ML20058E256
| ML20058E256 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 11/05/1993 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058E244 | List: |
| References | |
| NUDOCS 9312060225 | |
| Download: ML20058E256 (8) | |
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UNITED STAYES j
NUCLEAR REGULATORY COMMISSION
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MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.143 l
License No. DPR-36 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
i A.
The application for amendment filed by the Maine Yankee Atomic Power f
Company (the licensee) dated June 7, 1993, as supplemented October 1, 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 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 i
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; 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 amended by changes to~ the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.B.(6)(b) of Facility Operating License No. DPR-36 is hereby amended to read as follows:
9312060225 931105 PDR ADOCK 05CK)O309 p
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Jechnical specifications The Technical Specifications contained in Appendix A, as revised through. Amendment No.143, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of its date of issuance and i
shall be inplemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION 1,
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1 Walter R. Butler, Director Project Directorate I-3 i
Division of Reactor Projects - 1/II l
Office of Nuclear Reacter Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 5, 1993
6 O'
ATTACHMENT TO LICENSE AMENDMENT NO.143 FACILITY OPERATING LICENSE NO. DPR-36 DOCKET NO. 50-309 Replace the following pages_of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and i
contain vertical lines indicating the areas of change.
Remove Insert 4.6-1 4.6-1 4.6-3 4.6-3 l
4.6-4 4.6-4 4.6-5 4.6-5 4.6-6 4.6-6 l
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J 4.6 PERIODIC TESTING SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS STEAM GENERATOR EMERGENCY FEED PUMPS MAIN STEAM EXCESS FLOW CHECK VALVES FEEDWATER TRIP SYSTEM REACTOR COOLANT SYSTEM EMERGENCY VENT SYSTEM i
Apolicability Applies to the safety injection system, the containment spray system, chemical injection system, the containment cooling system, the emergency feedwater system, the main steam excess flow check valves, the feedwater trip system, and the reactor coolant system emergency v?nt system.
Ob.iective To verify that the subject systems will respond promptly and perform their intended functions, if required.
Specification A.
SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS l.
The following tests will be performed whenever plant conditions are as defined in Section 3.6. A of these Specifications.
a.
Emergency Core Cooling System (ECCS) Pumps:
Both operable high pressure safety injection (HPSI) pumps shall be tested quarterly in accordance with Specification 4.7 by operating i
in the charging mode.
Both operable low pressure safety injection (!'SI) pumps and both operable containment spray (CS) pumps shall be tested quarterly in accordance with Specification 4.7 by operating in the recirculation mode.
b.
ECCS Valves:
i All automatically operated valves that are required to operate to
)
assure core flooding, or containment spray shall be exercised quarterly in accordance with Specification 4.7.
As part of this quarterly testing, the volume centrol tank (VCT) outlet to charging pump suction valves need only be exercised through part travel, j
Monthly all other valves required for safeguards operation in i
accordance with Specification 3.6.A.2 shall be visually checked to i
verify proper operating position.
Amendment No. 45, 58, 62, 87, MO, H4, B7,143 4.6-1 w. w....
O' f.
l The check. valve barriers defined in Technical Specification 3.19.A.4 shall be deterrained to be intact by leak testing.
Check valve barriers shall be determined to be intact if each barrier (a) leakage is less than 15 gpm and each barrier (b) leakage is less than 5 gpm.
g.
Containment isolation Valves:
Each automatic containment isolation valve that cannot be tested every three months during reactor operation in accordance with Specification 4.6.A.4 shall be tested in accordance with Specification 4.7.
3.
The containment spray flow nozzles shall be tested every ten years in accordance with Specification 4.7.
The test shall consist of pressurizing the headers vith air and verifying that the nozzles are free of obstruction.
4.
Containment Isolation Valves:
Where practicable, each containment isolation valve shall be stroked to the position required to fulfill its safety function every three months in accordance with the Specification 4.7.
B.
EMERGENCY FEEDWATER PUMPS Prior %
lant startup following an extended cold shutdown, a flow test will be perfm to verify the normal flow path from the demineralized water storage tank to the steam generators.
The flow test shall be conducted with the l
emergency feedwater system valves in their normal alignment.
1 When the reactor is in a power operation condition (Condition 7), monthly inspections shall be performed to verify that all manual valves in the emergency and auxiliary feedwater systems necessary to assure emergency and auxiliary feedwater flow from the primary water source to the steam generators are locked in the proper position.
When the reactor is in a power operation condition (liary feedwater pump andConditio driven emergency feed pump, the turbine driven auxi system valves shall be tested quarterly in accordance with Specification 4.7 to demonstrate operability.
Bistable actuation setpoints of the motor driven emergency feed pumps shall be tested monthly in accordance with Table 4.1-2, number 21a.
During each refueling shutdown, a verification test shall be conducted to assure 1
that each motor driven emergency feed pump auto-start circuitry actuates upon i
receipt of an emergency feedwater actuation test signal in accordance with Table 4.1-2, number 21b.
C.
MAIN STEAM EXCESS FLOW CHECK VALVES The main steam excess flow check valves shall be tested once every 3 months for movement of the valve disc through a distance of approximately one and one-half inches. These valves shall be tested through full travel distance during each l
i refueling interval.
1 Amendment No. E, 65, 68, 87, H7,143 4.6-3 I
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D.
FEEDWATER TRIP SYSTEM 1.
The followinetests shall be performed at each refueling interval:
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a.
Main Feedwater Pumps Each main feedwater pump, condensate pump, and heater drain pump trip system shall be tested by tripping the actuation circuitry with a safety injection signal coincident with steam generator low pressure signal.
l b.
teedwater Valves Each main feedwater regulating valve, main feedwater regulating i
bypass valve, and emergency feenwater control and isolation valve trip system shall be tested by tripping the valves with a low pressure signal from their respective steam generators.
E.
REACTOR COOLANT SYSTEM EMERGENCY VENT SYSTEM 4
At least once each refueling interval during cold shutdown condition:
1.
All remotely operated valves shall be cycled 2.
Flow through each vent path shall be verified B.iLLLL The safety injection system and the containment spray system are principal plant safeguards systems that are normally operable during reactor operation.
Complete system tests cannot be performed when the reactor is operating because of their inter-relation with operating systems. The method of assuring operability of these systems is a combination of complete system tests performed during refueling shutdowns and inservice testing of active systein components (pumps and valves) which can be performed during reactor operation.
The test interval is based on the judgement that more frequent testing would not significantly increase the reliability (i.e., the probability that the component would operate when required), yet more frequent tests would result in increased wear over a long period of time.
Valve performance is monitored as part of the In-Service lesting Program, which encompassesSection XI of the ASME Code.Section XI of the ASME Code provides the activities and frequencies ne:.assary to satisfy the surveillance requirements.
The part travel exercising of the VCT outlet to charging pump suction valves, in lieu of the full travel acrcise, is conducted to preclude an interruption of normal plant operations.
Redundant valves have been used to assure proper lineup in the event of ECCS actuation.
Amendment 45, 55, 58, 62, 76, 87, MO, 4M,143 Correction Letter of 08/18/87 4.6-4 u\\sMono
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Other ECCS valves whose operation is not required to assure core flooding or l
containment spray shall be tested during each refueling shutdown period in accordance with Specification A.2.b.
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The three check valves in the ECCS line to each loop provide assurance that 3
a valve failure will not result in unrestricted flow of pressurized reactor coolant into lower pressure connecting piping outside the containment. The valve integrity testing required by Technical Specification 4.6.A.2.f l
assures that the rate of flow under a valve failure condition will not exceed the pressure relief capacity of the line.
It further provides periodic assurance that the check valves are intact.
The two check valves closest to the loop are grouped together as a single check valve barrier for test purposes.
The first valve provides a thermal barrier preventing thermal distortion from affecting the tightness of the second valve. The third valve alone constitutes a check valve barrier.
The check valves are hard seated swing checks designed to withstand the rigors of long term RHR operation without damage and the greatest assurance of integrity and dependability.
The leakage criteria provide an acceptable balance between the need to maintain a degree of tightness as a criterion of integrity and ALARA on one hand and power dependability considerations on the other.
This criteria gives due credit to the design of, and protection provided by, the three check valves in series.
Verification that the spray piping and nozzles are open shall be made initially by a suitably sensitive method, and at least every ten years thereafter.
Since cll piping material is all stainless steel, normally in a dry condition and with no plugging mechanism available, the retest every ten years is considered to be more than adequate.
l.
Other systems that are important to the emergency cooling function are the SI tanks, the component cooling system and the service water system.
The SI tanks are a passive safety feature.
In accordance with Specification 4.1 (Table 4.1-2, Item 11), the water volume and pressure in the SI tanks are checked periodically. The component cooling and service water systems operate when the reactor is in operation and are continuously monitored for satisfactory performance.
The quarterly test of the steam generator motor driven emergency feed pumps and l
the turbine driven auxiliary feedwater pump verifies their operability by recirculcting water to the demineralized water tank.
The bistable actuation setpoints for the motor driven emergency feed pumps are checked monthly in accordance with f21 of Table 4.1-2 to verify that the setpoints have not drifted.
Prior to plant startup following an extended cold shutdown, a flow test 1s performed on the emergency feedwater system to functionally verify the system l
alignment from the demineralized water storage tank to the steam generators.
Amendment No. 60, M, MG, +B+, 4W,143 4.6-5 L: \\shmge w
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Monthly inspections are performed to verify that all manual valves in the p"egency and auxiliary feedwater systems from the primary water source to the l
im generators are locked in the proper position.
Proper functioning of the steam turbine admission valve and starting of the auxiliary feed pump will demonstrate the operability of the steam driven pump.
Verification of correct operation shall be made both from j
instrumentation in the Main Control Room and direct visual observation of the pumps.
The main steam excess flow check valves serve to limit an excessive reactor coolant system cooldown rate and resultant reactivity insertion following a main steam break incident. Their freedom to move shall be verified l
periodically.
The feedwater trip system acts to limit excessive reactor coolant system cooldown and the resultant reactivity insertion produced by excessive feedwater flow to the steam generators in the event of a main steam line break. The system acts to trip feedwater pumps, condensate pumps, and he.
drain pumps, and close the main feedwater regulating valve, feedwater regulating valve bypass valve, and emergency feedwater control and isolation valves to the affected steam generator.
Signals activating the system are developed by instrumentation, logic, and relaying associated with the safety injection actuation system and the excess flow check valve actuation system.
The circuitry which develops these signals is subject to surveillance requirements of Tables 4.1-1 and 4.1-2 which assure their reliability.
The main feedwater pumps, condensate pumps, and heater drain pumps trip upon coincidence of SIAS and a low steam generator pressure.
The valves close on the low steam generator pressure in the associated steam generator.
The reliability of the coincidence logic is assured by testing in accordance with
- 20 of Table 4.1-2.
Cycling of the reactor coolant system emergency vent system valves and verification of flow through each path documents the operability of. the system.
Operability of this system assures the ability to remove noncondensible gases from the reactor coolant system under severe post accident conditions in order to help prevent or mitigate core damage.
4 Amendment No. 62, 76, 87, MO,143 4.6-6 Li\\sw\\eno
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