ML16138A769

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Amends 203 & 200 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Revising LCOs & Surveillance Requirements Re Low Pressure Svc Water Sys
ML16138A769
Person / Time
Site: Oconee  
Issue date: 01/13/1994
From: Plisco L
Office of Nuclear Reactor Regulation
To:
Shared Package
ML16138A770 List:
References
NUDOCS 9401180065
Download: ML16138A769 (22)


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 203 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 May 3, 1993, as supplemented August 11, 1993, 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 3.8 of Facility Operating License No. DPR-38 is hereby amended to read as follows:

9401180065 940113 PDR ADOCK 05000269

-2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 203,.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 shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Loren R. Plisco, Acting Director Project Directorate 11-3 Division of Reactor Projects -

I/Il Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

January 13, 1994

0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 203 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 May 3, 1993, as supplemented August 11, 1993, 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 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.203, 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 shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Loren R. Plisco, Acting Director Project Directorate 11-3 Division of Reactor Projects -

I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

January 13, 1994

,NkRE 0

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 200 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 May 3, 1993, as supplemented August 11, 1993, 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 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 200, 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 shall be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION oren R. Plisco, Acting Director Project Directorate 11-3 Division of Reactor Projects -

I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance:

January 13, 1994

ATTACHMENT TO LICENSE AMENDMENT NO, 203 FACILITY OPERATING LICENSE NO. DPR-38 DOCKET NO. 50-269 AND TO LICENSE AMENDMENT NO.203 FACILITY OPERATING LICENSE NO. DPR-47 DOCKET NO. 50-270 AND TO LICENSE AMENDMENT NO.200 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 Paqes iv iv viii viii ix x

xi 3.3-2 3.3-2 3.3-3 3.3-3 3.3-5 3.3-5 3.3-6 3.3-6 3.3-7 3.3-7 4.5 4.5-12 4.5 4.5-9

Section Page 3.10 GAS STORAGE TANK AND EXPLOSIVE GAS MIXTURE 3.10-1 3.11 (Not Used) 3.11-1 3.12 REACTOR BUILDING POLAR CRANE AND AUXILIARY HOIST 3.12-1 3.13 SECONDARY SYSTEM ACTIVITY 3.13-1 3.14 SNUBBERS 3.14-1 3.15 CONTROL ROOM PRESSURIZATION AND FILTERING SYSTEM 3.15-1 AND PENETRATION ROOM VENTILATION SYSTEMS 3.16 HYDROGEN PURGE SYSTEM 3.16-1 3.17 (NOT USED) 3,18 STANDBY SHUTDOWN FACILITY 3.18-1 4

SURVEILLANCE REQUIREMENTS 4.0-1 4.0 SURVEILLANCE STANDARDS 4.0-1 4.1 OPERATIONAL SAFETY REVIEW 4.1-1 4.2 STRUCTURAL INTEGRITY OF ASME CODE CLASS 1, 2 4.2-1 AND 3 COMPONENTS 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4.3-1 4.4 REACTOR BUILDING 4.4-1 4.4.1 Containment Leakage Tests 4.4-1 4.4.2 Structural Integrity 4.4-14 4.4.3 Hydrogen Purge System 4.4-17 4.4.4 Reactor Building Purge System 4.4-20 4.5 EMERGENCY CORE COOLING SYSTEMS AND REACTOR 4.5-1 BUILDING COOLING SYSTEMS PERIODIC TESTING 4.5.1 Emergency Core Cooling Systems 4.5-1 4.5.2 Reactor Building Cooling Systems 4.5-4 4.5.3 Containment Heat Removal Capability 4.5-6 4.5.4 Penetration Room Ventilation System 4.5-7 4.5.5 Low Pressure Injection System Leakage 4.5-9 4.6 EMERGENCY POWER PERIODIC TESTING 4.6-1 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7-1 4.7.1 Control Rod Trip Insertion Time 4.7-1 4.7.2 Control Rod Program Verification 4.7-2 4.8 MAIN STEAM STOP VALVES 4.8-1 Oconee 1, 2, and 3 iv Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

INTRODUCTION These Technical Specifications apply to the Oconee Nuclear Station, Units 1, 2, and 3 and are in accordance with the requirements of 10CFR50, Section 50.36.

The bases, which provide technical support or reference the pertinent FSAR section for technical support of the individual specifications, are included for informational purposes and to clarify the intent of the specification. These bases are not part of the Technical Specifications, and they do not constitute limitations or requirements for the licensee.

The Technical Specifications while applying to Units 1, 2, and 3 are written on a single unit basis; exceptions to this are identified.

Oconee Units 1, 2, & 3 viii Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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.b(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 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 />.

c.

For all Units, when reactor power is greater than 60% FP:

(1)

In addition to the requirements of Specification 3.3.1.a(l) and 3.3.1.b(l) 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 operable. 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 />.

3.3.2 Low Pressure Injection (LPI) System

a.

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'F:

(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 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.

Oconee Units 1, 2, & 3 3.3-2 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

(0 3.3.3 Core Flood Tank (CFT) System When the RCS is in a condition with prqssure 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 boron concentration within the limit specified in the Core Operating Limits Report in each CFT; 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*F:

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 3.3.4.a 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 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 />.

b.

The BWST shall contain a minimum level of 46 feet of water having a minimum concentration of boron within the limit specified in the Core Operating Limits Report at a minimum temperature of 50*F. 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 (RBC) System

a.

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'F and subcritical:

(1)

Two independent RBC trains, each comprised of an RBC fan, associated cooling unit, and associated ESF valves shall be operable. Valve LPSW-108 shall be locked open.

(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 3.3.5.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 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 3.3-3 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

b.

When the reactor is critical:

(1)

In addition to the requirements of Specifications 3.3.6.a(l) 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 automaticallyupon 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 RBS 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(l) 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(l) 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 />.

3.3.7 Low Pressure Service Water (LPSW)

a.

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*F:

(1)

Three LPSW pumps for the shared Unit 1, 2 LPSW system shall be operable, except as provided in (2) below.

(2)

Two LPSW pumps for the shared Unit 1, 2 LPSW system shall be operable if Unit 1 or Unit 2 has been defueled and one LPSW pump is capable of mitigating the consequences of a design basis accident in the remaining Unit.

(3)

Two pumps for the Unit 3 LPSW system shall be operable.

b.

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 above within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 3.3.7.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 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 />.

Oconee Units 1, 2, & 3 3.3-5 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

AmendmentNo. 200 (Unit 3)

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 2,200'F 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 insufficient 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 requirement for a flowpath from LPI discharge to HPI pump suction is provided to assure availability of long term core cooling following a small break LOCA in which the BWST is depleted and RCS pressure remains above the shutoff head of the LPI pumps.

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)

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 requirements. Heaters maintain the borated water supply at a temperature above 50*F 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 Ak/k subcritical at 70*F without any control rods in the core. The minimum boron concentration is specified in the Core Operating Limits Report.

It has been shown for the worst design basis loss-of-coolant accident (a 14.1 ft*

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 operable for seven days or one Reactor Building spray system provided all three Reactor Building cooling units are operable. Valve LPSW-108 is the LPSW isolation valve on the discharge side of each Unit's RBCUs. This valve is required to be locked open in order to assure the LPSW flowpath for the RBCUs is available.

Three low pressure service water pumps serve Oconee Units 1 and 2 and two low Oconee Units 1, 2, & 3 3.3-6 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

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 Unit 1 and 2 LPSW system requires two pumps to meet the single failure criterion provided that one of the Units has been defueled and the following LPSW system loads on the defueled Unit are isolated:

RBCUs, Component Cooling, main turbine oil tank, RC pumps, and LPI coolers.

In this configuration, if two of the three LPSW pumps are inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are permitted by TS 3.3.7.b to restore two of the three LPSW pumps to operable status. At all other times when the RCS of Unit 1 or 2 is 2 350 psig or a 250F, all three LPSW pumps are required to meet the single failure criterion. When all three LPSW pumps are required to be operable and one of the three pumps is inoperable, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are permitted by TS 3.3.7.b to restore the pump to operable status.

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 3.3-7 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

System, verificat shall be made that the check a 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.

4.5.1.2 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.

During each refueling outage 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 testinga 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 1 and 2.

The LPI system test required by Specification 4.5.1.1.2 verifies that the LPI system responds as required to actuation of ES channels 3 and 4. In addition, this test verifies that the LPSW pumps and LPSW-4 and -5 (LPSW supply to LPI coolers) respond as required to actuation of ES channels 3 and 4. The test required by Specification 4.5.3 verifies the containment heat-removal capability of the LPI coolers (in conjunction with the RBCUs and RB Spray system).

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 Units 1, 2, & 3 4.5-2 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.

Testing the manual operability of power-operated valves in the Low Pressure Injection System gives assurance that flow can be established in a timely manner even if the capability to operate a valve from the control room is lost.

With the reactor shut down, the valves in each core flooding line are checked for operability by reducing the Reactor Coolant System Pressure until the indicated level in the core flood tanks verify that the check and isolation valves have opened.

Power Operated Valves LP-17 and LP-18, are boundary valves between high pressure and low pressure design piping. As such, functional testing of these valves is performed during cold shutdown conditions when the Reactor Coolant System pressure is below the design pressure of the Low Pressure Injection System piping and the potential for over-pressurization of the low pressure system is eliminated. Check Valves CF-12, CF-14, LP-47, and LP-48 are located on the high pressure piping and therefore can be leak tested with the Reactor Coolant System at hot shutdown conditions.

REFERENCE (1)

FSAR, Section 6 Oconee Units 1, 2, & 3 4.5-3 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (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)

During each refueling outage, 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.

During each refueling outage, 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.

Bases The Reactor Building Cooling System and Reactor Building Spray System are designed to remove heat in the containment atmosphere to control the rate of depressurization in the containment. The peak transient pressure in the containment is not affected by the two heat removal systems.

The delivery capability of one reactor building spray pump at a time can be tested Oconee Units 1, 2, & 3 4.5-4 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corresponding pump. Pump discharge pressure and flow indication demonstrate performance.

With the pumps shut down and the borated water storage tank outlet closed, the reactor building spray injection valves can each be opened and closed by operator action. With the reactor building spray inlet valves closed, low pressure air or fog can be blown through the test connections of the reactor building spray nozzles to demonstrate that the flow paths are open.

The RB Spray system test required by Specification 4.5.2.1.1 verifies that the RB Spray pumps and valves respond as required to actuation of ES channels 7 and 8. In addition, this test verifies that LP-21, and LP-22 (BWST supply to the RB Spray pumps) respond as required to actuation of ES channels 7 and 8. The test required by Specification 4.5.3 verifies the containment heat removal capability of the RB Spray system (in conjunction with the LPI coolers and RBCUs).

The equipment, piping, valves, and instrumentation of the Reactor Building Cooling System are arranged so that they can be visually inspected. The cooling units and associated piping are located outside the secondary concrete shield. Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment. The service water piping and valves out-side the Reactor Building are inspectable at all times.

The reactor building fans are normally operated periodically, constituting the test that these fans are operable.

The RBCU system test required by Specification 4.5.2.1.2 verifies that the RBCU fans respond as required to actuation of ES channels 5 and 6. In addition, this test verifies that LPSW-18 (LPSW for "A" RBCU), LPSW-21, LPSW-565, and LPSW-566 (LPSW for "B" RBCU), and LPSW-24 (LPSW for "C" RBCU) respond as required to actuation of ES channels 5 and 6. The LPI system test required by Specification 4.5.1.1.2 verifies that the LPSW pumps respond as required to actuation of ES channels 3 and 4. The test required by Specification 4.5.3 verifies the containment heat removal capability of the RBCUs (in conjunction with the LPI coolers and RB Spray system).

REFERENCE (1)

FSAR, Section 6 Oconee Units 1, 2, & 3 4.5-5 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

4.5.3 Containment Heat Removal Capability Applicability 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.

On a refueling frequency, 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 4.5.3.1.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 4.5.3.1.a requires that at a minimum the surveillance be performed on a refueling frequency. 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 Units 1, 2, & 3 4.5-6 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (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.

Specification 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.

During each refueling outage, 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 (+/- 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.

During each refueling outage;

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 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 299%

halogenated hydrocarbon removal, respectively, when tested in accordance with ANSI N510-1975.

Oconee Units 1, 2, & 3 4.5-7 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)

e.

During each refueling outage, 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 a90% radioactive methyl iodide removal when tested in accordance with ANSI N510-1975 (130*C, 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 Units 1, 2, & 3 4.5-8 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (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 During each refueling outage, 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 Units 1, 2, & 3 4.5-9 Amendment No. 203 (Unit 1)

Amendment No. 203 (Unit 2)

Amendment No. 200 (Unit 3)