ML16161A279
| ML16161A279 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/26/1998 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML16161A280 | List: |
| References | |
| NUDOCS 9803130145 | |
| Download: ML16161A279 (39) | |
Text
11 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CORPORATION DOCKET NO. 50-269 OCONEE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 228 License No. DPR-38
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility)
Facility Operating License No. DPR-38 filed by the Duke Energy Corporation (the licensee) dated February 2, 1998, as supplemented February 18, 1998, complies with the standards and requirements of the>Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-38 is hereby amended to read as follows:
9803130145 980226 PDR ADOCK 05000269 P-PDR
-2 B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.
228, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and will be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Berkow? 2 to er ert N.
Di ector roject Directorate 11-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
February 26, 1998
o A UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CORPORATION DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.229 License No. DPR-47
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility)
Facility Operating License No. DPR-47 filed by the Duke Energy Corporation (the licensee) dated February 2, 1998, as supplemented February 18, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-47 is hereby amended to read as follows:
-2 B. Technical SDecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 229, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and will be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION erbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
February 26, 1998
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CORPORATION DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
225 License No. DPR-55 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility)
Facility Operating License No. DPR-55 filed by the Duke Energy Corporation (the licensee) dated February 2, 1998, as supplemented February 18, 1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:
-2 B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 225, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and will be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION erbert N. Berkow, Director Project Directorate 11-2 Division' of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
February 26, 1998
ATTACHMENT TO LICENSE AMENDMENT NO. 228 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO. 229 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 225 FACILITY OPERATING LICENSE NO. DPR-55 DOCKET NO. 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Insert 4.0-1 4.0-1 4.1-3 4.1-3 4.1-4 4.1-4 4.1-5 4.1-5 4.1-6 4.1-6 4.1-7 4.1-7 4.1-8 4.1-8 4.1-8a 4.1-8a 4.1-8b 4.1-9 4.1-9 4.2-2 4.2-2 4.4-17 4.4-17 4.4-20 4.4-20 4.5-1 4.5-1 4.5-2 4.5-2
-2 Remove Insert 4.5-4 4.5-4 4.5-6 4.5-6 4.5-7 4.5-7 4.5-8 4.5-8 4.5-9 4.5-9 4.6-1 4.6-1 4.7-1 4.7-1 4.8-1 4.8-1 4.9-1 4.9-1 4.12-1 4.12-1 4.14-1 4.14-1 4-14-2 4.14-2 4.18-1 4.18-1
. 4.18-2 4.18-2 4.18-4 4.18-4 4.20-5 4.20-5
4 SURVELLANCE REQUIREMENTS 4.0 SURVEILLANCE STANDARDS Applicability Applies to surveillance requirements which relate to tests, calibrations and inspections necessary to assure that the quality of structures, systems and components is maintained and that operation is within the safety limits and limiting conditions for operation.
Objective To specify minimum acceptable surveillance requirements.
Specification 4.0.1 Surveillance of structures, systems, components and parameters shall be as specified in the various subsections to this Technical Specification section, Section 4.0, except as permitted by Technical Specifications 4.0.2 and 4.0.3 below.
4.0.2 Minimum surveillance frequencies, unless' specified otherwise, may be adjusted as follows to facilitate test scheduling:
Maximum Allowable Specified Frequency Interval Between Surveillances Five times per week 2 days Two times per week 5 days Weekly 10 days Bi-Weekly 20 days Monthly 45 days Bi-Monthly 90 days Quarterly 135 days Semiannually 270 days Annually 18 months 18 months 22 months, I5 days Refueling Outage 22 months, 15 days Clarifying words in individual specifications such as "every," "at least," or "at least once every" are not intended to alter the frequencies defined by this specification.
4.0.3 If conditions exist such that surveillance of an item is not necessary to assure that operation is within the safety limits and limiting conditions for operation, surveillance need not be performed if such conditions continue for a length of time greater than the specified surveillance interval.
Surveillance waived as a result of this specification shall be performed prior to returning to condi tions for which the surveillance is necessary to assure that operation is within safety limits and limiting conditions for operation.
4.0.4 Inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50 Section 50.55a(gX4) to the extent practicable within the limitations of design, geometry and materials of construction of the components.
Oconee 1, 2, and 3 4.0-1 Amendment No.228(Unit 1)
Amendment No. 229(Unit 2)
Amendment No. 22 5(Unit 3)
Table 4.1-1 INSTRUMENT SURVEILLANCE REQUIREMENTS Channel Description Check Test Calibrate Remarks
- 1.
Protective Channel NA MO NA Coincidence Logic in the Reactor Trip Modules
- 2.
Control Rod Drive NA MO(l)
NA (1)
This test shall independently Trip Breaker, SCR confirm the operability of the Control Relays E and F shunt trip device and the undervoltage device.
- 3.
Power Range Amplifier ES(l)
NA (1)
(1)
Heat balance check each shift.
Heat balance calibration whenever indicated core thermal power exceeds neutron power by more than 2 percent.
- 4.
Power Range ES 45 Days MO(1)(2)
(1)
Using incore instrumentation.
STB (2)
Axial offset upper and lower chambers after each startup if not done previous week.
- 5.
Wide Range ES(1)
PS NA (1)
When in service.
- 6.
Source Range ES(1)
PS NA (1)
When in service.
- 7.
Reactor Coolant ES 45 Days 18 months Temperature STB
- 8.
High Reactor Coolant ES 45 Days 18 months Pressure STB
- 9.
Low Reactor Coolant ES 45 Days 18 months Pressure STB Oconee 1, 2, and 3 4.1-3 Amendment No. 228(Unit 1)
Amendment No. 22 nit 2)
Amendment No. 22 Unit 3)
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 10.
Flux-Reactor Coolant ES 45 Days 18 months Flow Comparator STB
- 11.
Reactor Coolant Pressure ES 45 Days 18 months Temperature Comparator STB
- 12.
Pump-Flux Comparator ES 45 Days 18 months STB
- 13.
High Reactor Building DA 45 Days 18 months Pressure STB
- 14.
High Pressure Injection &
NA MO NA Includes Reactor Building Reactor Building Isolation Isolation of non-essential Logic (Non-essential systems) systems
- 15.
High Pressure Injection Analog Channels:
- a.
Reactor Coolant ES MO 18 months Pressure
- b.
Reactor Building ES MO 18 months Pressure (4 psig)
- 16.
Low Pressure Injection NA MO NA Logic Oconee 1, 2, and 3 4.1-4 Amendment No.228(Unit 1)
Amendment No.2 2 9(Unit 2)
Amendment No.2 2 5(Unit 3)
TABLE 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 17.
Low Pressure Injection Analog Channels:
- a.
Reactor Coolant ES MO 18 months Pressure
- b.
Reactor Building ES MO 18 months Pressure (4 psig)
- 18.
Reactor Building Emergency NA MO NA Reactor Building isolation Cooling and Isolation includes essential systems System Logic (Essential Systems)
- 19.
Reactor Building Emergency ES MO 18 months Cooling and Isolation System Analog Channel Reactor Building Pressure (4 psig)
- 20.
Reactor Building Spray NA MO NA System Logic
- 21.
Reactor Building Spray NA MO 18 months System Analog Channel Reactor Building High Pressure
- 22.
Pressurizer Temperature ES NA 18 months
- 23.
Control Rod Absolute ES(l)
NA 18 months (2)
(1) Check with Relative Position Indicator.
Position (2) Calibrate rod misalignment channel.
Oconee 1, 2, and 3 4.1-5 Amendment No. 228(Unit 1)
Amendment No. 2 2 9(Unit 2)
Amendment No.22 5(Unit 3)
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 24.
Control Rod Relative ES(l)
NA 18 months (2)
(1) Check with Absolute Position Indicator.
Position (2) Calibrate rod misalignment channel.
- 25.
Core Flood Tanks
- a.
Pressure ES NA 18 months
- b.
Level ES NA 18 months
- 26.
Pressurizer Level ES NA 18 months I
- 27.
Letdown Storage Tank DA NA 18 months Level
- 28.
Delete
- 29.
High and Low Pressure NA NA 18 months j
Injection Systems Flow Channels
- 30.
Borated Water Storage WE NA 18 months Tank Level Indicator
- 31.
Boric Acid Mix Tank:
- a.
Level NA NA AN
- b.
Temperature MO NA AN
- 32.
Concentrated Boric Acid Storage Tank:
- a.
Level NA NA AN
- b.
Temperature MO NA AN Oconee 1, 2, and 3 4.1-6 Amendment No.228 (Unit 1)
Amendment No. 2 29 (Unit 2)
Amendment No.225(Unit 3)
.-Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 33.
Containment Temperature NA NA 18 months
- 34.
Incore Neutron Detectors MO(1)
NA NA (1)
Check functioning; including functioning of computer readout or recorder readout.
- 35.
Emergency Plant MO(1)
NA 18 months (1)
Battery check.
Radiation Instruments
- 36.
Environmental Monitors MO(1)
NA 18 months (1)
Check functioning.
- 37.
Reactor Manual Trip NA PS NA
- 38.
Reactor Building Emergency NA NA 18 months Sump Level
- 39.
Steam Generator Water Level WE NA 18 months
- 40.
Turbine Overspeed Trip NA NA 18 months
- 41.
Engineered Safeguards NA 18 months NA Includes Reactor Building isolation Channel 1 HP Injection &
of non-essential systems only Reactor Building Isolation Manual Trip
- 42.
Engineered Safeguards NA 18 months NA Includes Reactor Building isolation Channel 2 HP Injection &
of non-essential systems only Reactor Building Isolation Manual Trip
- 43.
Engineered Safeguards NA 18 months NA Channel 3 LP Injection Manual Trip Oconee 1, 2, and 3 4.1-7 Amendment No. 228(Unit 1)
Amendment No. 229(Unit 2)
Amendment No. 225(Unit 3)
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 44.
Engineered Safeguards NA 18 months NA Channel 4 LP Injection Manual Trip
- 45.
Engineered Safeguards NA 18 months(1)
NA Includes Reactor Building isolation of essential systems only.
Channel 5 RB Isolation (1) A one-time extension of the test frequency to a maximum of
& Cooling Manual Trip 23 months is allowed for Oconee Unit 2 during operating cycle 16.
- 46.
Engineered Safeguards NA 18 months(1)
NA Includes Reactor Building isolation of essential systems only.
Channel 6 RB Isolation (1) A one-time extension of the test frequency to a maximum of
& Cooling Manual Trip 23 months is allowed for Oconee Unit 2 during operating cycle 16.
- 47.
Engineered Safeguards NA 18 months NA Channel 7 Spray Manual Trip
- 48.
Engineered Safeguards NA 18 months NAj Channel 8 Spray Manual Trip
- 49.
Emergency Feedwater MO NA 18 months Flow Indicators
- 50.
PORV and Safety Valve MO NA 18 months Position Indicators
- 51.
RPS Anticipatory NA 45 Days 18 months Reactor Trip System Loss ST3 of Turbine Emergency Trip System Pressure Switches Oconee 1, 2, and 3 4.1-8 Amendment No. 22(Unit 1)
Amendment No. 22qUnit 2)
Amendment No. 2 2 Unit 3)
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 52.
RPS Anticipatory Reactor Trip System Loss of Main Feedwater a)
Control Oil Pressure NA 45 Days 18 months Switches STB
- 53.
Emergency Feedwater Initiation Circuits a)
Control Oil Pressure NA MO 18 months j
Switches
- 54.
Containment High Range NA MO 18 months TMI Item II.F.1.3 Radiation Monitor (RIA-57, 58)
- 55.
Containment Pressure MO NA AN TMI Item II.F.1.4 Monitor (PT-230, 231)
- 56.
Containment Water Level MO NA 18 months TMI Item II.F.1.5 Monitor-Wide Range (LT-90, -91)
- 57.
Containment Hydrogen NA MO AN TMI Item II.F.1.6 Monitor (MT-80,-81)
- 58.
Wide Range Hot Leg Level NA 18 months(1) 18 months(1)
(1) A one-time extension of the channel test and calibration frequency to a maximum of 24 months is allowed for Oconee Unit 2 during operating cycle 16.
- 59.
Reactor Vessel Head Level NA 18 months(1) 18 months(1)
(1) A one-time extension of the channel test and calibration frequency to a maximum of 24 months is allowed for Oconee Unit 2 during operating cycle 16.
Oconee 1, 2, and 3 4.1-8a Amendment No. 228(Unit 1)
Amendment No. 229Unit 2)
Amendment No. 22EUnit 3)
Table 4.1-1 (CONTINUED)
Channel Description Check Test Calibrate Remarks
- 60.
Core Exit Thermocouples MO NA 18 months(1)
(1) A one-time extension of the calibration frequency to a maximum of 24 months is allowed for Oconee Unit 2 during operating cycle 16.
- 61.
Subcooling Monitors MO 18 months(1) 18 months(l)
(1) A one-time extension of the channel test and calibration frequency to a maximum of 24 months is allowed for Oconee Unit 2 during operating cycle 16.
ES - Each Shift QU - Quarterly DA - Daily AN - Annually WE - Weekly PS - Prior to startup, if not performed previous week MO - Monthly NA - Not Applicable STB - STAGGERED TEST BASIS Oconee 1, 2, and 3 4.1-8b Amendment No. 228(Unit 1)
Amendment No. 229(Unit 2)
Amendment No.2 2 5(Unit 3)
Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test FreQuency
- 1.
Control Rod Movement )
Movement of Each Rod Monthly
- 2.
Pressurizer Safety Valves Setpoint 18 months4
- 3.
Main Steam Safety Valves Setpoint 18 months(4)
- 4.
Refueling System Interlocks Functional Prior to Refueling
- 5.
Main Steam Stop Valves ()
Movement of Each Stop Monthly Valve
- 6.
Evaluate Daily Leakage
- 7.
Condenser Circulating Water 6)
Functional 18 months Flow Test
- 8.
High Pressure Service Functional Monthly Water Pumps and Power Supplies
- 9.
Spent Fuel Cooling System Functional Prior to Refueling
- 10.
High Pressure and Low. )
Vent Pump Casings Monthly and Prior Pressure Injection System to Testing
- 11.
Emergency Feedwater Functional 18 months Pump Automatic Start and Automatic Valve Actuation Feature
()
Applicable only when the reactor is critical.
()
Applicable only when the reactor coolant is above 200*F and at a steady-state temperature and pressure.
()
Operating pumps excluded.
(4)
Number of safety valves to be tested every 18 months shall be in accordance with ASME CodesSection XI, Article IWV-351 1, such that each valve is tested at least once every 5 years.
(s>
Applicable.only to the interlocks associated with the Reactor Building Purge System.
(6)
Verification of the Emergency Condenser Circulating Water (ECCW) System function to supply siphon suction to the Low Pressure Service Water System shall be performed to ensure operability of the LPSW System.
Oconee 1, 2, and 3 4.1-9 Amendment No.228(Unit 1)
Amendment No. 2 2 (Unit 2)
Amendment No.2 2 5(Unit 3)
4.2.6 The power operated relief valve (PORV) is used for low temperature overpressure protection of the RCS and shall be demonstrated operable by:
- a.
Performing an operability test prior to each startup from cold shutdown.
- b.
Performing a calibration of the actuation circuit every 18 months.
- c.
Performing an inspection of the PORV at least once every two refueling cycles.
4.2.7 Each shift, the RCS vent(s) (as defined in Specification 3.1.2.9) shall be verified to be open, if the vent(s) is(are) being used for overpressure protection. If the vent pathway is provided wiih a valve which is locked, sealed, or otherwise secured in the open position, then these valves will open at least once per 31 days.
Bases The surveillance program has been developed to comply with the applicable edition of Section XI and addenda of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor Coolant Systems, as required by 10 CFR 50.55(a) to the extent practicable within limitations of design, geometry and materials of construction. The program places mkjor emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties.
Oconee 1, 2, and 3 4.2-2 Amendment No. 2 2 8(Unit 1)
Amendment No. 2 2 9(Unit 2)
Amendment No. 225(Unit 3)
4.4.3 Containment Hydrogen Control Systems App2licability Applies to the Containment Hydrogen Control Systems.
Objective To verify that the Containment Hydrogen Control Systems are operable.
Specifications 4.4.3.1 Containment Hydrogen Control System Piping Every 18 months, the permanent piping for the Containment Hydrogen Control System shall be tested as follows:
- a. The post-LOCA flow paths shall be verified by connecting and operating either the Hydrogen Purge Unit or the Hydrogen Recombiner through each flow path as follows:
- 1. The hydrogen Recombiner flow path circulates Reactor Building atmosphere at a flow greater than 50 SCFM.
- 2. The Hydrogen Purge flow path removes Reactor Building atmosphere and discharges to the Unit vent stack at a flow greater than or equal to 45 SCFM.
- b. The blind isolation flanges on the Containment Hydrogen Control System permanent piping shall be leak tested after each installation to ensure adequate isolation.
4.4.3.2 Containment Hydrogen Recombiner System Operational Performance Testing
- a. The testing requirement of this section may be performed without connecting the system to either of the Reactor Buildings.
- b. Every 18 months:
- 1. Visual inspection of the unit.
- 2. Calibrate all recombiner instrumentation and control circuits.
- 3. Operate a recombiner unit at design flow rate 10% and allow unit to reach recombination temperature.
4.4.3.3 Reactor Building Hydrogen Purge System, Pre-Operational Testing
- a. Prior to declaring this system operable, a Pre-operational system test shall be performed.
Oconee 1, 2, and 3 4.4-17 Amendment No. 228(Unit 1)
Amendment No. 229(Unit 2)
Amendment No. 225(Unit 3)
4.4.4 Reactor Building Purme System AVplicability Applies to the Reactor Building Purge System.
Objective To verify that the Reactor Building Purge System is operable.
SNecification 4.4.4.1 Each shutdown, when the purge valves have been operated, leakage integrity tests shall be performed on the containment purge isolation valves after final closing and prior to going above hot shutdown. If the purge valves have not been operated, leakage integrity tests shall be performed prior to going above hot shutdown unless such, tests have been conducted within the proceeding six months. If the acceptance criteria of Specification 4.4.1.23 are not met, Specification 3.6.6 shall apply. Unit shutdown to conduct the test and/or effect repairs is specifically not required.
4.4.4.2 Monthly, when the unit is above 250*F and 350 psig, the containment purge isolation valves shall be verified closed.
4.4.4.3 Every 18 months, the valve seals of the containment purge isolation valves shall be visually inspected and adjusted or replaced as appropriate.
4.4.4.4 Prior to use of the purge system at conditions between cold shutdown and 250*F and 350 psig, the isolation valves shall be exercise tested in accordance with the requirements (except test frequency) of the applicable edition of the ASME Boiler and Pressure Vessel Code,Section XI.
4.4.4.5 The pneumatically operated purge isolation valves shall be verified to close in response to a control signal from RIA-45 when the system is tested prior to refueling operations per Specification 3.8.10.
Bases Leakage integrity tests of the purge supply and isolation valves are conducted in order to identify excessive degradation of the resilient seals. Excessive leakage past resilient seals is typically caused by severe environmental conditions and/or wear due to frequent use.
The pneumatically operated purge isolation valves are tested prior to refueling operations because the only automatic isolation system in service at refueling is through RIA-45, which only closes the pneumatic isolation valves.
Oconee 1, 2, and 3 4.4-20 Amendment No. 228(Unit 1)
Amendment No. 229(Unit 2)
Amendment No. 225(Unit 3)
4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR BUILDING COOLING SYsmM PERIODIC TESTING 4.5.1 Emergency Core Cooling Systems Applicability Applies to periodic testing requirements for the Emergency Core Cooling Systems.
Objective To verify that the Emergency Core Cooling Systems are operable.
Specification 4.5.1.1 System Tests 4.5.1.1.1 High Pressure Injection System
- a.
Every 18 months, a system test shall be conducted to demonstrate that the system is operable. A test signal will be applied to demonstrate actuation of the High Pressure Injection System for emergency core cooling operation.
- b.
The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly; all appropriate pump breakers shall have opened or closed and all valves shall have completed their travel.
4.5.1.1.2 Low Pressure Injection System
- a.
Every 18 months, a system test shall be conducted to demonstrate that the system is operable. The test shall be performed in accordance with the procedure summarized below:
(1) A test signal will be applied to demonstrate actuation of the Low Pressure Injection System for emergency core cooling operation.
(2) Verification of the engineered safety features function of the Low Pressure Service Water pumps and manual alignment from the control room of valves LPSW-4 and LPSW-5 shall be made to demonstrate operability of the Low Pressure Injection coolers.'
- b.
The test will be considered satisfactory if control board indication verifies that all components have responded to the ES actuation signal properly; all appropriate ES actuated pump breakers shall have opened or closed, and all ES actuated valves shall have completed their travel. In addition, valves LPSW-4 and LPSW-5 shall have completed their travel.
The ES function of valves LPSW-4 and LPSW-5 shall be verified every 18 months. This surveillance requirement. may be discontinued and replaced by the valve surveillance in 4.5.1.1.2.a.(2) when the ES signals are removed from LPSW-4 and LPSW-5. Removal of the ES signal from valves LPSW-4 and LPSW-5 is scheduled in the U3EOC 16, UIEOCl7, and U2EOCl6 refueling outages successively.
Oconee 1. 2, and 3 4.5-1 Amendment No228 (Unit 1)
Amendment No229 (Unit 2)
Amendment No.2 2 5 (Unit 3)
4.5.1.13 Core Flooding System
- a.
Every 18 months, a system test shall be conducted to demonstrate proper operation of the system.
During pressurization of the Reactor Coolant System, verification shall be made that the check and isolation valves in the core flooding tank discharge lines operate properly.
- b.
The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened.
45.1.2 Component Tests 4.5.1.2.1 Valves - Power Operated
- a.
Valves LP-17, -18, shall only be tested every cold shutdown unless previously tested during the current quarter.
- b.
Every 18 months, the following LPI system valves shall be cycled manually to verify the manual operability of these power operated valves:
(1)
LPI pump discharge (ES) LP-17,-18 (2)
LPI discharge throttling LP-12,-14 (3)
LPI discharge header crossover LP-9,-10 (4)
LPI discharge to HPI/RBS LP-15,-16 4.5.1.2.2 Check Valves Periodic individual leakage testings of valves CF-12, CF-14, LP-47 and LP-48 shall be accomplished prior to power operation after every time the plant is placed in the cold shutdown condition for refueling, after each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is performed. Whenever integrity of these valves cannot be demonstrated, the integrity of the remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of the other closed valve located in the high pressure piping shall be recorded daily. For the allowable leakage rates and limiting conditions for operation, see Technical Specification 3.1.6.10.
Bases The Emergency Core Cooling Systems are the principle reactor safety features in the event of loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.
The High Pressure Injection System under normal operating conditions has one pump operating. The HPI system test required by Specification 4.5.1.1.1 verifies that the HPI system responds as required to actuation of ES channels I and 2.
(a)
To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
Oconee 1, 2, and 3 4.5-2 Amendment No. 228(Unit 1)
Amendment No. 2 2 9(Unit 2)
Amendment No. 2 2 5(Unit 3)
4.5.2 Reactor Building Cooling Systems ApplicabilitY Applies to testing of the Reactor Building Cooling Systems.
Objective To verify that the Reactor Building Cooling Systems are operable.
Specification 4.5.2.1 System Tests 4.5.2.1.1 Reactor Building Spray System
- a.
(1) Every 18 months, a system test shall be conducted to demonstrate proper operation of the system. A test signal will be applied to demonstrate actuation of the Reactor Building Spray System.
(2) The test will be considered satisfactory if visual observation and control board indication verifies that all components have responded to the actuation signal properly; the appropriate pump breakers shall have closed, and all valves shall have completed their travel.
- b.
Station compressed air will be introduced into the spray headers to verify the availability of the headers and spray nozzles at least every ten years.
4.5.2.1.2 Reactor Building Cooling System
- a.
Every 18 months', a system test shall be conducted to demonstrate proper operation of the system. The test shall be performed in accordance with the procedure summarized below:
(1) A test signal will be applied to actuate the Reactor Building Cooling System for reactor building cooling operation.
(2) Verification of the engineered safety features function of the Low Pressure Service Water System which supplies coolant to the reactor building coolers shall be made to demonstrate operability of the coolers.
- b.
The test will be considered satisfactory if control board indication verifies that all components have responded to the actuation signal properly, the appropriate valves have completed their travel, and fans are running at half speed.
Oconee 1, 2, and 3 4.5-4 Amendment No.228 (Unit 1)
Amendment No.2 2 9 (Unit 2)
Amendment No.2 2 5 (Unit 3)
'A one-time extension of the Reactor Building Cooling system test frequency to a maximum of 23 months is allowed for Oconee Unit 2 during operating cycle 16.
4.5.3 Containment Heat Removal Capability Aulicability Applies to verification of adequate containment heat removal capability.
Objective To verify that containment heat removal capability is sufficient to maintain post accident conditions within design limits.
Specification 4.5.3.1 Containment Heat Removal Capability
- a.
Every 18 months, containment heat removal capability shall be verified to be sufficient to maintain post accident conditions within design limits.
- b.
In addition to the requirements of 4.53.1.a, on a frequency consistent with the LPI cooler and RBCU fouling rate, containment heat removal capability shall be verified to be sufficient to maintain post accident conditions within design limits.
Bases The safety functions of the LPI system, RB Spray system, and RBCUs include maintaining containment pressure and temperature below design limits following an accident. This surveillance assures that containment heat removal capability is adequate assuming a worst case single failure. Specification 4.53.1.a requires that at a minimum the surveillance be performed every 18 months. In addition, since service induced fouling can reduce containment heat removal capability, Specification 4.5.3.1.b requires that a fouling rate be determined in order to establish a more frequent test interval if required.
REFERENCES:
FSAR Section 6.2 FSAR Section 15.14 Oconee 1, 2, and 3 4.5-6 Amendment No.228(Unit 1)
Amendment No.229(Unit 2)
Amendment No.225(Unit 3)
4.5.4 Penetration Room Ventilation System Applicability Applies to testing of the Penetration Room Ventilation System Objective To verify that the Penetration Room Ventilation System is operable.
Specificati6n 4.5.4.1 Operational and Performance Testing
- a. Monthly, each train of the Penetration Room Ventilation System shall be operated for at least 15 minutes at design flow *10%.
- b. Every 18 months, it shall be demonstrated that:
- 1. The Penetration Room Ventilation System fans operate at design flow (+/- 10%) when tested in accordance with ANSI N510-1975.
- 2. The pressure drop across the combined HEPA filters and charcoal adsorber banks is less than six inches of water at the system design flow rate (t 10%).
- 3. Each branch of the Penetration Room Ventilation System is capable of automatic initiation.
- 4. The bypass valve for filter cooling is manually operable.
- c. Leak tests using DOP or halogenated hydrocarbon, as appropriate shall be performed on the Penetration Room purge filters:
- 1. Every 18 months;
- 2. After each complete or partial replacement of a HEPA filter bank or charcoal adsorber bank;
- 3. After any structural maintenance on the system housing;
- 4. After painting, fire, or chemical release in any ventilation zone comnunicating with the system.
- d. The results of the DOP and halogenated hydrocarbon tests on HEPA filters and charcoal adsorber banks shall show 99% DOP removal and 99% halogenated hydrocarbon removal, respectively, when tested in accordance with ANSI N510-1975.
Oconee 1, 2, and 3 4.5-7 Amendment No2 28 (Unit 1)
Amendment No? 2 9 (Unit 2)
Amendment No? 2 5 (Unit 3)
- e.
Every 18 months, or following 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, or after painting, fire, or chemical release in any ventilation zone communicating with the system, a carbon sample shall be removed from the Penetration Room Ventilation system filters for laboratory analysis.
Within 31 days of removal, this sample shall be verified to show ?90%h radioactive methyl iodide removal when tested in accordance with ASTM D3803-1989 (30C, 95% R.H.).
Otherwise, the filter system shall be declared inoperable.
Bases Pressure drop across the combined high efficiency particulate air (HEPA) filters and charcoal adsorbers of less than six inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. A test frequency of once per year operating cycle establishes performance capability.
(HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radioiodine. Bypass leakage for the charcoal adsorbers and particulate removal efficiency for HEPA filters are determined by halogenated hydrocarbon and DOP respectively. The laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the performances are as specified, the calculated doses would be less than the guidelines stated in 10 CFR 100 for the accidents analyzed.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be replaced. Any HEPA filters found defective should be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.
Operation of the system every month will demonstrate operability of the filters and adsorber system.
Operation for 15 minutes demonstrates operability and minimizes the moisture build up during testing.
If painting, fire or chemical release occurs during system operation such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign materials, the same tests and sample analysis should be performed as required for operational use.
Demonstration of the automatic initiation capability is necessary to assure system performance capability.
Oconee 1, 2, and 3 4.5-8 Amendment No.228 (Unit 1)
Amendment No.229 (Unit 2)
Amendment No.2 2 5 (Unit 3)
4.5.5 Low Pressure Injection System Leakage Applicability Applies to Low Pressure Injection System leakage.
Objective To maintain a preventive leakage rate for the Low Pressure Injection System which will prevent significant off-site exposures.
Specification 4.5.5.1 Acceptance Limit The maximum allowable leakage from the Low Pressure Injection System components (which includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.
4.5.5.2 Test Every 18 months, the following tests of the Low Pressure Injection System shall be conducted to determine leakage:
- a.
The portion of the Low Pressure Injection System, except as specified in (b), that is outside the containment shall be tested either by use in normal operation or by hydrostatically testing at 350 psig.
- b.
Piping from the containment emergency sump to the low pressure injection pump suction isolation valve shall be pressure tested at no less than 59 psig.
- c.
Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collection and weighing or by another equivalent method.
Bases The leakage rate limit for the Low Pressure Injection System is a judgement value based on assuring that the components can be expected to operate with-out mechanical failure for a period on the order of 200 days after a loss of coolant accident. The test pressure (350 psig) achieved either by normal system operation or by hydrostatically testing, gives an adequate margin over the highest pressure within the-system after a design basis accident. Similarly, the pressure test for the return lines from the containment to the Low Pressure Injection System (59 psig) is equivalent to the design pressure of the containment. The dose to the thyroid calculated as a result of this leakage is 0.76 rem for a two-hour exposure at the site boundary.
REFERENCE FSAR, Section 15.15.4, and 6.3.3.2.2 Oconee 1, 2, and 3 4.5-9 Amendment No2 28 (Unit 1)
Amendment No2 2 9 (Unit 2)
Amendment No. 2 5 (Unit 3)
4.6 EMERGENCY POWER PERIODIC TESTING Applicability Applies to the periodic testing surveillance of the emergency power sources.
Objective To verify that the emergency power sources and equipment will respond promptly and properly when required.
Specification 4.6.1 Monthly, a test of the Keowee Hydro units shall be performed to verify proper operation of these emergency power sources and associated equipment. This test shall assure that:
- a. Each hydro unit can be automatically started from the Unit I and 2 control room.
- b. Each hydro unit can be synchronized through the 230 Kv overhead circuit to the startup transformers.
- c.
Each hydro unit can energize the 13.8 Kv underground feeder.
- d. The 4160 volt startup transformer main feeder bus breakers and standby bus breaker shall be exercised.
4.6.2
- a. Annually, the Keowee Hydro units will be started using the emergency start circuits in each control room to verify that each hydro unit and associated equipment is available to carry load within 25 seconds of a simulated requirement for engineered safety features.
- b. Promptly following the above annual test, each hydro unit will be loaded to at least the combined load of the auxiliaries actuated by ESG signal in one unit and the auxiliaries of the other two units in hot shutdown by synchronizing the hydro unit to the offsite power system and assuming the load at the maximum practical rate.
- c. Also, the ability of the Keowee Unit ACBs to close automatically to the underground path will be tested on an annual frequency.
4.6.3 Monthly, the Keowee Underground Feeder Breaker Interlock shall be verified to be operable.
4.6.4 Every 18 months, a simulated emergency transfer of the 4160 volt main feeder buses to the startup Oconee 1, 2, and 3 4.6-1 Amendment No.228 (Unit 1)
Amendment No.2 2 9 (Unit 2)
Amendment No.2 2 5 (Unit 3)
4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7.1 Control Rod Trip Insertion Time Test Applicability Applies to the surveillance of the control rod trip insertion time.
Objective To assure the control rod trip insertion time is within that used in the safety analyses.
Specification The control rod insertion time shall be measured at either full flow or no flow conditions as follows:
- a.
For all rods following each removal of the reactor vessel head,
- b.
For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and
- c.
For all rods at least once every 18 months.
The maximum control rod trip insertion time for an operable control rod drive mechanism, except for the Axial Power Shaping Rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66* seconds at reactor coolant full flow conditions or 1.40 seconds for no flow conditions.
For the APSRs it shall be demonstrated that loss of power will not cause rod movement.
If the trip insertion time above is not met, the rod shall be declared inoperable.
For Unit I Cycle 15, Group 1, Rod 8 and Group 2, Rod 5 may be considered operable with an insertion time 5 3.00 sec provided:
- 1) the average insertion time for the remaining rods in Groups I and 2 is 5 1.50 sec, and
- 2) the core average negative reactivity insertion rate is within the assumptions of the safety analysis.
Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The specified trip time is based upon the safety analysis in FSAR Chapter 15.
A rod is considered inoperable if the trip insertion time is greater than the specified allowable time or the core average negative reactivity insertion rate is less than the assumptions of the safety analysis.
REFERENCES (1)
FSAR, Section 15 (2)
Technical Specification 3.5.2 Oconee 1, 2, and 3 4.7-1 Amendment No? 2 8 (Unit 1)
Amendment No.229 (Unit 2)
Amendment No.225 (Unit 3)
4.8 MAIN STEAM STOP VALVES Applicability Applies to the main steam stop valves.
Objective To verify the ability of the main steam stop valves to close upon signal.
'Specification 4.8 Using Channels A and B, the operation of each of the main steam stop valves shall be tested every 18 months to demonstrate a closure time of one second or less in Channel A and a closure time of 15 seconds or less for Channel B.
Bases The main steam stop valves limit the Reactor Coolant System cooldown rate and resultant reactivity insertion following a main steam line break accident. Their, ability to promptly close upon redundant signals will be verified every 18 months. Channel A solenoid valves are designed to close all four turbine stop valves in 240 milliseconds. The backup Channel B solenoid valves are designed to close the turbine stop valves in approximately 12 seconds.
Using the maximum 15 second stop valve closing time, the fouled steam generator inventories and the minimum tripped rod worth with the maximum stuck rod worth, an analysis similar to that presented in FSAR Section 15.13, (but considering a blowdown of both steam generators) shows that the reactor will remain sub critical after reactor trip following a double-ended steam line break.
REFERENCES (1) FSAR, Section 10.3.4, and 15.13 Oconee 1, 2, and 3 4.8-1 Amendment No228 (Unit 1)
Amendment No229 (Unit 2)
Amendment No.2 2 5 (Unit 3)
4.9 EMERGENCY FEEDWATER PUMP AND VALVE PERIODIC TESTING Applicability Applies to the periodic testing of the turbine-driven and motor-driven emergency feedwater pumps and associated valves.
Objective To verify that the emergency feedwater pumps and associated valves are operable.
Specification 4.9.1 Pump Test The turbine-driven and motor-driven feedwater pumps shall be operated on recirculation to the upper surge tank for a minimum of one hour in accordance with the requirements of Specification 4.0.4.
4.9.2 Valve Test Automatic valves in the emergency feedwater flow path will be determined to be operable in accordance with the requirements of Specification 4.0.4.
4.9.3 System Flow Test Prior to Unit operation above 25% Full Power following any modifications or repairs to the emergency feedwater system which could degrade the flow path and at least once every 18 months, the emergency feedwater system shall be given either a manual or an automatic initiation signal.
4.9.4 Acceptance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly. In addition, during operation of the System Flow Test (Item 4.9.3 above), flow to the steam generators shall be verified by control room indication.
Bases The monthly testing frequency is sufficient to verify that the emergency feed-water pumps are operable.
Verification of correct operation is made both from the control room instrumentation and direct visual observation of the pumps. The parameters which are observed are detailed in the applicable edition of the ASME Boiler and Pressure Vessel Code,Section XI. The System Flow Test verifies correct total system operation following modifications or repairs.
REFERENCES (1)
FSAR, Section 10.4.7.4 Oconee 1, 2, and 3 4.9-1 Amendment No.228 (Unit 1)
Amendment No.229 (Unit 2)
Amendment No.225 (Unit 3)
4.12 CONTROL ROOM PRESSURIZATION AND FILTERING SYSTEM Aolicability Applies to control room pressurization and filtering system components Objective To verify that these systems and components will be able to perform their design functions.
Specification 4.12.1 Operating Tests
- a.
Control room outside air booster fan system tests shall be performed quarterly. These tests shall consist of an external visual inspection, a flow measurement for each unit and pressure drop measurements across each filter bank. Pressure drop across pre-filter shall not exceed I inch H20 and pressure drop across HEPA shall not exceed 2 inches H20. Fan motors shall be operated continuously for at least one hour, and all louvers shall be proven operable.
- b.
Every 18 months, verify the system maintains the control room at a positive pressure with both outside air booster fans on during system operation.
4.12.2 Filter Tests Every 18 months, for the Unit 1 and 2 and the Unit 3 control room an in-place leakage test using DOP on HEPA units and Freon-l 12 (or equivalent) on carbon units shall be performed at design flow on each filter train. Removal of 99.5 percent DOP by each entire HEPA filter unit and removal of 99.0 percent Freon-1 12 (or equivalent) by each entire carbon adsorber unit shall constitute acceptance performance. These tests must also be performed after any maintenance which may affect the structural integrity of either the filtration system units or of the housing.
Bases The purpose of the control room pressurization filtering system is to protect the control room operators from the effects of accidental release of radioactive effluents or toxic gases in the Turbine Building or Auxiliary Building only. The system is designed with two 50 percent capacity filter trains each of which consists of a prefilter, high efficiency particulate filters, carbon filters, booster fans, air handling unit fans, and associated ductwork to pressurize the control room with outside air.
Since these systems are not normally operated, a periodic test is required to insure their operability when needed. Quarterly testing of this system will show that the system is available.
Testing of the installed carbon adsorber stage and absolute filters every 18 months will verify the leak integrity of the cleanup system. Testing every 18 months will also verify the ability of the system to maintain the control room at a positive pressure to minimize infiltration of hazardous effluents.
Oconee 1, 2, and 3 4.12-1 Amendment No2 28 (Unit 1)
Amendment No? 2 9 (Unit 2)
Amendment No.22 5 (Unit 3)
4.14 REACTOR BUILDING PURGE FILTERS AND SPENT FUEL POOL VENTILATION SYSTEM Applicability Applies to testing of the Reactor Building purge filters for Units 2 and 3 and the spent fuel pool ventilation systems.
Objective To verify that the Unit 2 and Unit 3 Reactor Building purge filters will per-form their design function and that when used with the spent fuel pool ventilation system, will reduce the off-site dose due to a fuel handling accident.
Specification 4.14.1 Operational and Performance Testing
- a. Monthly, each train of the spent fuel pool ventilation system shall be operated through the respective Reactor Building purge filters for at least 15 minutes at design flow +/- 10%.
- b. Every 18 months, the spent fuel pool ventilation fans shall be shown to operate at design flow +/-
10% when tested in accordance with ANSI N510-1975.
- c. Leak tests using DOP or halogenated hydrocarbon, as appropriate shall be performed on the Reactor Building purge filters:
- 1. Every 18 months;
- 2. After each complete or partial replacement of HEPA filter bank or charcoal adsorber bank;
- 3. After any structural maintenance on the system housing;
- 4. After painting, fire, or chemical release in any ventilation zone communicating with the system.
- d. The results of the DOP and halogenated hydrocarbon tests on HEPA filters and charcoal adsorber banks shall show 299% DOP removal and 199% halogenated hydrocarbon removal, respectively, when tested in accordance with ANSI N510-1975.
Oconee 1, 2, and 3 4.14-1 Amendment No. 228(Unit 1)
Amendment No. 22(Unit 2)
Amendment No. 2 2 Unit 3)
- e. Every 18 months, or following 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation, or after painting, fire, or chemical release in any ventilation zone communicating with the system, a carbon sample shall be removed from the Reactor Building purge filters for laboratory analysis. Within 31 days of removal, this sample shall be verified to show k 90% radioactive methyl iodide removal when tested in accordance with ASTM D3803-1989 (300C, 95% R.H.). Otherwise, the filter system shall be declared inoperable.
Bases The Unit 2 Reactor Building purge filter is used in the ventilation system for the common spent fuel pool for Units 1 and 2. The Unit 3 Reactor Building purge filter is used in the Unit 3 spent fuel pool ventilation system. Each filter is constructed with a prefilter, an absolute filter and a charcoal filter in series. The high efficiency particulate air (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radioiodine.
Bypass leakage for the charcoal adsorbers and particulate removal efficiency for HEPA filters are determined by halogenated hydrocarbon and DOP respectively. The laboratory carbon sample test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the performances are as specified, the doses for a fuel handling accident would be minimized.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples. Each sample should be replaced. Any HEPA filters found defective should be replaced with filters qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52.
Operation of the spent fuel pool ventilation system every month will demonstrate operability of the fans, filters and adsorber system.
If painting, fire or chemical release occurs during system operation such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign materials, the same tests and sample analysis should be performed as required for operational use.
Oconee 1, 2, and 3 4.14-2 Amendment No2 28 (Unit 1)
Amendment No 29 (Unit 2)
Amendment No.22 5 (Unit 3)
418 SNUBBERS Applicability Applies to hydraulic and mechanical snubbers used to protect the Reactor Coolant System and other safety-related systems.
Objective To verify that the required hydraulic and mechanical snubbers are operable.
Specification 4.18.1 Each snubber associated with the Reactor Coolant System and other safety-related systems, as specified in the appropriate Station Procedure shall be visually inspected. Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) in those locations where mechanical snubber movement can be manually induced, the snubbers shall be inspected as follows:
(a)
Every 18 months, the inaccessible snubbers shall be inspected near the beginning and the end of an outage.
(b)
In the event of a severe dynamic event, snubbers in that system which experienced the event shall be inspected during the refueling outage to assure that the snubbers have freedom of movement and are not frozen up.
The inspection shall consist of verifying freedom of motion using one of the following: (i) Manually induced snubber movement, (ii) evaluation of in place snubber piston setting; (iii) stroking the mechanical snubber through its full range of travel. If one or more mechanical snubbers are found to be frozen up during this inspection, those snubbers shall be replaced (or overhauled) before returning to power. Re-inspection shall subsequently be performed according to the schedule listed below.
Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is dearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.18.4. However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be tested by starting with the piston at the as found setting Oconee 1, 2, and 3 4.18-1 Amendment No.228 (Unit 1)
Amendment No.2 29 (Unit 2)
Amendment No.2 2 5 (Unit 3)
and extending the piston rod in the tension mode direction. All snubbers connected toan inoperable common hydraulic fluid reservoir shall be counted as inoperable snubbers Snubber operability will be verified in accordance with the following schedule*
No. Inoperable Snubbers Subsequent Visual per Inspection Period Inspection Period 0
18 months* 25%
1 12 months +/- 25%
2 6 months
- 25%
3,4 4 months
- 25%
5,6,7 2 months
- 25%
8 1month
- 25%
Note: (1)
The required inspection interval shall not be lengthened more than two steps per inspection.
(2)
Snubbers may be categorized in two groups, "accessible" or "inaccessible,"
based on their accessibility during reactor operation. These two groups may be inspected indepenqently according to be above schedule.
(3)
Hydraulic and mechanical snubber Inspection schedules are independent 4.18.2 The seal service life of hydraulic snubbers shall be monitored to ensure that the seals do not exceed their expected service life by more than 10% between surveillance inspections.
The maximum expected service life for the various seals, seal materials, and applications shall be estimated based on engineering information, and the seals shall be replaced so that the maximum expected service life is not exceeded by more than 10% during a period when the snubber is required to be OPERABLE. The seal replacements shall be documented and the documentation shall be retained in accordance with Specification 6.5.1.m..
4.18.3 At least once every 18 months, a representative sample, a minimum of 10% of the total of hydraulic snubbers in use in the plant, shall be functionally tested either in place or in a bench test. For each hydraulic snubber that does not meet the functional test acceptance criteria of Specification 4.18.4, an additional minimum of 10% of the hydraulic snubbers shall be functionally tested until none are found inoperative or all have been functionally tested.
The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of hydraulic snubbers. The representative sample shall be selected randomly from the total population of safety-related hydraulic snubbers.
Oconee 1, 2, and 3 4.18-2 Amendment No228 (Unit 1)
Amendment No229 (Unit 2)
Amendment No.2 2 5 (Unit 3)
4.18.6 Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions. Snubbers so exempted shall be listed in a permanent record which references the exemption letter date.
Bases All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection.
Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nminal time less 25%) may not be used to lengthen the required inspection interval unless so determined, by the engineer, from a previous window of a schedule. Any inspection whose results require a shorter inspection interval will override the previous schedule.
When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.
When a snubber is found inoperable, an engineering evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperability of the snubber.
To provide assurance of snubber functional reliability, a representative sample of the installed hydraulic snubbers will be functionally tested every 18 months. Observed failures of these sample snubbers shall require functional testing of additional units.
Hydraulic snubbers and mechanical snubbers may each be treated as a differernt entity for the above surveillance programs.
Oconee 1, 2, and 3 4.18-4 Amendment No228 (Unit 1)
Amendment No.2 2 9 (Unit 2)
Amendment No.225 (Unit 3)
TABLE 4.20-1 SSF INSTRUMENTATION SURVEILLANCE REQUIREMENTS Check Calibrate Remarks
- 1.
RCS Pressure (3)
WE 18 months Loop A B
- 2.
SSF RC Makeup Pump (3)
Suction Pressure QU(1) 18 months Discharge Pressure QU(1) 18 months Suction Temperature QU(1) 18 months Discharge Flow QU(1) 18 months
- 3.
RC System Temperature (3)
NA(2) 18 months Loop A, B Hot, Cold
- 4.
Pressurizer Water Level (3)
WE 18 months
- 5.
SSF Auxiliary Service Water Pump Suction Pressure QU(1)
AN Discharge Pressure QU(1)
AN Unit 1 Discharge Pressure NA AN Unit 2 Discharge Pressure NA AN Unit 3 Discharge Pressure NA AN Discharge Test Flow QU(1)
AN Suction Temperature QU(1)
AN
- 6.
Steam Generator Levels (3)
WE 18 months A,B
- 7.
Underground Fuel Oil Storage NA AN Tank Inventory
- 8.
D/G Service Water Pump Discharge Flow QU(1)
AN Discharge Pressure QU(1)
AN
- 9.
D/G Air Start System WE AN Pressure (1)
Check when pump operated/tested per IST.
(2)
This instrumentation is normally aligned through a transfer/isolation device to each Unit Control Room and is thus checked in accordance with Specification 4.1, Table 4.1-1, Item 7. Every 18 months, the instrument string to the SSF Control Room will be checked and calibrated.
(3)
Units 1, 2,3.
Oconee 1, 2, and 3 4.20-5 Amendment No. 228(Unit 1)
Amendment No. 22 Unit 2)
Amendment No. 225(Unit 3)