ML062080621
| ML062080621 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 07/27/2006 |
| From: | NRC/NRR/DLPM/LPD4 |
| To: | Arizona Public Service Co |
| Fields M B, NRR/DORL, 415-3062 | |
| Shared Package | |
| ML062090274 | List: |
| References | |
| TAC MC7182, TAC MC7183, TAC MC7184 | |
| Download: ML062080621 (26) | |
Text
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.161, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license.
APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3)
Antitrust Conditions This license is subject to the antitrust conditions delineated in Appendix C to this license.
(4)
Operating Staff Experience Requirements Deleted (5)
Post-Fuel-Loading Initial Test Program (Section 14, SER and SSER 2" Deleted (6)
Environmental Qualification Deleted (7)
Fire Protection Program APS 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, as supplemented and amended, and as approved in the SER through Supplement 11, subject to the following provision:
APS may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
(8)
Emergency Preparedness Deleted The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Amendment No.
161 (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.161, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license.
APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3)
Antitrust Conditions This license is subject to the antitrust conditions delineated in Appendix C to this license.
(4)
Operating Staff Experience Requirements (Section 13.1.2. SSER 9)
Deleted (5)
Initial Test Program (Section 14, SER and SSER 21 Deleted (6)
Fire Protection Program APS 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, as supplemented and amended, and as approved in the SER through Supplement 11, subject to the following provision:
APS may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
(7)
Inservice Inspection Program (Sections 5.2.4 and 6.6. SER and SSER Deleted (8)
Supplement No. 1 to NUREG-0737 Requirements Deleted The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Amendment No.
161 (1)
Maximum Power Level Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3876 megawatts thermal (100% power) through operating cycle 13, and 3990 megawatts thermal (100% power) after operating cycle 13, in accordance with the conditions specified herein:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.161, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. APS shall operate the facility in accordance with the Technical Specifications
-and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
(3)
Antitrust Conditions This license is subject to the antitrust conditions delineated in Appendix C to this license.
(4)
Initial Test Program (Section 14, SER and SSER 2)
Deleted (5)
Additional Conditions Deleted D.
APS has previously been granted an exemption from Paragraph IIl.D.2(b)(ii) of Appendix J to 10 CFR Part 50. This exemption was previously granted in Facility Operating License NPF-65 pursuant to 10 CFR 50.12.
With the granting of this exemption, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.
E.
The licensees shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Palo Amendment No.
161
PALO VERDE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.3 INSTRUMENTATION 3.3.1 Reactor Protective System (RPS) Instrumentation -- Operating 3.3.2 RPS Instrumentation -- Shutdown 3.3.3 Control Element Assembly Calculators (CEACs) 3.3.4 RPS Logic and. Trip Initiation 3.3.5 Engineered Safety Features Actuation System (ESFAS)
Instrumentation 3.3.6 ESFAS Logic and Manual Trip 3.3.7 Diesel Generator (DG)
-- Loss of Voltage Start (LOVS) 3.3.8 Containment Purge Isolation Actuation Signal (CPIAS) 3.3.9 Control Room Essential Filtration Actuation Signal (CREFAS) 3.3.10 Post Accident Monitoring (PAM)
Instrumentation 3.3.11 Remote Shutdown System 3.3.12 Boron Dilution Alarm System (BDAS) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops --
MODES 1 and 2 3.4.5 RCS Loops --
MODE 3 3.4.6 RCS Loops -- MODE 4 3.4.7 RCS Loops -- MODE 5, Loops Filled 3.4.8 RCS Loops -- MODE 5. Loops Not Filled 3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves -- MODES 1, 2. and 3 3.4.11 Pressurizer Safety Valves --
MODE 4 3.4.12 Pressurizer Vents 3.4.13 Low Temperature Overpressure Protection (LTOP) System 3.4.14 RCS Operational Leakage 3.4.15 RCS Pressure Isolation Valve (PIV) Leakage 3.4.16 RCS Leakage Detection Instrumentation 3.4.17 RCS Specific Activity 3.4.18 Steam Generator (SG) Tube Integrity 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Safety Injection Tanks (SITs) -- Operating 3.5.2 SITs -- Shutdown 3.5.3 ECCS -- Operating 3.5.4 ECCS --
Shutdown 3.5.5 Refueling Water Tank (RWT) 3.5.6 Trisodium Phosphate (TSP)
PALO VERDE UNITS 1.2,3 ii AMENDMENT NO.
161
Defini ti ons 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 methodology for verification have been previously reviewed and approved by the NRC.
Kn1 is the K effective calculated by considering the actual CEA configuration and assuming that the fully or partially inserted full strength CEA of highest worth is fully withdrawn.
LEAKAGE shall be:
- a.
Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) sealwater 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
LEAKAGE through a steam generator (SG) to the Secondary System (primary to secondary LEAKAGE).
LEAKAGE I
(continued)
PALO VERDE UNITS 1,2,3 1.1-4 AMENDMENT NO. 4&2, 161
Defi ni tions 1.1 1.1 Definitions LEAKAGE (continued)
- b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE:
- c.
Pressure Boundary LEAKAGE MODE NEUTRON RATED THERMAL POWER (NRTP)
OPERABLE OPERABILITY PHYSICS TESTS LEAKAGE (except primary to secondary LEAKAGE) through a noniso1able fault in an RCS component body, pipe wall, or vessel wall.
A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, cold leg reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
The-indicated neutron flux at RTP.
A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
- a.
Described in Chapter 14, Initial Test Program of the UFSAR:
- b.
Authorized under the provisions of 10 CFR 50.59; or
- c.
Otherwise approved by the Nuclear Regulatory Commission.
I (continued)
PALO VERDE UNITS 1,2,3 1.1-5 AMENDMENT NO.
161
RCS Operational LEAKAGE 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.14 PCS Operational LEAKAGE LD0 3.4.14 RCS operational LEAKAGE shall be limited to:
- a.
No pressure boundary LEAKAGE:
- b.
1 gpm unidentified LEAKAGE:
- c.
10 gpm identified LEAKAGE: and
- d.
150 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 AND not 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 limit.
I PALO VERDE UNITS 1,2,3 3.4.14-1 AMENDMENT NO. 4,24,161
RCS Operational LEAKAGE 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1
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 Perform RCS water inventory balance.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> SR 3.4.14.2 ------------------ 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 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 150 gallons per day through any one SG.
PALO VERDE UNITS 1,2,3 3.4.14-2 AMENDMENT NO. 14-ý, 161
SG Tube fritegrity 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 ACT----------ONS------ 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 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained not plugged in until the next accordance with the refueling outage or Steam Generator SG tube inspection.
Program.
AND A.2 Plug the affected Prior to tube(s) in accordance entering MODE 4 with the Steam 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 AND not 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.
PALO VERDE UNITS 1.2,3 3.4.18-1 AMENDMENT NO. 161
SG Tube Integrity SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program.
with the Steam Generator Program.
SR 3.4.18.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged in accordance with the Steam following a SG Generator Program.
tube inspection.
PALO VERDE UNITS 1,2,3 3.4.18-2 AMENDMENT NO. 161
Programs and Manuals 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG)
Program A Steam Generator (SG)
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 of tubes.
Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected and 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 SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.
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, (continued)
PALO VERDE UNITS 1.2,3 5.5-6 AMENDMENT NO. 424. 161
Programs and Manuals 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG)
Program (continued) 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 0.5 gpm per SG and 1 gpm through both SGs.
- 3.
The operational LEAKAGE performance criterion is specified in LCO 3.4.14, "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.
- 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.l, d.2a, d.2b, 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.
2a.
Original SGs with Alloy 600 MA tubes:
Inspect 100% of the tubes at sequential periods of 60 effective full power months.
The first sequential period shall be (continued)
PALO VERDE UNITS 1,2,3 5.5-7 AMENDMENT NO. 424, 161
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG)
Program (continued) considered to begin after the first inservice inspection of the SGs.
No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.
2b.
Replacement*SGs with Alloy 690 TT tubes:
Inspect 100% of the tubes at sequential periods of 144, 108, 92, 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)
PALO VERDE UNITS 1.2,3 5.5-8 AMENDMENT NO.
J-2-G, 161
Programs and Manuals S55 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 Testing Program (VFTP)
A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, and in accordance with Regulatory Guide 1.52.
Revision 2 and ANSI N510-1980 at the system flowrate specified below +/- 10%.
- 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 - 1.0 % when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980. at the system flowrate specified as follows
+/- 10%:
(continued)
PALO VERDE UNITS 1,2,3 5.5-9 AMENDMENT NO. 440. 161
Programs and Manual5.5-5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP)
(continued)
ESF Ventilation System Control Room Essential Filtration System (CREFS)
Engineered Safety Feature (ESF)
Pump Room Exhaust Air Cleanup System (PREACS)
- b.
Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass ! 1.0 % when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the system flowrate specified as follows +/- 10%:
ESF Ventilation System CREFS Flowrate 28,600 CFM 6,000 CFM ESF PREACS
- c.
Demonstrate for each of the ESF systems that a charcoal adsorber sample, when obtained in accordance with the application of Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, as described in Section 1.8 of the UFSAR, 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 300C and to the relative humidity specified as follows:
ESF Ventilation System CREFS Penetration 2.5%
2.5%
RH 70%
70%
ESF PREACS (continued)
PALO VERDE UNITS 1.2,3 5.5-10 AMENDMENT NO. 44G, 161
Programs and Manual]s
- r 5.5 Programs and Manuals (continued) 5.5.11 Ventilation Filter Testing Program (VFTP)
(continued)
- d.
For each of the ESF systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than or equal to the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980 at the system flowrate specified as follows +/- 10%:
ESF Ventilation System Delta P Flowrate CREFS 4.8 inches water gauge 28,600 CFM ESF PREACS 5.2 inches water gauge 6,000 CFM
- e.
Demonstrate that the heaters for each of the ESF systems dissipate the following specified value when tested in accordance with ANSI N510-1980:
ESF Ventilation System Wattage ESF PREACS
> 19 kW The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.
5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides control 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 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".
(continued)
PALO VERDE UNITS 1.2.3 5.5-11 AMENDMENT NO.
4422. 161
Programs and Manuals 5.5 Programs and Manuals (continued) 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)
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 Ž 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents:
and
- c.
A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR Part 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 as referenced in the UFSAR.
The purpose of the program is to establish the following:
(continued)
PALO VERDE UNITS 1.2,3 5.5-12 AMENDMENT NO.
-L29, 161
Programs and Manuals 5.5 Programs and Manuals (continued) 5.5.13 Diesel Fuel Oil Testing Program (continued)
- 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.
Water and sediment within limits when tested in accordance with ASTM D1796:
- b.
Other properties for ASTM 2D fuel oil within 31 days following sampling and tanks; and are within limits addition to storage 5.5.14
- c.
Total particulate concentration of the stored fuel oil is
! 10 mg/l when tested every 92 days in accordance with ASTM D-2276. Method A-2 or A-3.
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 approval to the updated FSAR or Bases that requires NRC 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)
PALO VERDE UNITS 1,2,3 5.5-13 AMENDMENT NO. 1P, 161
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Technical Specifications (TS)
Bases Control Program (continued)
- d.
Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
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 limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exceptionto 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, 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:
- a.
A required system redundant to system(s) supported by the inoperable support system is also inoperable: or (continued)
PALO VERDE UNITS 1.2.3 5.5-14 AMENDMENT NO 161
Programs and Manuals 5.5 Programs and Manuals (continued) 5.5.15 Safety Function Determination Program (continued)
- b.
A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
- c.
A required system redundant to 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.
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.
5.5.16 Containment Leakage Rate Testing Program
- a.
A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.
This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:
- 1.
The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Code Section XI. Subsection IWL, except where relief has been authorized by the NRC.
The containment concrete visual examination may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage.
- 2.
The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Code-Section XI, Subsection IWE, except where relief has been authorized by the NRC.
(continued)
PALO VERDE UNITS 1, 2 AND 3 5.5-15 AMENDMENT NO. 4-54,161
Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 Containment Leakage Rate Testing Program (continued)
- b.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 52.0 psig for Units 1 and 3, and 58.0 psig for Unit 2.
The containment design pressure is 60 psig.
- c.
The maximum allowable containment leakage rate, La, at Pa. shall be 0.1 % 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 are < 0.60 La for the Type B and C tests and
- 0.75 La for Type A tests.
- 2.
Air lock testing acceptance criteria are:
a)
Overall air lock leakage rate is
- 0.05 La when tested at Ž Pa.
b)
For each door, leakage rate is
- 0.01 La when pressurized to 2 14.5 psig.
- e.
The provisions of SR 3.0.2 do not apply to the test frequencies in the Containment Leakage Rate Testing Program.
- f.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
PALO VERDE UNITS 12.35 5.5-16 AMENDMENT NO.
161
Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 PAM Report When a report is required by Condition B or G of LCO 3.3.10, "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 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days.
The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking. and the corrective action taken.
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.
PALO VERDE UNITS 1,2.3 5.6-6 AMENDMENT NO. -I-8i, 161