ML16138A683
| ML16138A683 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/29/1989 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16138A684 | List: |
| References | |
| NUDOCS 8912070107 | |
| Download: ML16138A683 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 178 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 Power Company (the licensee) dated July 28, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and-regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; D. The issuance of this license 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:
89f2070107 89f129 Pr)R-.- ADOCK 0:)5000269 P
- m m
-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 178, 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.
FOR THE NUCLEAR PEGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of issuance: November 29, 1989
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 178 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 Power Company (the licensee) dated july 28, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license 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:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 178, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
-3.
This license amendmnenit is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: November 29, 1989
R'l RE G o7 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO.
50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Anierdment No. 175 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 Power Company (the licensee) dated July 28, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; D. The issuance of this license amendment will riot 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 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 175, 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.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Miatthews, Director Project Directorate 11-3 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: November 29, 1989
ATTACHMENT TO LICENSE AMENDMENT NO. 178 FACILITY OPERATING LICENSE NO. DPR-38 COCKET NO. 50-269 AND TO LICENSE AMENDMENT NO.
178 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO. 175 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 Pages Insert Pages 3.3-1 3.3-1 3.3-2 3.3-2 3.3-3 3.3-3 3.3-4 3.3-4 3.3-5 3.3-5 3.3-6 3.3-6 3.7-3 3.7-3 4.1-9 4.1-9 4.6-1 4.6-1 4.18-1 4.18-1 4.18-2 4.18-2
3.3 EMRGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR BUILDING
- SPRAY, AND LOW PRESSURE SERVICE WATER SYSTEMS Applicability Applies to the emergency core cooling, reactor building cooling, reactor building spray, and low pressure service water systems.
Objective To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building cooling, reactor building spray and low pressure service water systems.
Specification 3.3.1 High Pressure Injection (HPI) System
- a.
When the reactor coolant system (RCS), with fuel in the core, is in a condition with temperature above 350 0 F and reactor power less than 60% FP:
(1) Two independent trains, each comprised of an HPI pump and a flow path capable of taking suction from the borated water storage tank and discharging into the reactor coolant system automatically upon Engineered Safeguards Protective System (ESPS) actuation (HPI segment) shall be operable.
(2) Test or maintenance shall be allowed on any component of the HPI system provided one train of the HPI system is oper able.
If the HPI system is not restored to meet the require ments of Specification 3.3.1.a(1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the requirements of Specification 3.3.1.a(1) are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS temperature below 350'F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b.
For all Units, when reactor power is greater than 60% FP:
(1) In addition to the requirements of Specification 3.3.1.a(1) above, the remaining HPI pump and valves HP-409 and HP-410 shall be operable and valves HP-99 and HP-100 shall be open.
(2) Tests or maintenance shall be allowed on any component of the HPI system, provided two trains of HPI system are oper able.
If the inoperable component is not restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, reactor power shall be reduced below 60% FP within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
OCONEE -
UNITS 1, 2 & 3 3.3-1 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
3.3.2 Low Pressure Injection (LPI) System
- a.
When the RCS, with fuel in the core, is in a condition with pres sure equal to or greater than 350 psig or temperature equal to or greater than 250 0F:
(1) Two independent LPI trains, each comprised of an LPI pump and a flowpath capable of taking suction from the borated water storage tank and discharging into the RCS automatically upon ESPS actuation (LPI segment), together.with two LPI coolers and two reactor building emergency sump isolation valves (manual or remote-manual) shall be operable.
(2) Tests or maintenance shall be allowed on any component of the LPI system provided the redundant train of the LPI system is operable.
If the LPI system is not restored to meet the requirements of Specification 3.3.2.a(l) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the requirements of Specification 3.3.2.a(l) are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250'F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.3.3 Core Flood Tank (CFT) System When the RCS is in a condition with pressure above 800 psig both CFT's shall be operable with the electrically operated discharge valves open and breakers locked open and tagged; a minimum level of 13 +/-.44 feet (1040 +/- 30 ft. 3 ) and one level instrument channel per CFT; a minimum concentration of borated water in each CFT of 1835 ppm boron; and pressure at 600 +/- 25 psig with one pressure instrument channel per CFT.
3.3.4 Borated Water Storage Tank (BWST)
When the RCS, with fuel in the core, is in a condition with pressure equal to or greater than 350 psig or temperature equal to or greater than 250 0F:
- a.
The BWST shall have operable two level instrument channels.
(1) Tests or maintenance shall be allowed on one channel of BWST level instrumentation provided the other channel is operable.
(2) If the BWST level instrumentation is not restored to meet the requirements of Specification 3.3.4a above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the requirements of Specification 3.3.4.a are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 2501F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
OCONEE -
UNITS 1, 2 & 3 3.3-2 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
- b.
The BWST shall contain a minimum level of 46 feet of water having a minimum concentration of 1950 ppm boron at a minimum temperature of 50aF. The manual valve, LP-28, on the discharge line shall be locked open.
If these requirements are not met, the BWST shall be considered unavailable and action initiated in accordance with Specification 3.2.
3.3.5 Reactor Building Cooling (REC) System
- a.
When the RCS, with fuel in the core, is in a condition with pres sure equal to or greater than 350 psig or temperature equal to or greater than 250'F and subcritical:
(1) Two independent RBC trains, each comprised of an RBC fan, associated cooling unit, and associated ESF valves shall be operable.
(2) Tests or maintenance shall be allowed on any component of the RBC system provided one train of the RBC and one train of the RBS are operable. If the RBC system is not restored to meet the requirements of Specification 3.3.5a(l) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250oF within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b.
When the reactor is critical:
(1) In addition to the requirements of Specifications 3.3.5.a(1) above, the remaining RBC fan, associated cooling unit, and associated ESF valves shall be operable.
(2) Tests or maintenance shall be allowed on one RBC train under either of the following conditions:
(a) One REC train may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(b) One RBC train may be out of service for 7 days provided both RBS trains are operable.
(c) If the inoperable RBC train is not restored to meet the requirements of Specification 3.3.5.b(l) within the time permitted by Specification 3.3.5.b(2) (a) or (b), the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the requirements of Specification 3.3.5.b(1) are not met within an addi tional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250 0 F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
OCONEE -
UNITS 1, 2 & 3 3.3-3 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
.-3.6 Reactor Building Spray (RBS) System
- a.
When the RCS, with fuel in the core, is in a condition with pres sure equal to or greater than 350 psig or temperature equal to or greater than 2501F and subcritical:
(1) One RBS train comprised of an RBS pump and a flowpath capable of taking suction from the LPI system and discharging through the spray nozzle header automatically upon ESPS actuation (RBS segment) shall be operable.
(2) Tests or maintenance shall be allowed on any component of the RBS system under the following conditions:
(a) One RBS train may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided two RBC trains are operable.
(b) If the inoperable RBS train is not restored to meet the requirements of Specification 3.3.6.a(1) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a condition with the RCS pressure below 350 psig and RCS temperature below 250 0 F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- b.
When the reactor is critical:
(1) In addition to the requirements of Specifications 3.3.6.a(1) above, the other RBS train comprised of an RBS pump and a flowpath capable of taking suction of the LPI system and discharging through the spray nozzle header automatically upon ESPS actuation (RBS segment) shall be operable.
(2) Tests or maintenance shall be allowed on one RBS train under either of the following conditions:
(a) One RBS train may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(b) One RES train may be out of service for 7 days provided all three RBC trains are operable.
(c) If the inoperable RBS train is not restored to meet the requirements of Specification 3.3.6.b(1) above within the time permitted by Specification 3.3.6.b(2) (a) or (b),
the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the requirements of Specification 3.3.6.b(1) are not met within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in a condition with RCS pressure below 350 psig and RCS temperature below 250'F within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
OCONEE -
UNITS 1, 2 & 3 3.3-4 Amendment No.178 (Unit 1)
Amendment No.178 (Unit 2)
Amendment No.175 (Unit 3)
3.3.7 Low Pressure Service Water (LPSW)
- a.
When the RCS, with fuel in the core, is in a condition with pres sure equal to or greater than 350 psig or temperature equal to or greater than 2501F:
(1) Two LPSW pumps for the shared Unit 1, 2 LPSW system and two LPSW pumps for the Unit 3 LPSW system shall be operable with valves LPSW-108, 2LPSW-108, and 3LPSW-108 locked open.
(2) Tests or maintenance shall be allowed on any component of the LPSW system provided the redundant train of the LPSW system is operable.
If the LPSW system is not restored to meet the requirements of Specification 3.3.7.a(l) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the requirements of Specifica tion 3.3.7.a(1) are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following hot shutdown, the reactor shall be placed in condition with RCS pressure below 350 psig and RCS temperature below 2500 within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Bases Specification 3.3 assures that, for whatever condition the reactor coolant system is in, adequate engineered safety feature equipment is operable.
For operation up to 60% FP, two high pressure injection pumps are specified.
Also, two low pressure injection pumps and both core flood tanks are required.
In the event that the need for emergency core cooling should occur, functioning of one high pressure injection pump, one low pressure injection pump, and both core flood tanks will protect the core, and in the event of a main coolant loop severance, limit the peak clad temperature to less than 2,2001F and the metal water reaction to that representing less than 1 percent of the clad. (1)
Both core flooding tanks are required as a single core flood tank has insuffi cient inventory to reflood the core.
The requirement to have three HPI pumps and two HPI flowpaths operable during power operation above 60% FP is based on considerations of potential small breaks at the reactor coolant pump discharge piping for which two HPI trains (two pumps and two flow paths) are required to assure adequate core cooling.(2)
The analysis of these breaks indicates that for operation at or below 60% FP only a single train of the HPI system is needed to provide the necessary core cooling.
The borated water storage tanks are used for two purposes:
(a) As a supply of borated water for accident conditions.
(b) As a supply of borated water for flooding the fuel transfer canal during refueling operation.(3)
OCONEE -
UNITS 1, 2 & 3 3.3-5 Amendment No.178 (Unit 1)
Amendment No.178 (Unit 2)
Amendment No.175 (Unit 3)
Three-hundred and fifty thousand (350,000) gallons of borated water ( a level of 46 feet in the BWST) are required to supply emergency core cooling and reactor building spray in the event of a loss-of-core cooling accident. This amount fulfills requirements for emergency core cooling. The borated water storage tank capacity of 388,000 gallons is based on refueling.volume require ments. Heaters maintain the borated water supply at a temperature above 501F to lessen the potential for thermal shock of the reactor vessel during high pressure injection system operation. The boron concentration is set at the amount of boron required to maintain the core 1 percent subcritical at 700F without any control rods in the core.
The minimum value specified in the tanks is 1950 ppm boron.
It has been shown for the worst design basis loss-of-coolant accident (a 14.1 ft2 hot leg break) that the Reactor Building design pressure will not be exceeded with one spray and two coolers operable.
(4) Therefore, a maintenance period of seven days is acceptable for one Reactor Building cooling fan and its associated cooling unit provided two Reactor Building spray systems are oper able for seven days or one Reactor Building spray system provided all three Reactor Building cooling units are operable.
Three low pressure service water pumps serve Oconee Units 1 and 2 and two low pressure service water pumps serve Oconee Unit 3. There is a manual cross connection on the supply headers for Unit 1, 2, and 3. One low pressure service water pump per unit is required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of-coolant accident.
The operability of redundant equipment(s) is determined based on the results of inservice inspection and testing as required by Technical Specification 4.5 and ASME Section XI.
REFERENCES (1) ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Babcock &
Wilcox, Lynchburg, Virginia, June 1975.
(2) Duke Power Company to NRC letter, July 14, 1978, "Proposed Modifications of High Pressure Injection System".
(3) FSAR, Section 9.3.3.2 (4) FSAR, Section 15.14.5 OCONEE -
UNITS 1, 2 & 3 3.3-6 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
The circuits or channels of more than one functional unit of the EPSL may be inoperable only if:
- 1.
The inoperability results from a loss of power due to the in operability of a 125 VDC instrumentation and control panelboard (see 3.7.2(e) below); and
- 2.
The conditions of Table 3.7-1 for degraded operation are satis fied for the affected functional units.
If any event, if the reactor is subcritical, the inoperable cir cuit(s) or channel(s) shall be restored to operability and the conditions of Table 3.7-1 for normal operation shall be satisifed for all functional units before the reactor is returned to criticality.
(c) One 4160 volt main feeder bus may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(d) One complete single string (i.e., 4160 volt switchgear (TC, TD, or TE), 600 volt load center, (X8, X9, or X10), 600-208 volt MCC (XS1, XS2, or XS3), and their loads) of each unit's 4160 volt Engineered Safety Features Power System may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(e) One or more of the following DC distribution components may be in operable for periods not exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (except as noted in 3.7.2(g) below):
- 1.
One complete single string or single component (i.e., 125VDC battery, charger, distribution center, and panelboards) of the 125VDC 230KV Switching Station Power System.
- 2.
One complete single string or single component (i.e., 125VDC battery, charger, and distribution center) of the Keowee 125VDC Power System may be inoperable provided the remaining string of Keowee is operable and electrically connected to an operable Keowee hydro unit.
- 3.
One complete single string or single component (i.e., 125VDC battery, charger, distribution center, and associated isolating and transfer diodes) of any units 125VDC Instrumentation and Control Power System. Only one battery more than the number allowed to be inoperable per 3.7.1 (f) for the Station may be removed from service under this paragraph.
- 4.
One 125 VDC instrumentation and control panelboard and its associated loads, per unit, provided that no additional AC buses are made inoperable beyond the provisions of 3.7.2(a), (c), and (d), and provided that the conditions of Table 3.7-1 for normal operation are satisfied for all functional units of the EPSL before the 125 VDC instrumentation and control panelboard becomes inoperable. Additionally, the provisions of 3.7.2.(h) must be observed for the 120 VAC vital instrumentation power panelboard which is powered by the affected 125 VDC panelboard.
OCONEE -
UNITS 1, 2 & 3 3.7-3 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
Table 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency
- 1.
Control Rod Movement'(1)
Movement of Each Rod Monthly
- 2.
Pressurizer Safety Valves Setpoint Each Refueling
- 3.
Main Steam Safety Valves Setpoint Each Refueling
- 4.
Refueling System Interlocks(5)
Functional Prior to Refueling
- 5.
Main Steam Stop Valves (1)
Movement of Each Stop Monthly Valve
- 6.
Evaluate Daily Leakage
- 7.
Condenser Cooling Water Functional Each Refueling System Gravity Flow Test
- 8. -High Pressure Service Functional Monthly Water Pumps and Power Supplies
- 9.
Spent Fuel Cooling System Functional Prior to Refueling
- 10.
igh Pressure and Low st Vent Pump Casings Monthly and Prior Pressure Injection System to Testing
- 11.
Emergency Feedwater Functional Each Refueling Pump Automatic Start and Automatic Valve Actuation Feature (1)
Applicable only when the reactor is critical.
(2) Applicable only when the reactor coolant is above 200OF and at a steady state temperature and pressure.
Operating pumps excluded.
Number of safety valves to be tested each refueling shall be in accordance with ASME CodesSection XI, Article IWV-3511, such that each valve is tested at least once every 3 years.
Applicable only to the interlocks associated with the Reactor Building Purge System.
OCONEE -
UNITS 1, 2 & 3 4.1-9 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
4.16 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 equip ment. This test shall assure that:
- a.
Each hydro unit can be automatically started from the Unit 1 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 emer gency 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 prac tical rate.
4.6.3 Monthly, the Keowee Underground Feeder Breaker Interlock shall be verified to be operable.
4.6.4 During each refueling outage, a simulated emergency transfer of the 4160 volt main feeder buses to the startup transformer (i.e., CT1, CT2 or CT3) and to the 4160 volt standby buses shall be made to verify proper operation.
4.6.5 Quarterly, the External Grid Trouble Protection System logic shall be tested to demonstrate its ability to provide an isolated power path between Keowee and Oconee.
4.6.6 Annually and prior to planned extended Keowee outages, it shall be demonstrated that a Lee Station combustion turbine can be started and OCONEE -
UNITS 1, 2 & 3 4.6-1 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
S 4.18 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) At each refueling, the inaccessible snubbers shall be inspected near the beginning and the end of the 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 clearly 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 OCONEE -
UNITS 1, 2 & 3 4.18-1 Amendment No. 178 (Unit 1 Amendment No. 178 (Unit 2)
Amendment No. 175 (Unit 3)
Sfound setting and extending the piston rod in the tension mode direction. All snubbers connected to an 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 1 month +/- 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 accessiblity during reactor operation. These two groups may be inspected independently according to the 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 re quired to be OPERABLE.
The seal replacements shall be documented and the documentation shall be retained in accordance with Specifi cation 6.5.1.m.
4.18.3 At least once per refueling outage, 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 -
UNITS 1, 2 & 3 4.18-2 Amendment No. 178 (Unit 1)
Amendment No. 178 (Unit 2)
Amentment No. 175 (Unit 3)