ML070600660
| ML070600660 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 03/01/2007 |
| From: | Stang J NRC/NRR/ADRO/DORL/LPLII-1 |
| To: | |
| Shared Package | |
| ML070320068 | List: |
| References | |
| TAC MD2036, TAC MD2037 | |
| Download: ML070600660 (18) | |
Text
(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 calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units I and 2, and; (6)
Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center.
C.
This renewed operating 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 The licensee is authorized to operate the facility at a reactor core full steady state power level of 3411 megawatts thermal (100%).
(2)
Technical Specifications The Technical Specifications contained In Appendix A, as revised through Amendment No. 237 are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.
Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC In writing when implementation of these activities is complete and can be verified by NRC Inspection.
The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be Included In the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4), following Issuance of this renewed operating license.
Until that update is complete, Duke may make changes to the programs described In such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
Renewed License No. NPF-9 Amendment No. 237
\\ \\ (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 calibration or associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2; and, (6)
Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training andTechnology Center.
C.
This renewed operating 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 The licensee is authorized to operate the facility at a reactor core full steady state power level of 341.1 megawatts thermal (100%).
(2)
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 218 are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
(3)
,Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation.
Duke shall complete these activities no later than March 3, 2023, and shall notify the NRC in writing when Implementation of these activities is complete and can be verified by NRC inspection.
The Updated Final Safety Analysis Report supjlement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license.
Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth In 10 CFR 50.59, and otherwise complies with the requirements in that section.
Amendment No. 21 8
TABLE OF CONTENTS (continued) 3.4 REACTOR COOLANT SYSTEM (RCS) (continued) 3.4.6 RCS Loops-MODE 4...................................................................
3.4.6-1 3.4.7 RCS Loops-MODE 5, Loops Filled...............................................
3.4.7-1 3.4.8 RCS Loops-MODE 5, Loops Not Filled........................................
3.4.8-1 3.4.9 P ressurizer.....................................................................................
3.4.9-1 3.4.10 Pressurizer Safety Valves..............................................................
3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)..................... 3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP) System.......... 3.4.12-1 3.4.13 RCS Operational LEAKAGE...........................
3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage.................................
3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation.......................................
3.4.15-1 3.4.16 RCS Specific Activity......................................................................
3.4.16-1 3.4.17 RCS Loop-Test Exceptions............................................................
3.4.17-1 3.4.18 Steam Generator (SG) Tube Integrity...............................................
3.4.18-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)............................
3.5.1-1 3.5.1 A ccum ulators.................................................................................
3.5.1-1 3.5.2 ECCS-Operating..........................................................................
3.5.2-1 3.5.3 ECCS-Shutdown..........................................................................
3.5.3-1 3.5.4 Refueling Water Storage Tank (RWST).........................................
3.5.4-1 3.5.5 Seal Injection Flow.........................................................................
3.5.5-1 3.6 CONTAINMENT SYSTEMS...................................................................
3.6.1-1 3.6.1 Containment...................................................
3.6.1-1 3.6.2 Containment Air Locks...................................................................
3.6.2-1 3.6.3 Containment Isolation Valves.........................................................
3.6.3-1 3.6.4 Containment Pressure...................................................................
3.6.4-1 3.6.5 Containment Air Temperature........................................................
3.6.5-1 3.6.6 Containment Spray System............................................................
3.6.6-1 3.6.7 N ot U sed........................................................................................
3.6.8 Hydrogen Skimmer System (HSS).................................................
3.6.8-i 3.6.9 Hydrogen Mitigation System (HMS)...............................................
3.6.9-1 3.6.10 Annulus Ventilation System (AVS).................................................
3.6.10-1 3.6.11 Air Return System (ARS)...............................................................
3.6.11-1 3.6.12 Ice B ed...........................................................................................
3.6.12-1 3.6.13 Ice Condenser Doors.....................................................................
3.6.13-1 3.6.14 Divider Barrier Integrity...................................................................
3.6.14-1 3.6.15 Containment Recirculation Drains..................................................
3.6.15-1 3.6.16 Reactor Building.............................................................................
3.6.16-1 3.7 PLANT SYSTEMS.................................................................................
3.7.1-1 3.7.1 Main Steam Safety Valves (MSSVs)..............................................
3.7.1-1 3.7.2 Main Steam Isolation Valves (MSIVs)............................................
3.7.2-1 3.7.3 Main Feedwater Isolation Valves (MFIVs),
Main Feedwater Control Valves (MFCVs), MFCV's Bypass Valves and Main Feedwater (MFW) to Auxiliary Feedwater (AFW) Nozzle Bypass Valves (MFW/AFW NBVs)........................................................
3.7.3-1 3.7.4 Steam Generator Power Operated Relief Valves (S G P O R V s).............................................................................
3.7.4-1 McGuire Units 1 and 2 ii Amendment Nos. 237,218
Definitions 1.1 1.1 Definitions (continued)
ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE LEAKAGE shall be:
- a.
Identified LEAKAGE
- 1.
LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal
,water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or 3
Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
- b.
Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;
- c.
Pressure Boundarv LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
(continued)
McGuire Units 1 and 2 1.1-3 Amendment Nos. 237,218
RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:
- a.
- b.
1 gpm unidentified LEAKAGE;
- c.
10 gpm identified LEAKAGE;
- d.
389 gallons per day total primary to secondary LEAKAGE through all steam generators (SGs); and
- e.
135 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
I APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A,
RCS Operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.
limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
B.
Required Action and B.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.
B.2 Be in MODE 5.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.
OR Primary to secondary LEAKAGE not within lim it.
I McGuire Units 1 and 2 3.4.13-1 Amendment Nos. 237,218
RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4-SR 3.4.13.1
NOTES
- 1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS Operational LEAKAGE is within limits by performance of RCS water inventory balance.
NOTE -------
Only required to be performed during steady state operation 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> SR 3.4.13.2
NOTE Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
Verify primary to secondary LEAKAGE is < 135 gallons 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per day through any one SG and <389 gallons per day total through all SGs.
McGuire Units 1 and 2 3.4.13-2 Amendment Nos. 237,218
SG Tube Integrity 3.4.18 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity LCO 3.4.18 SG tube integrity shall be maintained.
AND All SG tubes satisfying the tube repair 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 A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.
Program.
AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program.
next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.
B.2 Be in MODE 5.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.
McGuire Units 1 and 2 3.4.18-1 Amendment Nos. 237,218
SG Tube Integrity 3.4.18
¶ SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program.
with the Steam Generator Program SR 3.4.18.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program.
a SG tube inspection McGuire Units 1 and 2 3.4.18-2 Amendment Nos. 237, 218
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:
- a.
Testing frequencies specified in Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:
ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing activities Required Frequencies for performing inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days At least once per 731 days
- b.
The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;
- c.
The provisions of SR 3.0.3 are applicable to inservice testing activities; and
- d.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.
5.5.9 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 provisions:
- 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 (continued)
McGuire Units 1 and 2 5.5-6 Amendment Nos. 237, 218
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG).Program (continued) 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 and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and 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 0.27 gallons per minute total.
- 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.
(continued)
McGuire Units 1 and 2 5.5-7 M U 15Amendment Nos. 237, 218,
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program (continued)
- 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 repair 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. 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.
- 1.
Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
- 2.
Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.
- 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 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.
(continued)
McGuire Units I and 2 5.5-8 Amendment Nos. 237,218
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation and low pressure turbine disc stress corrosion cracking. The program shall include:
- a.
Identification of a sampling schedule for the critical variables and control points for these variables;
- b.
Identification of the procedures used to measure the values of the critical variables;
- c.
Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage; d-Procedures for the recording and management of data;
- e.
Procedures defining corrective actions for all off control point chemistry conditions; and
- f.
A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.
5.5.11 Ventilation Filter Testinqg Program (VFTP)
AA program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975, with exceptions as noted in the UFSAR.
- a.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows the following penetration and system bypass when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 (N510-1980 for Auxiliary Building Filtered Exhaust) at the flowrate specified below +/- 10%.
(continued)
McGuire Units 1 and 2 5.5-9 Amendment Nos. 237,218
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP) (continued)
ESF Ventilation System Penetration Flowrate Annulus Ventilation Control Area Ventilation Aux. Bldg. Filtered Exhaust (2 fans)(Unit 1)
Aux. Bldg. Filtered Exhaust (2 fans)(Unit 2)
Containment Purge (non-ESF) (2 fans)
Fuel Bldg. Ventilation (non-ESF)
<1%
< 0.05%
<1%
<1%
<1%
<1%
8000 cfm 2000 cfm 45,700 cfm 40,500 cfm 21,000 cfm 35,000 cfm
- b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows the following penetration and system bypass when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 (N510-1980 for Auxiliary Building Filtered Exhaust) at the flowrate specified below +/- 10%.
ESF Ventilation System Penetration Flowrate Annulus Ventilation Control Area Ventilation Aux. Bldg. Filtered Exhaust (2 fans)(Unit 1)
Aux. Bldg. Filtered Exhaust (2 fans)(Unit 2)
Containment Purge (non-ESF) (2 fans)
Fuel Bldg. Ventilation (non-ESF)
<1%
< 0.05%
<1%
<1%
<1%
<1%
8000 cfm 2000 cfm 45,700 cfm 40,500 cfm 21,000 cfm 35,000 cfm C.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at the temperature and relative humidity (RH) specified below.
ESF Ventilation System Annulus Ventilation Control Area Ventilation Aux. Bldg. Filtered Exhaust Containment Purge (non-ESF)
Fuel Bldg. Ventilation (non-ESF)
< 4%
< 0.95%
< 4%
< 4%
< 4%
RH 95%
95%
95%
95%
95%
Temp.
300C 300C 30 0C 80 0C 800C d.
Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 at the flowrate specified below +/- 10%.
(continued)
McGuire Units 1 and 2 5.5-10 Amendment Nos. 237 218 I
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testinq Proqram (VFTP) (continued)
ESF Ventilation System Annulus Ventilation Control Area Ventilation Aux. Bldg. Filtered Exhaust (2 fans)(Unit 1)
Aux. Bldg. Filtered Exhaust (2 fans)(Unit 2)
Containment Purge (non-ESF) (2 fans)
Fuel Bldg. Ventilation (non-ESF)
Delta P Flowrate 6.0 in wg 5.0 in wg 6.0 in wg 6.0 in wg 6.0 in wg 6.0 in wg 8000 cfm 2000 cfm 45,700 cfm 40,500 cfm 21,000 cfm 35,000 cfm
- e.
Demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ANSI N510-1975.
ESF Ventilation System Annulus Ventilation Control Area Ventilation Wattage @ 600 VAC 43.0 + 6.4 kW 10.0 + 1.0 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
Explosive Gas and Storaae Tank Radioactivity Monitorina Proaram 5.5.12 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 or fed into the offgas treatment system, 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 determined 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 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);
(continued)
McGuire Units 1 and 2 5.5-11 Amendment Nos. 237,218 I
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)
- b.
A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank or connected gas storage tanks and fed into the off gas treatment system is less than the amount that would result in a Deep Dose Equivalent of > 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and
- c.
A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations exceeding 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.
5.5.13 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 ASTM Standards. The purpose of the program is to establish the following:
- a.
Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
- 1.
an API gravity or an absolute specific gravity within limits,
- 2.
a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and
- 3.
a clear and bright appearance with proper color or a water and sediment content within limits;
- b.
Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and (continued)
McGuire Units 1 and 2 5.5-12 Amendment Nos. 237,218
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program (continued)
- c.
Total particulate concentration of the fuel oil is < 10 mg/I when tested evenr 31 days.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.
5.5.14 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 5.5.14.b.1 or 5.5.14.b.2 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 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. The SFDP shall contain the following:
(continued)
McGuire Units 1 and 2 5.5-13 Amendment Nos. 237,218 1
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)
- 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.
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.
McGuire Units 1 and 2 5.5-14 Amendment Nos. 237,218
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 Steam Generator Tube inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG,
- b.
Active degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e.
Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f.
Total number and percentage of tubes plugged to date,
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing, and
- h.
The effective plugging percentage for all plugging in each SG.
(continued)
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