ML20234B169
| ML20234B169 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 06/25/1987 |
| From: | Nerses V Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20234B144 | List: |
| References | |
| GL-83-37, NUDOCS 8707020256 | |
| Download: ML20234B169 (10) | |
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o UNITED STATES
'g . NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555
\...+/ MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 MAINE YANKEE f.TOMIC POWER STATION AMENDMENT TO FACILITY OPERATING L CENSE Amendment No.100 License No. DPR ~16
- 1. The Nuclear Regulatory Commission (the Comission) has found that:
A. The application for amendment by Maine Yankee Atomic Power Company (the licensee), dated March 13, 1985 as clarified by letters dated .lanuary 15, 1986 and January 13, 1987, complies with the standards and reouirements 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 ooerate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (1) 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; D. The issuance of this amendment will not be inimical to the comon 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.
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- 2. Accordingly, the license is amended by changes to the Technical Specifications 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:
l (b) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is ef4ctive 60 days from the date of this letter.
FOR T E NUCLEAR REGULATORY COMMISSION D
Victor Nerses, Acting Director Project Directorate I-3 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: June 25, 1987 k
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4 ATTACHMENT TO LICENSE AMENDMENT NO. 100 FACILITY OPERATING LICENSE NO. DPR-36 DOCKET NO. 50-309 Revise Appendix A as follows:
Remove Pages Insert Pages 3.3-2 3.3-2 3.3-3 3.3-3(unchanged. Repositioned)'
3.3-4* '
4.6-1 4.6-1 4.6-4 4.6-4 4.6-5 4.6-5 (unchanged. Repositioned) 4.6-6 ".6-6 l
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- New page >
4 C. Pressurizer
- 1. .The pressurizer shall be operable with at least one bank of proportional heaters and a water icvel between 28 and 60 percent during normal system operation whenever the reactor coolant system T avg is greater than 500 F.
- 2. The pressurizer spray system must be lined up to provide continuous pressurizer spray flow whenever the reactor is critical.
D. Reactor Coolant System Emergency Vent System
- 1. At least one reactor coolant system emergency vent path, consis-ting of at least two valves in series capable of being powered from emergency busses shall be operable and closed during trans-thermal and higher operating conditions (Conditions 4-7), at each of the following locations:
reactor head
- pressurizer
- a. Power shall be available at emergency busses MCC 7B/8B in order to satisfy the operability criteria for valve actuators PR-M-89/90 and RC-M-54/56,
- b. This Specification also requires that power be removed from these valve actuators. This does not affect valve. opera-bility.
Note: Power shall be provided to the valve actuators when the system is selected for use during post-accident conditions.
Remedial Action:
- 1. With one RCS emergency vent path inoperable, ensure that a PORV is operable providing a redundant vent path, and provide a report to the Commission within 30 days describing the cause and nature of the problem and schedule for repair of the inoperable emergency vent path.
- 2. With both of the RCS emergency vent paths inoperable, ensure that one PORV venting path is operable, and restore at least one RCS emergency vent path to operable within 30 days, or be in cold shutdown condition within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Note: A PORV is operable as a vent path if both the PORV and its respective block valve may be remotely operated.
3.3-2 Amendment No. EZ,Ef,EE,67,EE, 100
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Basis:
Reactor coolant pump flow and steam generator heet transfer capabilities ,
are specified to assure adequate core heat transfer capability under all operating conditions from criticality to full power. Three loop operation is specified to assure plant operation is restricted to conditions considered in the safety analyses.
The exception permits testing to detemine decay heat removal capabilities of the primary system while on natural circulation, prior to operation at
' higher power.
Following a loss of offsite power, stored and decay heat from the reactor would normally be removed by natural circulation using the steam generators as the heat sink. Water supply to the ' steam generators is maintained by the auxiliary feedwater system. Natural circulation cooling of the primary system requires the use of the pressurizer heaters or high pressure safety injection pumps to maintain a suitable overpressure on the reactor coolant system. Alternatively, in the event that natural circulation in the reactor coolant system is interrupted, the feed and bleed mode of reactor coolant system operation can be used to remove decay heat from the reactor. This method of decay heat removal requires the use of the emergency core cooling system (ECCS) and the power-operated relief valves (PORVs) in the pressurizer.
The p0RVs can be operated either manually or automatically in the Maine Yankee design. Block valves are provided upstream of the relief valves to isolate the valve in the event that a PORV fails.
The exception pemits hydrostatic testing of the Reactor Coolant System in accordance with the ASME code when the test pressure approaches the PORV setpoint.
When reactor coolant boron concentration is being reduced, the process must be uniform throughout the reactor coolant system volume to prevent stratifica-tion of reactor coolant at a lower boron concentration which could result in a reactivity insertion.
Sufficient mixing of the reactor coolant is assured by one low pressure safety injection (LPSI) pump operating in the RHR mode. When operated in this mode it will circulate the reactor coolant system volume in less than 12 minutes. The pressurizer volume is relatively inactive; therefore, it will tend to have a boron concentration higher than the rest of the reactor coolant system during a dilution operation. A continuous pressurizer spray l flow will maintain a nominal spread between the boron concentration in the I pressurizer and the reactor coolant system during the addition of boron.
Without residual heat removal, the amount of steam which could be generated at safety valve lift pressure with the reactor subcritical would be less than one valve's capacity. One valve, therefore, provides adequate defense against overpressurization when the reactor is subcritical and the RHR l system is isolated.
l ll 3.3-3 Amendment No, M,67,PJ,100
____________- ___ _ _-____________-______ a
Overpressure protection is provided for all critical conditions. The safety valves are sized to relieve steam at a rate equivalent to the peak volumetric pressure surge rate. For this purpose one safety valve is sufficient; ;
however, a minimum of two safety valves is required by Section III of the ASME Code.
Reactor coolant system emergency vent paths provide a means of removing non-condensible gasses from the reactor coolant system following an extremely unlikely severe nuclear accident. The requirement that the RCS vent system be closed with power removed from the actuators at or above operating Condition' 4 removes the possibility of inadvertent opening while at power.
The power operated relief valves provide an alternate means of removing-non-condensible gasses should the reactor coolant system emergency vent
' system become inoperable.
3.3-4 Amendment No. 100
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. 4.6 PERIODIC TESTING SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS STEAM GENERATOR EMERGENCY FEED PUMPS MAIN STEAM EXCESS FLOW CHEW VALVES FEEDWATER TRIP SYSTEM l REACTOR COOLANT SYSTEM EMERGENCY VENT SYSTEM i l 1
Applicability:
Applies to the safety injection system, the containment spray system, q chemical injection system, the containment cooling system, the emergency y feedwater system, the main steam excess flow check valves, the feedwater !
trip system, and the reactor coolant system emergency vent system.
Objective-l To verify that the subject systems will respond promptly and perform their intended functions, if required.
Specification:
A. SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS
- 1. The following tests will be performed monthly whenever plant conditions are as defined in Section 3.6. A of these Specifications.
- a. Emergency Core Cooling System (ECCS) Pumps:
1 Both operable high pressure safety injection (HPSI) pumps shall be l tested by operating in the charging mode. 1 Both operable low pressure safety injection (LPSI) pumps and both
- operable containment spray (CS) pumps shall be tested by operating in the recirculation mode.
Acceptable performance shall be that pumps attain rated heads, operate for at'least 15 minutes, and that the associated instru-mentation and controls function properly,
- b. ECCS Valves:
All automatically operated valves that are required to operate to assure core flooding, or containment spray shall be exercised.
The volume control tank (VCT) outlet to charging pump suction valves shall be exercised through part travel and all other valves shall be visually checked to verify proper operating position.
Exception: LSI-M-ll, 21 or 31 shall not be tested when the associated ECCS check valve barrier leakage falls into Condition 2 or 3, as defined in Specifica-tion 4.6.a.2.f.
The following tests will be performed at each refueling interval: '
2.
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- a. ECCS Pumps:
One HPSI pump shall be flow tested at 1000 psig discharge head.
4.6-1 Amendment No. O,EE,62,E7,100
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 (Condition 7), each motor drivem emergency feed pump and the turbine driven auxiliary feedwater pump and systems valves shall be tested at monthly' intervals to demonstrate operability. Bistable actuation setpoints of the motor driven emergency 1 feed pumps shall be tested monthly in accordance with Table 4.1-2, number 21a.
C. MAIN STEAM EXCESS FLOW CHECK VALVES )
The main steam excess flow check valves shall be tested once every 6 weeks for movement of the valve disc through a distance of approximately one and one-half inches. These valves will be tested through full travel distance during each refueling interval.
D. FEEDWATER TRIP SYSTEM
- 1. The following tests will be performed at each refueling interval:
- 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.
- b. Feedwater Valves
- Each main feedwater regulating valve, main feedwater regulating bypass valve, and emergency feedwater control and isolation i 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 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.
Basis:
The safety injection system and the containment spray system are principal plant safeguards systems that are nonna11y operable during reactor opera-tion.
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 4.6-4 Amendment No. 45,5E,fE,57,76,E7,100
I tests performed during refueling shutdowns and monthly tests of active system components (pumps and velves) which can be performed during reactor operation. The test interval is based on 'the judgment 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.
The monthly part travel exercising of the VCT outlet to charging pump suction valves, in lieu of the full travel exercise, 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.
Other ECCS valves whose operation -is not required to assure core flooding or containment spray shall be tested during each refueling shutdown period in accordance with 2.b.
The three check valves in the ECCS line to each loop provide assurance that 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 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. '
In addition to the check valves the ECCS line'to each loop contains a Motor Operated Valve (MOV) which is closed except for periodic monthly testing.
The M0V and reactor side piping is designed for full system pressure and is also capable of preventing an overpressure condition of connecting piping.
The leakage criteria provide an acceptable balance between the need to maintain a degree of tightness as a criterion of integrity on one hand and ALARA and power dependability considerations on the other giving due credit to the unique design feature of and protection provided by the four valves in series.
Verification that the spray piping and nozzles are open will be made initially by a suitably. sensitive method, and at least every five years thereafter. Since all piping material is all stainless steel, normally in a dry condition and with no plugging mechanism available, the retest every five years is considered to be more than adequate.
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 -
4.6-5 Amendment $2,S7,100
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are checked periodically. The component cooling and service water systems j operate when the reactor is in operation and are continuously monitored i for satisfactory performance.
The monthly testing interval of the steam generator motor driven emergency feed pumps' and the turbine driven auxiliary feedwater pump verifies their operability by recirculating water to the demineralized water tank.
The bistable actuation setpoints for the motor driven emergency feed pumps are checked monthly in accordance with #21 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 is perfonned on the Auxiliary Feedwater System to functionally verify the system alignment from the demineralized water storage tank to the steam generators.
Monthly inspections are performed to verify that all manual valves in the Auxiliary Feedwater System from the primary water source to the steam 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 will be made both from instrumenta-tion 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 will be verified periodically.
I 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 heater 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 l 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 non- +
condensible gasses from the reactor coolant system under severe post :
accident conditions in order to help prevent or mitigate core damage. *
- 4. 6- 6 Amendment No. 62,76,$7,100
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