ML062570205
| ML062570205 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  (NPF-068, NPF-081) |
| Issue date: | 09/12/2006 |
| From: | NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | |
| Gratton C, NRR/DORL 415-1055 | |
| Shared Package | |
| ML062260302 | List: |
| References | |
| TAC MC2642, TAC MC2643 | |
| Download: ML062570205 (12) | |
Text
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 146 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.
(4)
DELETED (5)
DELETED (6)
DELETED (7)
DELETED (8)
DELETED (9)
DELETED (10) Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 102, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Additional Conditions.
D.
The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include (a) an exemption from the requirements of 10 CFR 70.24 for two criticality monitors around the fuel storage area, and (b) an exemption from the requirements of Paragraph IIl.D.2(b)(ii) of Appendix J of 10 CFR 50, the testing of containment air locks at times when containment integrity is not required.
The special circumstances regarding exemption b are identified in Section 6.2.6 of SSER 5.
An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1967, issued August 21, 1986, and relieved GPC from the requirement of having a criticality alarm system. GPC and Southern Nuclear are hereby exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.
These exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. The exemptions in items b and c above are granted pursuant to 10 CFR 50.1?. With Amendment No. 146 C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.
(1)
Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2)
Technical SDecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 126, and the Environmental Protection Plan contained in Appendix B, both ot which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
The Surveillance Requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 74. The SRs listed below shall be successfully demonstrated prior to the time and condition specified below for each:
a)
DELETED b)
DELETED c)
SR 3.8.1.20 shall be successfully demonstrated at the first regularly scheduled performance after implementation of this license amendment.
(3)
Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.
(4)
Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 80, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Additional Conditions.
D.
The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70. These include (a) an exemption from the requirements of 10 CFR 70.24 for two criticality monitors around the fuel storage area, and (b) an exemption from the requirements of Paragraph IIl.D.2(b)(ii) of Appendix J of 10 CFR 50, the testing of containment air locks at times when containment integrity is not required.
The special circumstances regarding exemption b are identified in Section 6.2.6 of SSER 8.
Amendment No.126
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Pro-gram (continued)
- 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 I gpm per SG.
- 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."
- c.
Provisions for SG tube repair 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.
The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:
- 1. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation found in the portion of the tube below 17 inches from the top of the hot leg tubesheet does not require plugging.
For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, degradation Identified in the portion of the tube from the top of the hot leg tubesheet to 17 Inches below the top of the tubesheet shall be plugged upon detection.
- 2. For Unit 1 during Refueling Outage 13 and the subsequent operating cycle, and for Unit 2 during Refueling Outage 12 and the subsequent operating cycle, degradation Identified in the portion of the tube below 17 Inches from the top of the hot leg tubesheet does not require plugging.
For Unit I during Refueling Outage 13 and the subsequent operating cycle, and for Unit 2 during Refueling Outage 12 and the subsequent operating cycle, degradation identified In the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the hot leg tubesheet shall be plugged upon detection.
- 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 (continued)
Vogtle Units 1 and 2 5.5-8 Amendment N0146 (Unit 1)
Amendment Nol 26 (Unit 2)
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator.(SG) Program (continued) present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Unit 2 during Refueling Outage 11 and the subsequent operating cycle, the portion of the tube below 17 Inches from the top of the hot leg tubesheet is excluded. For Unit 1 during Refueling Outage 13 and the subsequent operating cycle, and for Unit 2 during Refueling Outage 12 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded. 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. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.
- 2. Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.
- 3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever Is less). If definitive information, such as from examination of a pulled tube, diagnostic nondestructive 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.
(continued)
Vogtle Units 1 and 2 5.5-9 Amendment No146 (Unit 1)
Amendment No.126 (Unit 2)
Programs and Manuals 5.5 5.5 Programs andManuals (continued) 5.5.10 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.
Ventilation Filter Testing Program (VFTP)
A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1980:
- a.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% when tested In accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1980 at the system flow rate specified below
- 10%.
5.5.11 ESF Ventilation System Control Room Emergency Filtration System (CREFS)
Piping Penetration Area Filtration and Exhaust (PPAFES)
Flow Rate 19.000 CFM 15,500 CFM (continued)
Vogtle Units I and 2 5.5-10 Amendment Noj146 (Unit 1) I Amendment No.126 (Unit 2) I
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)
- b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1980 at the system flow rate specified below :t 10%.
ESF Ventilation System Flow Rate CREFS 19,000 CFM PPAFES 15,500 CFM
- c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl Iodide penetration less than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C and greater than or equal to the relative humidity specified below.
ESF Ventilation System Penetration RH CREFS
.2%
70%
PPAFES 10%
95%
- d.
Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the charcoal adsorbers, and CREFS cooling coils is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ASME N510-1989 at the system flow rate specified below +/- 10%.
ESF Ventilation System Delta P Flow Rate CREFS 7.1 in.
19,000 CFM water gauge PPAFES 6 in.
- e.
Demonstrate that the heaters for the CREFS dissipate 2 95 kW when corrected to 460 V when tested in accordance with ASME N510-1989.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
(continued)
Vogtle Units 1 and 2 5.5-11 Amendment NoJ46 (Unit 1)
Amendment No. 2 6 (Unit 2)
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.12 Explosive Gas and Storac-e Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Gaseous Waste Processing System, the quantity of radioactivity contained in each Gas Decay Tank, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure." The liquid radwaste quantities shall be limited to 10 curies per outdoor tank in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures."
The program shall include:
- a.
The limits for concentrations of hydrogen and oxygen in the Gaseous Waste Processing System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);
- b.
A surveillance program to ensure that the quantity of radioactivity contained in each gas decay tank is less than the amount that would result in a whole body exposure of 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 liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is limited to 1 10 curies per tank, excluding tritium and dissolved or entrained noble gases. This surveillance program provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations would be 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.
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.
5.5.13 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 (continued)
Vogtle Units 1 and 2 5.5-12 Amendment No. 146(Unit 1)
Amendment No. 126(Unit 2)
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program (continued)
ASTM Standards. The purpose of the program is to establish the following:
- a.
Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
- 1.
an API gravity or an absolute specific gravity within limits, or an API gravity or specific gravity within limits when compared to the supplier's certificate;
- 2.
a flash point within limits for ASTM 2D fuel oil, and, if gravity was not determined by comparison with supplier's certification, a kinematic viscosity within limits for ASTM 2D fuel oil; and
- 3.
a clear and bright appearance with proper color.
- b.
Other properties for ASTM 2D fuel oil are within limits within 30 days following sampling and addition to storage tanks; and
- c.
Total particulate concentration of the fuel oil is < 10 mg/i when tested every 31 days.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program surveillance frequencies.
5.5.14 Technical Specifications (TS) Bases Control Pro-gram This program provides a means for processing changes to the Bases of these Technical Specifications.
- a.
Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews
- b.
Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
- 1.
a change in the TS incorporated in the license; or
- 2.
a change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
- c.
The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
(continued)
Vogtle Units 1 and 2 5.5-13 Amendment No. 146 (Unit 1) I Amendment No. 126 (Unit 2) I
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Technical Specifications (TS) Bases Control Program (continued)
- d.
Proposed changes that meet the criteria of (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).
5.5.15 Safety Function Determination Pro-gram (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. The SFDP shall contain the following:
- a.
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;
- b.
Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
- c.
Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
- d.
Other appropriate limitations and remedial or compensatory actions.
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
- a.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or
- b.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or
- c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
(continued)
Vogtle Units I and 2 5.5-14 Amendment No146 (Unit 1)
Amendment No.126 (Unit 2)
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)
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.
5.5.16 MS and FW Piping Inspection Program This program shall provide for the inspection of the four Main Steam and Feedwater lines from the containment penetration flued head outboard welds, up to the first five-way restraint. The extent of the inservice examinations completed during each inspection interval (ASME Code Section Xl) shall provide 100%
volumetric examination of circumferential and longitudinal welds to the extent practical. This augmented inservice inspection is consistent with the requirements of NRC Branch Technical Position MEB 3-1, "Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," November 1975 and Section 6.6 of the FSAR.
5.5.17 Containment Leaka-ge Rate Testinq Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Testing Program," dated September 1995, as modified by the following exceptions:
- 1.
Leakage rate testing for containment purge valves with resilient seals is performed once per 18 months in accordance with LCO 3.6.3, SR 3.6.3.6 and SR 3.0.2.
- 2.
Containment personnel air lock door seals will be tested prior to reestablishing containment integrity when the air lock has been used for containment entry. When containment integrity is required and the air lock has been used for containment entry, door seals will be tested at least once per 30 days during the period that containment entry(ies) is (are) being made.
- 3.
The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified (continued)
Vogtle Units 1 and 2 5.5-15 Amendment No.146 (Unit 1)
Amendment No. 1 2 6 (Unit 2)
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.17 Containment Leakage Rate Testing Program (continued) by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC. At the discretion of the licensee, the containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenancelrefueling outage.
- 4.
A one time exception to NEI 94-01, Rev. 0, uIndustry Guidelines for Implementing Performance-Based Option of 10 CFR 50, Appendix J":
Section 9.2.3:
The next Type A test, after the March 2002 test for Unit I and the March 1995 test for Unit 2, shall be performed within 15 years.
The peak calculated primary containment internal pressure for the design basis loss of coolant accident, P,, is 37 psig.
The maximum allowable containment leakage rate, L., at P., is 0.2% of primary containment air weight per day.
Leakage rate acceptance criteria are:
- a.
Containment overall leakage rate acceptance criteria are _ 1.0 L,. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L, for the combined Type B and Type C tests, and *0.75 L. for Type A tests;
- b.
Air lock testing acceptance criteria are:
- 1)
Overall air lock leakage rate is _ 0.05 L, when tested at > P,,
- 2)
For each door, the leakage rate is _ 0.01 L, when pressurized to
Ž: P,.
The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
(continued)
Vogtle Units I and 2 5.5-16 Amendment No. 146 (Unit 1)
Amendment No. 126 (Unit 2)
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.18 Configuration Risk Management Program The Configuration Risk Management Program (CRMP) provides a proceduralized risk-informed assessment to manage the risk associated with equipment inoperability. The program applies to technical specification structures, systems, or components for which a risk-informed allowed outage time has been granted. The program shall include the following elements:
- a.
Provisions for the control and implementation of a Level I at power internal events PRA-informed methodology. The assessment shall be capable of evaluating the applicable plant configuration.
- b.
Provisions for performing an assessment prior to entering the LCO Condition for preplanned activities.
- c.
Provisions for performing an assessment after entering the LCO Condition for unplanned entry into the LCO Condition.
- d.
Provisions for assessing the need for additional actions after the discovery of additional equipment out of service conditions while in the LCO Condition.
- e.
Provisions for considering other applicable risk significant contributors such as Level 2 issues and external events, qualitatively or quantitatively.
5.5.19 Battery Monitoring and Maintenance Program This program provides for 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." of the following:
- a.
Actions to restore battery cells with float voltage < 2.13 V, and b
Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates.
Vogtle Units 1 and 2 5.5-17 Amendment No.146(Unit 1)
Amendment No.126(Unit 2)