sentence of paragraph 5.5.9.b.1 would state:
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.
The basis for the change is that this sentence inappropriately includes anticipated transients in the description of normal operating conditions.
The NRC staff agrees the current wording is incorrect and that anticipated transients should be differentiated from normal operating conditions. Therefore, the NRC staff concludes that the change is acceptable.
3.2.3. Paragraph 5.5.9.c, "Provisions for SG tube repair criteria,"
Paragraph 5.5.9.d, "Provisions for SG tube inspections," and LCO 3.4.17, "Steam Generator (SG) Tube Integrity" The licensee proposed to change all references in paragraph 5.5.9.c, 5.5.9.d, and LCO 3.4.17 from "tube repair criteria" to "tube plugging criteria." This change is intended to be consistent with the treatment of SG tube repair throughout Specification 5.5.9, since CPNPP, Units 1 and 2 do not have an approved tube repair criteria.
The NRC staff concludes that the proposed change provides a more accurate description of the criteria and, therefore, adds clarity to the specification. This is because one of two actions must be taken when the tube plugging criteria are exceeded. One action is to remove the tube from service by plugging the tube at both tube ends, and the alternative action is to repair the tube.
However, per the application, the licensee does not have an approved tube repair criteria.
Therefore, the NRC staff concludes that the change is acceptable.
3.2.4 Paragraph 5.5.9.d: "Provisions for SG tube inspections" The licensee proposed to change the term "An assessment of degradation" to "A degradation assessment" in paragraph 5.5.9.d to be consistent with the terminology used in industry program documents.
The NRC staff agrees that the terminology should be consistent and concludes that the change is acceptable.
3.2.5 Paragraph 5.5.9.d.1 Current paragraph 5.5.9.d.1 states:
Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
Revised paragraph 5.5.9.d.1 would state:
Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.
The change would replace "SG replacement" with "SG installation." The basis for the change is that it SG installation encompasses SG replacement.
The NRC staff concludes that the change involves no change in the inspection requirement, and will be more generic terminology from plant to plant and is acceptable.
3.2.6 Paragraphs 5.5.9.d.2 and 5.5.9.d.3, "Unit 1 Model 076 and Unit 2 Model 05 Steam Generator (SG) Program" Current paragraph 5.5.9.d.2 states:
For the Unit 2 model 05 steam generators (Alloy 600 thermally treated) inspect 100% of the tubes at sequential periods of 120, 90, 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 48 effective full power months or two refueling outages (whichever is less) without being inspected.
Revised paragraph 5.5.9.d.2 would state:
For the Unit 2 model 05 steam generators (Alloy 600 thermally treated) after the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other 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, and c 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 120 effective full power months. This constitutes the first inspection period;
- b.
During the next 96 effective full power months, inspect 100% of the tubes.
This constitutes the second inspection period; and
- c.
During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.
Current paragraph 5.5.9.d.3 states:
For the Unit 1 model Delta-76 steam generators (Alloy 690 thermally treated) inspect 100% of the tubes at sequential periods of 144, 108, 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 paragraph 5.5.9.d.3 would state:
For the Unit 1 model Delta-76 steam generators (Alloy 690 thermally treated) 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 1 00% of the tubes during the next 144 effective full power months. 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 1 00% of the tubes every 72 effective full power months. This constitutes the fourth and subsequent inspection periods.
The proposed changes toTS paragraphs 5.5.9.d.2 and 5.5.9.3 are similar for each of the tube alloy types used in domestic SGs, but with differences that reflect the improved resistance of alloy 690 thermally treated (TT) to stress-corrosion cracking relative to both alloy 600 mill annealed (MA) and alloy 600 TT. These differences include progressively larger maximum inspection interval requirements and sequential inspection periods (during which 100 percent of the tubes must be inspected) for alloy 600 MA, 600 TT, and alloy 690 TT tubes, respectively. In addition, because of the longer maximum inspection intervals allowed for alloy 600 TT and 690 TT tubes, paragraphs 5.5.9.d.2 and 5.5.9.d.3 include a restriction on the distribution of sampling over each sequential inspection period for alloy 600 TT and 690 TT tubes that is not included for alloy 600 MA tubes.
The licensee proposes to move the first two sentences of paragraphs 5.5.9.d.2 and 5.5.9.d.3 to the end of the paragraph and make editorial changes to improve clarity. The NRC staff concludes that these changes are clarifying in nature and do not change the current intent of these two sentences. However, the licensee's application also includes changes to when inspections are performed for each SG tube type as follows:
CPNPP Unit 1 SGs (690 TT tubes)
The second inspection period would be revised from 108 to 120 effective full power months (EFPM).
The third inspection period would be revised from 72 to 96 EFPM.
The fourth and subsequent inspection periods would be revised from 60 to 72 EFPM.
CPNPP Unit 2 SGs (600 TT tubes)
The second inspection period would be revised from 90 to 96 EFPM.
The third and subsequent inspection periods would be revised from 60 to 72 EFPM.
The licensee characterizes these changes as marginal increases for consistency with typical fuel cycle lengths that better accommodate the scheduling of inspections. The NRC staff observes that this is clearly the case for plants operating with 18-or 36-month inspection intervals (one or two fuel cycles, respectively). With these intervals, the last scheduled inspection during the first inspection period would coincide with the end of the first, third, and subsequent inspection periods.
The proposed changes would generally increase the number of inspections for CPNPP Unit 1 in each of the third and subsequent inspection periods by up to one additional inspection, for CPNPP Unit 2 in each of the second and subsequent inspection periods by up to one additional inspection. This could reduce the required average minimum sample size during these periods.
However, inspection sample sizes will continue to be subject to paragraph 5.5.9.d which states that in addition to meeting the requirements of paragraphs 5.5.9.d.1 thru d.4, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure SG tube integrity is maintained until the next scheduled inspection. Therefore, the NRC staff concludes that with the proposed changes to the length of the second and subsequent inspection periods, compliance with the SG program requirements in Specification 5.5.9.d will continue to ensure both adequate inspection scopes and tube integrity.
For each inspection period, paragraphs 5.5.9.d.2 and 5.5.9.d.3 currently require that at least 50 percent of the tubes be inspected by the refueling outage nearest to the mid-point of the inspection period and the remaining 50 percent by the refueling outage nearest the end of the inspection period. The NRC staff notes that if there are not an equal number of inspections in the first half and second half of the inspection period, the average minimum sampling requirement may be markedly different for inspections in the first half of the inspection period compared to those in the second half, even when there are uniform intervals between each inspection.
For example, a plant in the 120 EFPM inspection period with a scheduled 36-month interval (two fuel cycles) between each inspection would currently be required to inspect 50 percent of the tubes by the refueling outage nearest the midpoint of the inspection which would be the third refueling outage in the period, 6 months before the mid-point (assuming an inspection was performed at the very end of the 144 EFPM inspection period). However, since no inspection is scheduled for that outage, then the full 50 percent sample must be performed during the inspection scheduled for the second refueling outage in the period. Two inspections would be scheduled to occur in the second half of the inspection period, at 72 and 108 months into the inspection period. Thus, the current sampling requirement could be satisfied by performing a 25 percent sample during each of these inspections or other combinations of sampling (e.g.,
1 0 percent during one and 40 percent in the other) totaling 50 percent.
Based on operational experience, the NRC staff concludes that there is no longer a basis to require the minimum initial sample size to vary so much from inspection to inspection. The licensee proposes to revise this requirement such that the minimum sample size for a given inspection in a given inspection period is 1 00 percent divided by the number of scheduled inspections during that inspection period. For the above example, the proposed change would result in a uniform initial minimum sample size of 33.3 percent for each of the three scheduled inspections during the inspection period. The NRC staff concludes this proposed revision to be an improvement to the existing requirement since it provides a more consistent minimum initial sampling requirement.
The proposed changes to paragraphs 5.5.9.d.2 and 5.5.9.d.3 include two new sentences addressing the prorating of required tube sample sizes 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. For example, new information from another similar plant becomes available indicating the potential for circumferential cracking at a specific location on the tube. Previous degradation assessments had not identified the potential for this type of degradation at this location. Thus, previous inspections of this location had not been performed with a technique capable of detecting circumferential cracks. However, now that the potential for circumferential cracking has been identified at this location, paragraph 5.5.9.d requires a method of inspection to be performed with the objective of detecting circumferential cracks which may be present at this location and that may satisfy the applicable tube plugging criteria.
Furthermore, if this inspection is performed for the first time during the third of four SG inspections scheduled for one of the inspection periods, then paragraphs 5.5.9.d.2 and 5.5.9.d.3 currently do not specify whether this location needs to be 100 percent inspected by the end of the inspection period, or whether a prorated approach may be taken. The NRC staff addressed this question in Issue 1 of NRC Regulatory Information Summary (RIS) 2009-04, "Steam Generator Tube Inspection Requirements," dated April 3, 2009 (ADAMS Accession No. ML083470557), as follows:
Issue 1: A licensee may identify a new potential degradation mechanism after the first inspection in a sequential period. If this occurs, what are the expectations concerning the scope of examinations for this new potential degradation mechanism for the remainder of the period (e.g., do 100 percent of the tubes have to be inspected by the end of the period or can the sample be prorated for the remaining part of the period)?
The TS contain requirements that are a mixture of prescriptive and performance-based elements. Paragraph "d" of these requirements indicates that the inspection scope, inspection methods, and inspection intervals shall be sufficient to ensure that SG tube integrity is maintained until the next SG inspection.
Paragraph "d" is a performance-based element because it describes the goal of the inspections but does not specify how to achieve the goal. However, paragraph "d.2" is a prescriptive element because it specifies that the licensee must inspect 100 percent of the tubes at specified periods.
If an assessment of degradation performed after the first inspection in a sequential period results in a licensee concluding that a new degradation mechanism (not anticipated during the prior inspections in that period) may potentially occur, the scope of inspections in the remaining portion of the period should be sufficient to ensure SG tube integrity for the period between inspections.
In addition, to satisfy the prescriptive requirements of paragraph "d.2" that the licensee must inspect 100 percent of the tubes within a specified period, a prorated sample for the remaining portion of the period is appropriate for this potentially new degradation mechanism. This prorated sample should be such that if the licensee had implemented it at the beginning of the period, the TS requirement for the 100 percent inspection in the entire period (for this degradation mechanism) would have been met. A prorated sample is appropriate because {1) the licensee would have performed the prior inspections in this sequential period consistently with the requirements, and (2) the scope of inspections must be sufficient to ensure that the licensee maintains SG tube integrity for the period between inspections.
The NRC staff concludes that relocation of information in sentences 3 and 4 as described above clarifies the existing requirement consistently with the NRC staff's position from RIS 2009-04 quoted above and is, therefore, acceptable.
The proposed fifth sentence in paragraphs 5.5.9.d.2 and 5.5.9.d.3 states, "Each inspection period defined below may be extended up to 3 effective full power months (EFPM) 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." Allowing extension of the inspection periods by up to an additional 3 EFPMs potentially impacts the average tube inspection sample size to be implemented during a given inspection in that period. For example, if four SG inspections are scheduled to occur within the nominal 120-EFPM period, then the minimum sample size for each of the four inspections could average as little as 25 percent of the tube population. If a fifth inspection can be included within the period by extending the period by 3 EFPM, then the minimum sample size for each of the five inspections could average as little as 20 percent of the tube population. Since the subsequent period begins at the end of the included SG inspection outage, the proposed change does not impact the required frequency of SG inspection.
Required tube inspection sample sizes are also subject to the performance-based requirement in paragraph 5.5.9.d, which states, in part, that in addition to meeting the requirements of paragraphs 5.5.9.d.1, d.2, d.3, and d.4, "the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next scheduled SG inspection." This requirement remains unchanged under the proposal. The NRC staff concludes the proposed fifth sentence involves only a relatively minor relaxation to the existing sampling requirements in paragraphs 5.5.9.d.2 and 5.5.9.d.3. However, the performance-based requirements in 5.5.9.d ensure that adequate inspection sampling will be performed to ensure tube integrity is maintained. The NRC staff concludes that the proposed change is acceptable.
Finally, the first sentence of the proposed revision to paragraphs 5.5.9.d.2 and 5.5.9.d.3 replaces the last sentence of the current paragraphs 5.5.9.d.2 and 5.5.9.d.3. This sentence establishes the minimum allowable SG inspection frequency. This minimum inspection frequency is unchanged from the current requirement in CPNPP technical specifications. The NRC staff concludes that the wording changes in the sentence are of an editorial and clarifying nature and are not material, such that the current intent of the requirement is unchanged. Thus, the NRC staff concludes the first sentence of proposed paragraphs 5.5.9.d.2 and 5.5.9.d.3 is acceptable.
3.2. 7 Change to Paragraph 5.5.9.d.4:
The first two sentences of paragraph 5.5.9.d.4 currently state:
For Unit 1, if crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indications shall not exceed 24 effective full power months or one refueling outage (whichever is less). For Unit 2, if crack indications are found in any SG tube from 14.01 inches below the top of the tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, then the next inspection for each SG for the degradation mechanism that caused the crack indications shall not exceed 24 effective full power months or one refueling outage (whichever is less).
The first two sentences of paragraph 5.5.9.d.4 would state:
For Unit 1, 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 indications shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). For Unit 2, if crack indications are found in any SG tube from 14.01 inches below the top of the tubesheet on the hot leg side to 14.01 inches below the top of the tubesheet on the cold leg side, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indications shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections).
The proposed change is replacing the words for each SG" with the words for each affected and potentially affected SG."
Proposed changes in paragraph 5.5.9.d.4 permit SG inspection intervals to extend over multiple fuel cycles for SGs with alloy 600 TT as well as 690 TT tubing, assuming that such intervals can be implemented while ensuring tube integrity is maintained in accordance with paragraph 5.5.9.d.4. However, stress-corrosion cracks may not become detectable by inspection until the crack depth approaches the tube repair limit. In addition, stress-corrosion cracks may exhibit high growth rates. For these reasons, once cracks have been found in any SG tube, paragraph 5.5.9.d.4 restricts the allowable interval to the next scheduled inspection to 24 EFPM or one refueling outage (whichever results in more frequent inspections). The intent of this requirement is that it applies to the affected SG and to any other SG which may be potentially affected by the degradation mechanism that caused the known crack(s).
For example, a root cause analysis in response to the initial finding of one or more cracks may reveal that the crack(s) are associated with a manufacturing anomaly which causes locally high residual stress which in turn caused the early initiation of cracks at the affected locations. If it can be established that the extent of condition of the manufacturing anomaly applies only to one SG and not the others, then the NRC staff agrees that only the affected SG needs to be inspected within 24 EFPM or one refueling cycle in accordance with paragraph 5.5.9.d.4. The next scheduled inspections of the other SGs will continue to be subject to all other provisions of paragraph 5.5.9.d. The NRC staff concludes that the proposed change to paragraph 5.5.9.d.4 is acceptable, because it clarifies the intent of the paragraph.
Also, the change from "whichever is less" to "whichever results in more frequent inspections" improves the sentence structure, does not represent a change to any inspection requirements, and, therefore, is acceptable.
3.2.8 Specification 5.6.9, "Unit 1 Model D76 and Unit 2 Model D5 Steam Generator Tube Inspection Report" This specification lists the items to be included in a report which shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." By letter dated August 29, 2013, the licensee proposed the following changes toTS 5.6.9:
Current item b, which states "Active degradation mechanisms found... " would be revised to state "Degradation mechanisms found... "
Current Item e, which states "Number of tubes plugged during the inspection outage for each active degradation mechanism... " would be revised to state "Number of tubes plugged during the inspection outage for each degradation mechanism... "
Current item f, which states "Total number and percentage of tubes plugged to date... " would be revised to state "The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator... "
The licensee proposed to delete the word "active" in items b and e above. Thus, all degradation mechanisms found, whether deemed to be active or not, would now be reportable. The proposed change to item f is an editorial change that does not materially change the reporting requirements. The NRC staff concludes that the proposed changes are acceptable.
The NRC reviewed the proposed changes to the licensee-controlled TS Bases for consistency with the proposed TS changes. One reference to tube repair methods was not removed from TS bases page B 3.4-88. This reference to tube repair methods should be removed, since the use of tube repair methods (such as sleeving) is no longer authorized at CPNPP.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement 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, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on October 1, 2013 (78 FR 60324 ).
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: R. Grover Da~: February 27, 2014
A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
Docket Nos. 50-445 and 50-446
Enclosures:
- 1. Amendment No. 161 to NPF-87
- 2. Amendment No. 161 to NPF-89
- 3. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
PUBLIC LPL4-1 Reading RidsAcrsAcnw_MaiiCTR Resource RidsNrrDeEsgb Resource RidsNrrDoriDpr Resource RidsNrrDorllpl4-1 Resource ADAMS Accession No.: ML14042A223 Sincerely,
/RAJ Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDssStsb Resource RidsNrrLAJBurkhardt Resource RidsNrrPMComanchePeak Resource RidsRgn4MaiiCenter Resource RGrover, NRR/DSS/STSB KKarwoski, NRR/DE/ESGB
- See memo dated 2/5/2014 OFFICE NRR/DORLILPL4-1 /PM NRR/DORLILPL4-1/PM NRR/DORL/LPL4-1/LA NRRIDSS/STSB/BC*
NAME A George BSingal JBurkhardt REIIiott DATE 2/18/14 2/18/14 2/13/14 2/5/14 OFFICE OGC NLO NRR/DORL/LPL4-1/BC NRR/DORL/LPL4-1/PM NAME A Ghosh MMarkley (Flyon for)
BSingal DATE 2/26/14 2/27/14 2/27/14 OFFICIAL AGENCY RECORD