ML13046A114
| ML13046A114 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/22/2013 |
| From: | Thomas Wengert Plant Licensing Branch III |
| To: | Weber L Indiana Michigan Power Co |
| Wengert T NRR/DORL/LPL3-1 301-415-4037 | |
| References | |
| TAC ME9560, TAC ME9561 | |
| Download: ML13046A114 (54) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 22, 2013 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106 SUB"IECT:
DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENTS RE: ADOPTION OF TSTF-51 0, REVISION 2, "REVISION TO STEAM GENERATOR PROGRAM INSPECTION FREQUENCIES AND TUBE SAMPLE SELECTION," USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NOS. ME9560 AND ME9561)
Dear Mr. Weber:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 320 to Renewed Facility Operating License No. DPR-58 and Amendment No. 304 to Renewed Facility Operating License No. DPR-74 for the Donald C. Cook Nuclear Plant, Units 1 and 2. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated September 12, 2012.
The amendments revise the TSs to adopt NRC-approved Technical Specifications Task Force (TSTF) Change Traveler TSTF-51 0, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," using the consolidated line item improvement process. Specifically, the amendments revise TS 3.4.17, "Steam Generator (SG)
Tube Integrity," TS 5.5.7, "Steam Generator (SG) Program," and TS 5.6.7, "Steam Generator Tube Inspection Report," and include TS Bases changes that summarize and clarify the purpose of the TS.
L. Weber
- 2 A copy of our related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
-t£'~~~
Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosures:
- 1. Amendment No. 320 to DPR-58
- 2. Amendment No. 304 to DPR-74
- 3. Safety Evaluation cc w/encls: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 320 License No. DPR-58
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated September 12, 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
- 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-58 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 320, are hereby incorporated into this license. The licensee 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 180 days.
FOR THE NUCLEAR REGULATORY COMMISSION Ya,J l1d~ ItA-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: March 22, 2013
ATTACHMENT TO LICENSE AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-315 Replace the following page of Renewed Facility Operating License DPR-58 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT 3
3 Replace the following pages of 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.
REMOVE INSERT 3.4.17-1 3.4.17-1 3.4.17-2 3.4.17-2 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 5.6-4 5.6-4
-3 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, 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; and (5) Pursuant to the Act and 10 CFR Parts 30 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 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, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal in accordance with the conditions specified herein.
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 320 are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Less than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined in Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operating license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than found loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.
(4) Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31, 1979, January 30, 1981, February 7,1983, November 22,1983, December 23, 1983, March 16, 1984, August 27, 1985, Renewed License No. DPR-58 Amendment No. 1 through 318. 320
SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
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
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.1 AND A.2 Verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection.
Plug the affected tube(s) in accordance with the Steam Generator Program.
7 days Prior to entering MODE 4 following the next refueling outage or SG tube inspection B. Required Action and associated Completion Time of Condition A not met.
8.1 AND Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> OR SG tube integrity not maintained.
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 /> Cook Nuclear Plant Unit 1 3.4.17-1 Amendment No. ~, 320
3.4.17 SG Tube Integrity SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam Generator Program.
In accordance with the Steam Generator Program SR 3.4.17.2 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 a SG tube inspection Cook Nuclear Plant Unit 1 3.4.17-2 Amendment No. ~, 320
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 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 for all SGs.
Cook Nuclear Plant Unit 1 5.5-5 Amendment No. ~, 298, 320
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)
- 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "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 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.
- 1. Inspect 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, 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 Cook Nuclear Plant Unit 1 5.5-6 Amendment No. 237, ~, 320
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 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; 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 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.
Cook Nuclear Plant Unit 1 5.5-7 Amendment No. ~, 298, 320
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. 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;
- 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.
5.5.9 Ventilation Filter Testing Program (VFTP)
The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal ad sorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.
Tests described in Specification 5.5.9.c shall be performed once per 24 months; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of ad sorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.
Tests described in Specification 5.5.9.d shall be performed once per 24 months.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.
Cook Nuclear Plant Unit 1 5.5-8 Amendment No. ~, 298
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program NFTP) (continued)
- a.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a removal efficiency of> 99%
of the dioctyl phthalate (DOP) when tested in accordance with the standard and at the system flowrate specified below:
ESF Ventilation System ANSI Standard Flowrate (cfm)
CREV System N510-1975
- 5,400 and S 6,600 ESF Ventilation System N510-1980
- 22,500 and S 27,500 FHAEV System N510-1980
- 27,000 and S 33,000
- b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a removal efficiency of ;;:: 99% of a halogenated hydrocarbon refrigerant test gas when tested in accordance with the standard and at the system flowrate specified below:
ESF Ventilation System ANSI Standard Flowrate (cfm)
CREV System N510-1975
- 5,400 and S 6,600 ESF Ventilation System N510-1980
- 22,500 and S 27,500 FHAEV System N510-1980
- 27,000 and S 33,000
- c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers, shows the methyl iodide penetration less than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity (RH) specified below:
Cook Nuclear Plant Unit 1 5.5-9 Amendment No. ~, 298
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program NFTP) (continued)
ESF Ventilation System Face Velocity (fpm) Penetration (%) RH (%)
CREV System NA 1
95 ESF Ventilation System 45.5 5
95 FHAEV System 46.8 5
95 In addition, the carbon samples not obtained from test canisters shall be prepared by either:
- 1.
Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or
- 2.
Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.
- d.
Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:
Delta P ESF Ventilation System (inches water gauge)
Flowrate (cfm)
CREV System 4
2: 5,400 and :S 6,600 ESF Ventilation System 4
2: 22,500 and :S 27,500 FHAEV System 4
2: 27,000 and :S 33,000 Cook Nuclear Plant Unit 1 5.5-10 Amendment No. ~, 298, 305
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor temporary liquid storage tanks.
The program shall include:
- a.
The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a Surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (Le., whether or not the system is designed to withstand a hydrogen explosion);
- b.
A Surveillance program to ensure that the quantity of radioactivity contained in each gas storage 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 temporary liquid storage 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, 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.
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.
5.5.11 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, an absolute specific gravity, or a specific gravity within limits; Cook Nuclear Plant Unit 1 5.5-11 Amendment No. ~, 298
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Program (continued)
- 2.
A flash point within limits and, if the gravity was not determined by comparison with the supplier's certification, a kinematic or saybolt viscosity within limits; and
- 3.
A clear and bright appearance with proper color;
- b.
Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.11.a above, are within limits; and
- c.
Total particulate concentration of the fuel oil is::; 10 mgtl when tested every 31 days in accordance with ASTM 0-2276, Method A.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.
5.5.12 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 UFSAR.
- d.
Proposed changes that meet the criteria of Specification 5.5.12.b 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).
Amendment No. ~, 298 Cook Nuclear Plant Unit 1
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP)
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.
- a.
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 unit 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.
- b.
A loss of safety function exists when, assuming no concurrent single failure.
no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), 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;
- 2.
A required system redundant to the system(s) in tum supported by the inoperable supported system is also inoperable; or
- 3.
A required system redundant to the support system(s) for the supported systems described in Specifications 5.5.13.b.1 and 5.5.13.b.2 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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system. the appropriate Conditions and Required Actions to enter are those of the support system.
Cook Nuclear Plant Unit 1 5.5-13 Amendment No. 2&7. 298
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program
- a.
A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) 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 Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in October 1992.
- 2.
A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing.
- b.
The calculated peak containment internal pressure for the design basis loss of coolant accident, P a, is 12 psig.
- c.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.25%
of containment air weight per day.
- d.
Leakage rate acceptance criteria are:
- 1.
Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the Type Band C tests and s 0.75 La for Type A tests.
- 2.
Air lock testing acceptance criterion is overall air lock leakage rate is S 0.05 La when tested at 2: Pa.
- e.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Cook Nuclear Plant Unit 1 5.5-14 Amendment No. ~f 298
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer including 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 minimum established design limit.
5.5.16 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 Ventilation (CREV) System, CRE occupants can control the reactor safely 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 total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
- a.
The definition of the CRE and the CRE boundary.
- 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 in leakage 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 O.
The following is an exception to Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0:
The appropriate application of ASTM E741-00 required by C.1.1 may include minor exceptions to the test methodology. These exceptions shall be documented in the test report.
Cook Nuclear Plant Unit 1 5.5-15 Amendment No. ~, ~, 307
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program (continued)
- 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 CREV System, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic 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 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 0, respectively.
Cook Nuclear Plant Unit 1 5.5-16 Amendment No. 307
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 8. WCAP-12610-P-A &CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO', July 2006 (Westinghouse Proprietary).
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Post Accident Monitoring Report When a report is required by Condition B or G of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 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 5.5.7, 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.
Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 287, ~, ~, ~,
a+e,320
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE
\\
Amendment No. 304 License No. DPR-74
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company (the licensee) dated September 12, 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
- 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-74 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 304, are hereby incorporated into this license. The licensee 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 180 days.
FOR THE NUCLEAR REGULATORY COMMISSION Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: ~1arch 22, 2013
ATTACHMENT TO LICENSE AMENDMENT NO. 304 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Replace the following page of Renewed Facility Operating License DPR-74 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT 3
3 Replace the following page of Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
REMOVE INSERT 3.4.17-1 3.4.17-1 3.4.17-2 3.4.17-2 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 5.6-4 5.6-4
- 3 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, 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; and (5) Pursuant to the Act and 10 CFR Parts 30 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 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, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3468 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to the renewed operating license.
The preoperational tests, startup tests and other items identified in Attachment 1 to this renewed operating license shall be completed. Attachment 1 is an integral part of this renewed operating license.
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 304, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Additional Conditions (a) Deleted by Amendment No. 76 (b) Deleted by Amendment No.2 (c) Leak Testing of Emergency Core Cooling System Valves Indiana Michigan Power Company shall prior to completion of the first inservice testing interval leak test each of the two valves in series in the Renewed License No. DPR-74 Amendment No. 1 through 302, 304
3.4.17 SG Tube Integrity 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.
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
NOTE----------------------------------------------------
Separate Condition entry is allowed for each SG tube.
CONDITION REQUIRED ACTION A. One or more SG tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program.
A.1 AND A.2 Verify tube integrity of the affected tube{s) is maintained until the next refueling outage or SG tube inspection.
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 B. Required Action and associated Completion Time of Condition A not B.1 AND Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> met.
OR 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 /> SG tube integrity not maintained.
Cook Nuclear Plant Unit 2 3.4.17-1 Amendment No..&79, 304
3.4.17 SG Tube Integrity SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam Generator Program.
In accordance with the Steam Generator Program SR 3.4.17.2 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 a SG tube inspection Cook Nuclear Plant Unit 2 3.4.17-2 Amendment No. ~! 304
5.5 Programs and Manuals 5.5 Programs and Manuals 5.5.7 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 for all SGs.
Cook Nuclear Plant Unit 2 5.5-5 Amendment No. ~, 279, 304
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)
- 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "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 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.
- 1. Inspect 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, 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 Cook Nuclear Plant Unit 2 5.5-6 Amendment No. ~, ;g.g, 304
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) 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; 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 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.
Cook Nuclear Plant Unit 2 5.5-7 Amendment No. ~, ~, 304
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. 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;
- 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.
5.5.9 Ventilation Filter Testing Program (VFTP)
The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or.charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.
Tests described in Specification 5.5.9.c shall be performed once per 24 months; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of adsorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.
Tests described in Specification 5.5.9.d shall be performed once per 24 months.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.
Cook Nuclear Plant Unit 2 5.5-8 Amendment No. ~, 279
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program NFTP) (continued)
- a.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a removal efficiency of > 99%
of the dioctyl phthalate (DOP) when tested in accordance with the standard and at the system flowrate specified below:
ESF Ventilation System ANSI Standard Flowrate (cfm)
CREV System N510-1975 2: 5,400 and S 6,600 ESF Ventilation System N510-1980 2: 22,500 and S 27,500 FHAEV System N510-1980 2: 27,000 and S 33,000
- b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal ad sorber shows a removal efficiency of 2: 99% of a halogenated hydrocarbon refrigerant test gas when tested in accordance with the standard and at the system flowrate specified below:
ESF Ventilation System ANSI Standard Flowrate (cfm)
CREV System N510-1975 2: 5,400 and S 6,600 ESF Ventilation System N510-1980 2: 22,500 and S 27,500 FHAEV System N510-1980 2: 27,000 and S 33,000
- c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers, shows the methyl iodide penetration less than or equal to the value specified below when tested in accordance with ASTM 03803-1989 at a temperature of 30°C (86°F) and the relative humidity (RH) specified below:
Cook Nuclear Plant Unit 2 5.5-9 Amendment No. ~, 279
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)
ESF Ventilation System Face Velocity (fpm) Penetration (%) RH (%)
CREV System NA 1
95 ESF Ventilation System 45.5 5
95 FHAEV System 46.8 5
95 In addition, the carbon samples not obtained from test canisters shall be prepared by either:
- 1.
Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or
- 2.
Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.
- d.
Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:
Delta P ESF Ventilation System (inches water gauge)
Flowrate (cfm)
CREV System 4
- 5,400 and s 6,600 ESF Ventilation System 4
- 22,500 and s 27,500 FHAEV System 4
- 27,000 and S 33,000 Cook Nuclear Plant Unit 2 5.5-10 Amendment No. ~, ~, 288
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor temporary liquid storage tanks.
The program shall include:
- a.
The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup 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);
- b.
A Surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 2: 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 temporary liquid storage 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, 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.
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.
5.5.11 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. an absolute specific gravity, or a specific gravity within limits; Cook Nuclear Plant Unit 2 5.5-11 Amendment No. 2&9. 279
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Program (continued)
- 2.
A flash point within limits and, if the gravity was not determined by comparison with the supplier's certification, a kinematic or saybolt viscosity within limits; and
- 3.
A clear and bright appearance with proper color;
- b.
Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.11.a above, are within limits; and
- c.
Total particulate concentration of the fuel oil is S 10 mgll when tested every 31 days in accordance with ASTM 0-2276, Method A.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test Frequencies.
5.5.12 Technical Specifications (IS) 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 UFSAR.
- d.
Proposed changes that meet the criteria of Specification 5.5.12.b 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).
Cook Nuclear Plant Unit 2 5.5-12 Amendment No. ~, 279
Programs and Manuals 5.5 5.5 Programs and Manuals I
5.5.13 Safety Function Determination Program (SFDP)
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.
- a.
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 unit 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.
- b.
A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), 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;
- 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 described in Specifications 5.5.13.b.1 and 5.5.13.b.2 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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.
Cook Nuclear Plant Unit 2 5.5-13 Amendment No. 269, 279
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program
- a.
A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) 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 Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in May 1992.
- b.
The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12 psig.
- c.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.25%
of containment air weight per day.
- d.
Leakage rate acceptance criteria are:
- 1.
Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are :s; 0.60 La for the Type Band C tests and :s; 0.75 La for Type A tests.
- 2.
Air lock testing acceptance criterion is overall air lock leakage rate is
- s; 0.05 La when tested at ~ Pa.
- e.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer including 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 minimum established design limit.
Cook Nuclear Plant Unit 2 5.5-14 Amendment No. ~, 279
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.16 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 Ventilation (CREV) System, CRE occupants can control the reactor safely 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 total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
- 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 O.
The following is an exception to Section C.1 and C.2 of Regulatory Guide 1.197, Revision 0:
The appropriate application of ASTM E741-00 required by C.1.1 may include minor exceptions to the test methodology. These exceptions shall be documented in the test report.
- 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 CREV System, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the periodic 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.
Cook Nuclear Plant Unit 2 5.5-15 Amendment No. 289
Programs and Manuals 5.5 I
5.5 Programs and Manuals 5.5.16 Control Room Envelope Habitability Program (continued)
- 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 0, respectively.
Cook Nuclear Plant Unit 2 5.5-16 Amendment No. 289
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COlR) (continued)
- 7. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOl Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary); and
- 8. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRlO', July 2006 (Westinghouse Proprietary)
- 9. 5.6 Reporting Requirements
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COlR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H 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 preplan ned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.7 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 5.5.7, 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,
- f.
The number and percentages of tubes plugged to date, and the effective plugging percentage in each generator, and Cook Nuclear Plant Unit 2 5.6-4 Amendment No. ~, ~, ~, 2-9&, 2Q7.
~,304
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 AND AMENDMENT NO. 304 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-315 AND 50-316
1.0 INTRODUCTION
By letter dated September 12, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12268A047), Indiana Michigan Power Company (the licensee),
proposed changes to the technical speCifications (TSs) for Donald C. Cook Nuclear Plant (CNP), Units 1 and 2, to adopt U.S. Nuclear Regulatory Commission (NRC, the Commission) approved Revision 2 to TS Task Force (TSTF) Standard Technical Specifications (STS) Change Traveler TSTF-510, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
The proposed amendments would revise TS 3.4.17, "Steam Generator (SG) Tube Integrity,"
TS 5.5.7, "Steam Generator (SG) Program," and TS 5.6.7, "Steam Generator Tube Inspection Report," and include TS Bases changes that summarize and clarify the purpose of the TS. The specific changes concern SG inspection periods and address applicable administrative changes and clarifications.
The licensee stated that the license amendment request is consistent with the Notice of Availability of TSTF-51 0, Revision 2, announced in the Federal Register (FR) on October 27, 2011 (76 FR 66763), as part of the consolidated line item improvement process.
The current STS requirements in the above specifications were established in May 2005 with the NRC staff's approval of TSTF-449, Revision 4, "Steam Generator Tube Integrity" (NRC Notice of Availability (70 FR 24126; May 6,2005)). 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
- 2 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).
2.0 REGULATORY EVALUATION
The SG tubes in PWRs have a number of safety functions. These tubes are an integral part of the reactor coolant pressure boundary (RCPS) and, as such, are relied upon to maintain primary system pressure and inventory. As part of the RCPS, 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 (10 CFR) establishes the requirements with respect to the integrity of the SG tubing. Specifically, the General Design Criteria (GOC) in Appendix A to 10 CFR Part 50 states that the RCPS:
- shall..... have an extremely low probability of abnormal leakage... and of gross rupture" (GOC 14), "... shall be designed with sU'fficient margin... " (GOC 15 and 31),
- shall be of..... the highest quality standards practical." (GOC 30), and shall be designed to permit..... periodic inspection and testing... to assess... structural and leaktight integrity... " (GOC 32).
The licensee's application states that its proposed amendment complies with the 10 CFR 50 and applicable GOCs' requirements specified above, and provides the following explanation:
The construction permits for CNP were issued and the majority of construction was completed prior to issuance of 10 CFR 50, Appendix A, General Design Criteria, in 1971 by the Atomic Energy Commission (AEC); therefore, CNP was not licensed to the 10 CFR 50 Appendix A GOC. CNP was designed and constructed to comply with the AEC GOC as proposed on July 10, 1967. The application of the AEC proposed GOC to the CNP is contained in the CNP UFSAR [Updated Final Safety Analysis Report], Section 1.4, as the Plant Specific Design Criteria (PSOC). Appendix A of 10 CFR Part 50 GOC differ both in numbering and content from the PSOC for CNP. However, the following information demonstrates that CNP's PSOC meets the intent of GOC 14, 30, and 32 of 10 CFR 50, Appendix A.
CNP's PSOC 9 is the equivalent of GOC 14 as can be seen in the PSOC language below. With the exception of the testing requirement which is also prescribed within GOC 32, which is discussed further below in CNP's PSOC
- 3 equivalent to GOC 32. These differences do not alter the conclusion that the proposed change is applicable to CNP.
PSOC 9 Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated and constructed so as to have an exceedingly low probability of gross rupture or significant uncontrolled leak~ge throughout its design lifetime.
There are no changes to the SG design that impact CNP's PSOC 9 which is the equivalent of GOC 14. The evaluation performed in Section 4.0 of TSTF-51 0 concludes that the proposed change will continue to comply with CNP's PSOC.
CNP's PSOC 34 and PSOC 16 are the equivalent of GOC 30 as can be seen in the PSOC language below. These differences do not alter the conclusion that the proposed change is applicable to CNP.
PSOC 34 Reactor Coolant Pressure Boundary Propagation Failure Prevention The reactor coolant pressure boundary shall be designed and operated to reduce to an acceptable level the probability of rapidly propagating type failure. Consideration is given (a) to the provisions for control over service temperature and irradiation effects which may require operational restrictions, (b) to the design and construction of the reactor pressure vessel in accordance with applicable codes, including those which establish requirements for absorption of energy within the elastic strain energy range and for absorption of energy by plastic deformation and (c) to the design and construction of reactor coolant pressure boundary piping and equipment in accordance with applicable codes.
PSOC 16 Monitoring Reactor Coolant Leakage Means shall be provided to detect significant uncontrolled leakage from the reactor coolant pressure boundary.
There are no changes to the SG design that impact CNP's PSOC 34 and 16 which is the equivalent of GOC 30. The evaluation performed in Section 4.0 of TSTF-510 concludes that the proposed change will continue to comply with CNP's PSOC.
CNP's PSOC 36 is the equivalent of GOC 32 as can be seen in the PSOC language below. This difference does not alter the conclusion that the proposed change is applicable to CNP.
-4 PSDC 36 Reactor Coolant Pressure Boundary Surveillance Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance of critical areas by appropriate means to assess the structural and leaktight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance program conforming with current applicable codes shall be provided.
There are no changes to the SG design that impact CNP's PSDC 36 which is the equivalent of GDC 32.
The licensee further states, "The evaluation performed in Section 4.0 of TSTF-S10 concludes that the proposed change will continue to comply with CNP's PSDC."
The regulations in 10 CFR SO.SSa specify that components, which 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 SO.SSa further requires, in part, that throughout the service life of a PWR facility, ASME Code Class 1 components meet the requirements, except design and access provisions and pre-service examination requirements, in Section XI, "Rules for Inservice Inspection [(lSI)] of Nuclear Power Plant Components," of the ASME Code, to the extent practical. This requirement includes the inspection and repair criteria of Section XI of the ASME Code.
The regulations in 10 CFR SO.36, 'Technical specifications," establish the requirements related to the content of the TS. Pursuant to 10 CFR SO.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 (S) administrative controls. LCOs and accompanying action statements and SRs in the STS relevant to SG tube integrity are in Specification 3.4.13 "RCS [reactor coolant system] Operational Leakage," and Specification 3.4.17 (SR 3.4.17.2), "Steam Generator (SG) Tube Integrity." The SRs in the "Steam Generator (SG) Tube Integrity" specification reference the SG Program, which is defined in the STS administrative controls. CNP's TSs 3.4.13, "RCS Operational Leakage," and 3.4.17, "Steam Generator (SG) Tube Integrity," address requirements similar to those specified in STS sections above.
The regulations in 10 CFR SO.36(c)(S) 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. The SG Program is defined in CNP's TS S.S.7, while the reporting requirements relating to implementation of the SG Program are in TS S.6.7.
Specification S.S.7 requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. SG tube integrity is maintained by meeting the performance criteria specified in TS S.S.7.b for structural and leakage integrity, consistent with the plant deSign and licensing basis. Specification S.S.7.a requires that a condition monitoring
- 5 assessment be performed during each outage in which the SG tubes are inspected, to confirm that the performance criteria are being met. Specification 5.5.7.d 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 repair criteria. The applicable tube repair criteria, specified in TS 5.5.7.c, are that tubes found during inservice inspection to contain flaws with a depth equal to or exceeding 40 percent of the nominal wall thickness shall be plugged.
3.0 TECHNICAL EVALUATION
Each proposed change to the TS is described individually below, followed by the NRC staff's assessment of the change.
3.1 SpeCification 5.5.7. "Steam Generator (SG) Program" The last sentence in the introductory paragraph currently states, "In addition, the Steam Generator Program shall include the following provisions:"
Proposed Change:
The sentence is revised to state, "In addition, the Steam Generator Program shall include the following:" The subsequent paragraph starts with "Provisions for... "
and stating "provisions" in the introductory paragraph is duplicative.
Assessment: The NRC staff has reviewed the licensee's proposed change to Specification 5.5.7 and agrees that the word, "provisions," in the introductory paragraph is duplicative. The NRC staff agrees that the change is administrative in nature, and therefore is acceptable.
3.2 Specification 5.5.7, Paragraph 5.5.7 b.1. "Structural integrity performance criterion" The first sentence currently states:
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.
Proposed Change:
Revise the sentence as follows:
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 the sentence inappropriately includes antiCipated transients in the description of normal operating conditions.
- 6 Assessment: The NRC staff agrees the current wording is incorrect and that anticipated transients should be differentiated from normal operating conditions. Therefore, the NRC staff finds the change acceptable.
3.3.
Paragraph S.S.7.c, "Provisions for SG tube repair criteria," Paragraph S.S.7.d, "Provisions for SG tube inspections," TS 3.4.17, "Steam Generator (SG) Tube Integrity" Proposed Change:
Change all references to "tube repair criteria" to "tube plugging criteria."
This change is intended to be consistent with the treatment of SG tube repair throughout Specification 5.5.7.
Assessment: The NRC staff finds that the proposed change provides a more accurate label of the criteria and, therefore, adds clarity to the specification. This is because only one action must be taken when the criteria are exceeded. There are no alternative SG tube repair actions in CNP's TSs. Therefore, the NRC staff finds the change acceptable.
3.4 Paragraph S.S.7.d, "Provisions for SG tube inspections" Proposed Change:
Change the term "an assessment of degradation" to "a degradation assessment" to be consistent with the terminology used in industry program documents.
Assessment: The NRC staff agrees that the terminology should be consistent and finds the change acceptable.
3.5 Paragraph S.S.7.d.1 The paragraph currently states: "Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement."
Proposed change The change would replace "SG replacement" with "SG installation." The basis for the change is that it will allow the SG Program to apply to both existing plants and new plants.
Assessment: The NRC staff agrees the SG Program can apply to both existing and new plants. Therefore, the NRC staff finds the change acceptable.
3.6 Paragraph S.S.7.d.2 for plants with SGs with alloy 690 thermally-treated (TT) tubes NOTE: The subject license admendment request (LAR) specifically states, 'The SGs in both [CNP] units are equipped with alloy 690 thermally treated tubing; therefore, the TSTF-S10 material specific requirements for that material type are reflected in the proposed change."
The paragraph currently states:
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
- 7 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.
Proposed Change:
Revise Paragraph 5.5.7.d.2 to state the following:
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. This constitutes the first inspection period;
- 8)
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.
Assessment:
Paragraph 5.5.7.d.2 in its current form, and with the proposed changes, is similar for each of the tube alloy types used in domestic steam generators, but with differences that reflect the improved resistance of alloy 690 TT to stress corrosion cracking relative to both alloy 600 mill annealed (MA) and alloy 600 TT. These differences include progressively larger maximum
- 8 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, Paragraph 5.5.7.d.2 includes 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 Paragraph 5.5.7.d.2 to the end of the paragraph and make editorial changes to improve clarity. The NRC staff finds these changes to be of a clarifying nature, not changing the current intent of these two sentences. However, the LAR also includes three changes to when inspections are performed as follows:
- 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.
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 NRC staff finds that, for plants operating with a 54-month inspection intervals (three fuel cycles), the end of each inspection period will not generally coincide with a scheduled inspection outage.
The proposed changes would generally increase the number of inspections in each of the third and subsequent inspection periods by up to one additional inspection. This could reduce the required average minimum sample size during these periods. For plants operating with 54 month inspection intervals, the proposed changes will usually have no effect on the required average minimum sample size during these periods. However, inspection sample sizes will continue to be subject to Paragraph S.S.7.d.2, which states that in addition to meeting the requirements of Paragraph 5.5.7.d.2, 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 5.5.7.d.2 will continue to ensure both adequate inspection scopes and tube integrity.
For each inspection period, Paragraph 5.5.7.d.2 currently requires 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
- 9 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 second (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, six months before the mid-point (assuming an inspection was performed at the very end of the 144 EPFM 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., 10 percent during one and 40 percent in the other) totaling 50 percent.
The NRC staff finds there is no 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 100 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 Paragraph 5.5.7.d.2 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 5.5.7.d.2 requires a method of inspection to be performed with the objective of detecting circumferential cracks that may be present at this location and that may satisfy the applicable tube plugging criteria.
Furthermore, suppose this inspection is performed for the first time during the third of four SG inspections scheduled for the 144 EFPM inspection period. Paragraph 5.5.7.d.2 currently does not specify whether this location needs to be 100 percent inspected by the end of the 144 EFPM 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
- 10 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)?
[NRC Staff Position:] The TS contain requirements that are a mixture of prescriptive and performance-based elements. Paragraph lid" 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 lid" 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 1 00 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 1 00 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 finds that relocation of information in sentences 3 and 4 as described above clarifies the existing requirement consistent with the NRC staff's position from RIS 2009-04 quoted above and are, therefore, acceptable.
The proposed fifth sentence in Paragraph 5.5.7.d.2 states, "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." Allowing extension of the inspection periods by up to an additional three EFPM 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 144 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 three EFPM, then the minimum sample size for each of the five inspections could average as little as 20 percent of the
- 11 tube population. Therefore, 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 5.5.7.d.2, which states, in part, that in addition to meeting the requirements of paragraph 5.5.7.d.2, "... 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 that the proposed fifth sentence involves only a relatively minor relaxation to the eXisting sampling requirements in Paragraph 5.5.7.d.2. However, the performance-based requirements in 5.5.7.d.2 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 paragraph 5.5.7.d.2 replaces the last sentence of the current paragraph 5.5.7.d.2. This sentence establishes the minimum allowable SG inspection frequency as at least every 72 EFPM or at least every third refueling outage (whichever results in more frequent inspections). This minimum inspection frequency is unchanged from the current requirement in CNP's TSs. The NRC staff finds 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 paragraph 5.5.7.d.2 is acceptable.
3.7 Paragraph 5.5.7.d.3 for plants with SGs with alloy 690 TT tubes The first sentence of paragraph 5.5.7.d.3 currently states:
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).
Proposed Change:
Revise this sentence as follows:
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).
The proposed change replaces the words "for each SG" with the words "for each affected and potentially affected SG." The licensee states that the existing wording can be misinterpreted.
The licensee further states that the intent is that those SGs that are affected and those SGs that are potentially affected must be inspected for the degradation mechanism that caused the crack indication. However, some licensees have questioned whether the current reference to "each SG" requires only the SGs that are affected to be inspected for the degradation mechanism.
The proposed revision is intended to clarify the intent of the requirement.
Assessment: Proposed changes in Paragraph 5.5.7.d.2 (Section 3.6 of this Safety Evaluation) permits SG inspection intervals to extend over multiple fuel cycles for SGs with alloy 690 TT
- 12 tubing, assuming that such intervals can be implemented while ensuring tube integrity is maintained in accordance with paragraph 5.5.7.d. However, stress corrosion cracks may not become detectable by inspection until the crack depth approaches the tube plugging limit. In addition, stress corrosion cracks may exhibit high growth rates. For these reasons, once cracks have been found in any SG tube, Paragraph 5.5.7.d.3 restricts the allowable interval to the next scheduled inspection to 24 EFPM or one refueling outage (whichever is less). The intent of this requirement is that it applies to the affected SG and to any other SG that 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 might reveal that the crack(s) are associated with a manufacturing anomaly that caused locally high residual stress, which in turn causes 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 5.5.7.d.3. The next scheduled inspections of the other SGs will continue to be subject to all other provisions of paragraph 5.5.7.d. The NRC staff finds the proposed change to Paragraph 5.5.7.d.3 acceptable, because it clarifies the intent the paragraph.
3.8 Specification 5.6.7, "Steam Generator Tube Inspection Report" This specification lists items a. through g. 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 5.5.7, "Steam Generator (SG) Program."
Proposed Changes:
Item b. currently reads: "Active degradation mechanisms found... " to be revised to read: "Degradation mechanisms found... "
Item e. currently reads: "Number of tubes plugged during the inspection outage for each active degradation mechanism" to be revised to read: "Number of tubes plugged during the inspection outage for each degradation mechanism" Item f. currently reads, "Total number and percentage of tubes plugged to date" to be revised to read: "The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator."
Assessment: This proposal would 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 NRC staff finds the proposed change acceptable.
The change in item f. "The number and percentage of tubes plugged to date," does not materially change the reporting requirements. The additional change in item f. "the effective plugging percentage in each steam generator," is currently not specified in the CNP's TSs.
However, it is consistent with the approved TSTF-510 change. The NRC staff finds changes in item f. acceptable.
- 13 3.9 Technical Specification Pages 5.5-8 through 5.5-16 TS Pages 5.5-8 through 5.5-16 for CNP Units 1 and 2 are modified solely to accommodate re pagination of the TS pages. These changes are purely administrative in nature and are, therefore, acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Michigan 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 (77 FR 76080, dated December 26. 2012). 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 Date of issuance: March 22, 2013
ML13046A114 OFFICE NRR/LPL3-1/PM N RR/LPL3-2/LA NRRISTSB/BC OGC-NLO NRR/LPL3-1/BC NRR/LPL3-1/PM i NAME TWengert SRohrer RElliott LSubin RCarlson TWengert I~
T03i,v, 'v (KFeintuch for) 03/11/13 03/11/13 103/11/13 03/22/13 03/22/13