ML19207A874
| ML19207A874 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19207A875 | List: |
| References | |
| NUDOCS 7908220540 | |
| Download: ML19207A874 (7) | |
Text
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UNITED STATES e ' j.c (
NUCLEAR REGULATORY COMMISSION n
WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.13 TO FACILITY OPERATING LICENSE NO. NPF-2 ALABAMA POWER COMPANY JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-348 I ntroduction By letter to Alabama Power Company (APC) aated December 29,1976, the NRC requested an evaluation of the Farley Nuclear Plant (FNP) system designs to determine susceptibility to overpressurization events. We also requested an analysis of the possible events, licensee proposed interim and permanent systems, and procedure modificatiol.s to reduce the likelihood and consequences of such events.
By letter dated March 17, 1977 APC proposed development and installation of an Overpressure Mitigating System (OPMS) using the existing pressurizer power operated relief valves as a long term solution to the problem.
However, by letter of September 6,1978 APC modified its earlier proposal.
APC's new proposal would use the existing pressure relieving capacity of the residual heat removal (RHR) system relief valves.
Subsequently, by 1etters dated November 3, 9 and 17,1978 and January 4, March 21, and April 17, 1979, APC provided infonnation on the OPMS design and transient analyses, proposed Technical Specifications for OPMS operability and surveillance, and responded to our specific concerns.
The OPMS, as now designed, is based upon operation of passiv-RHR relief valves without additional instrunentation (except alarms) or relays, solenoids and valve operators.
The system is designed to prevent a reactor coolant system (RCS) transient from exceeding the pressure and temperature limits of the Technical Specifications for Farley Nuclear Plant (FNP), as required by Appendix G to 10 CFR 50.
Background Discussion Over the last few years, incidents identified as pressure transients hdve occurred in pressurized water reactors (PWRs).
As used in this report " pressure transients" refers to events during which the temperaturo pressure limits of the reactor vessel, as shown in the Technical Specifica-tions, are exceeded. All of these incidents occurred at relatively low temperatures (less than 200 F) where the reactor vessel material toughness (resistance to brittle failure) is reduced from that which exists at normal operating temperatures.
7811Sii N#
7908220 The " Technical Report on Reactor Vessel Pressure Transients" in fiUREG-0138 summarizes the technical considerations relevant to this matter, discusses the safety conce/ns and existing safety margins of operating reactors, and describes the regulatory actions taken to resolve this issue by reducing the likelihood of future pressure transient events at operating reactors.
A brief discussion is presented here.
Reactor vessels are constructed in accordance with the ASME Boiler and Pressure Vessel Code. Steels used are particularly tough at normal reactor operating pressure and temperature conditions. However, these steels are less tough if subject to relatively high pressures at relatively low temperatures.
Thus, restrictions are placed on the system pressure during startup and shutdown operations.
At operating (hot) temperatures, the pressures allowed by Appendix G limits are above the setpoint of the installed pressurizer code safety relief valves. However, most operating PWRs did not have automatic pressure relief device setpoints low enough to mitigate pressure transients during cold. conditions (startup and shutdown) when the RCS is water-solid and non-vented.
By letter dated December 29, 1976 prior to issuing an operating license for FriP, we requested the applicant to begin efforts to design and install plant systems to mitigate the consequences of pressure transients at low temperatures. We also requested that operating procedures be examined and adninistrative -hanges be made to prevent initiat g overpressure events.
We also concluded that interim administrative controls should be imposed to assure safe operation until permanent overpressure mitigating hardware could be installed.
By 1etters dated January 24, March 17, and May 24,1977 (in response to our May 3,1977 letter) APC provided preliminary information describing interin measures to prevent these transients.
The applicant proposed several modifications to administrative procedures, design, and operator training as discussed below:
1.
Operator Training:
During cold license training, operators would be briefed on the types of events that could cause over-pressurization based on changes made in the procedures to minimize the probability of such events.
2.
RHR Relief Valve:
The setpoints of the two RHR relief valves would be set at 450 psig which is significantly below the isolation pressure of the RHR system.
One valve has a 900 gpm relief capacity which is above the flow capacity of a single charging pump.
781166 3.
Steam Bubble:
A steam bubble would be formed in the pressurizer at 160 F when the plant was being heated up.
The bubble would be collapsed at 160 F when the plant is cooled down. This procedure would minimize the amount of time in a water-solid condition.
4.
Charging Pump: Only one charging pump would be operable at RCS temperatures below 200 F.
This would limit the potential volumetric insurge.
With an open path to the RHR relief valve, no overpressuri-zation would occur if the RCS letdown line were inadvertently closed.
The otner charging pumps would have power removed.
5.
Letdown Line:
The RCS letdown heat exchanger control valve would be placed in manual control prior to starting or stopping an RHR pump when the RCS is in a water-solid condition. This would preclude a spurious isolation of the RHR system.
6.
Reactor Coolant Pumps:
The procedures would include a precaution to verify that RHR suction valves were open prior to starting an RCS punp during water-solid operation. Also, included would be procedure limitations while in water-solid conditions on starting an RCS pump after the loss of RCS flow when seal injection temperature was less than RCS temperature. This would require either (a) the start of an RCS punp within 5 minutes or (b) the establishment of a steam bubble in the pressurizer or_(c) a reduction of RCS pressure to 50-100 psig and the securing of seal flow for at least two hours prior to re-establishing scal flow and starting an RCS pump.
7.
Accumul ators:
The accumulator isolation valves would be closed and the power locked-out whenever the plant is on the RHR cooling mode.
8.
Al arm: An alarm utilizing existing RCS wide-range pressure instrumentatic would be installed.
This alarm would annunciate on the Main Control Board (audio and visual) when the RCS pressure and temperature approach Appendix G limits.
We found the proposed administrative and design changes acceptable as interim measures to minimize the likelihood of a water-solid overpressurization pending confirmation by the Inspection and Enforcement Office of NRC of their impl ementa tion.
Because of the minimal neutron damage to be suffered by the pressure vessel during its first operating cycle, we concluded that no credible event would cause vessel rupture due to overpressurization during this period.
Because of APC's proposed administrative procedures and the pressure vessel fracture toughness, we concluded that the reactor could operate for its first cycle with reasonable assurance that the health and safety of the public are protected.
?81187
. APC became a member of the utility group in developing long-tem solutions to mitigate the consequences of RCS pressure transients during water-solid operation. The long-tem design modifications considered by the group would use either the power-operated relief valves or RHR system re'ief valves to preclude violating Appendix G limits.
We required in lictnse condition 2.C.(3)(b) that an effective overpressure protection system be installed prior to the initiation of the second operating cycle.
This has been accomplished.
The purpose of this safety evaluation is to document the basis for our approval of APC's OPMS design, their analyses, and the issuance by fiRC of related Technical Specifications.
E val uation The proposed overall approach to eliminating overpressure events incorporates administrative, procedural and hardware controls with reliance upon the plant operator for the principal line of defense.
Preventive administrative and procedural measures include:
(1) procedural precautions, (2) de-energization of components during cold shutdown, (3) avoidance of water solid reactor coolant system whenever possible, and (4) addition of an overpressure protection system which incorporates lov: pressure relief using the existing RHR system relief valves.
Water-solid overpressure mitigation systems are designed to mitigate the consequences of an overpressure event in the RCS when at low temperatures.
The system at each plant must be designed to prevent violation of 10 CFR 50 Appendix G limits.
In particular, it must:
1.
Perfom its function assuming any single active component failure, 2.
For the worst mass and heat input events postulated, not violate Appendix G limits, ai demonstrated through appropriate calculational techniques, 3.
Meet IEE-279 requirements and provide overpressure alarm, 4.
Provide the ability to be tested to assure its operability, 5.
Function during and after an Operating Basis Earthquake (OBE), and 6.
tiot require offsite power for proper operation.
781168 The OPMS proposed by APC in letter uoted September 6,1978, as supplemented, is an integral part of the RHR system relying on the RHR pump suction line relief valves to provide pressure relief.
The RHR relief valves (RHRRV) are spring loaded, bellows type valves which have a setpoint of 450 psig.
The valves will be fully open at 495 psig (i.e., thew have 10% accumulation).
There are two isolation valves between each of the RHRRVs and the RCS.
The RHR system suction lines are automatically isolated when the RCS pressure exceeds 700 psig. The RHR system suction valves inside containment will be opened whenever the RCS temperature is 310 F or lower, thereby aligning the RHR relief valves for RCS pressure protection.
APC provided a single failure analysis which demonstrates that no single active failure can disable the mitigation system such that Appendix G limits are violated. Single failures considered included electrical faults, operator action, suction valve failure, relief valve failure, and loss of offsite power.
APC provided Appendix G curves for 100 F/hr cooldown and heatup rates (applicable for 8 effective full power years) for comparison against potential overpressure eve ' maximum pressures.
APC used the modified Westinghouse generic water-solid overpressure model with the specific RCS volume for the FNP and with a conserva-tive steam generator heat transfer area.
The worst mass input event was assumed to be the inadvertent operation of three high head safety injectica pumps with a maximum total flow rate of 1000 gallons per minute at zero psig backpressure.
The worst heat input event was assumed to be the starting of a single reactor coolant pump with a temperature differential of 50 F existing between the RCS and the steam generator. The maximum calculated RCS pressures for these postulated worst mass and heat input events remained below the pressures allowed by the Appendix G curves for transients initiated below 310 F and below 450 psig.
For transients above 310 F, the pressurizer safety valves would relieve pressure to prevent violation of Appendix G limits.
However, there is a hypothetical condition where it appears that the Appendix G curve might be exceeded by a pressure transient. The postulated case is one at a cooldown rate Fetween 60 F/hr and 100 F/hr.
For cooldown rates in this range while at lea RCS temperatures (below 126 F), starting one charging pump could result in violating the 60 F/hr or 100 /hr Appendix G curves. This condition is highly improbable as cooldown rates of 60 F/hr to 100 r/hr at such low temperatures are very unlikely if not impossible to achieve.
For these reasons the staff finds the maximum calculated water-solid overpressure event pressure analyses and consequences to be accept-able.
The RHR relief valves, which are installed on the two RHR pump suction lines, discharge to the pressurizer relief tank.
The FNP analysis conservatively
'/ 0 1 1 @
, takes no credit for RHR relief valve flow until 10 percent overpressure (495 psig) has been achieved at which point the valve was assumed to be instantaneously and fully open.
Valve backpressure was reviewed to assure that flow degradation due to flashing was included in the analysis.
The RHR valves are spring-loaded and have no electrical components.
The auto-closure interlock and the open permissive control circuits of the motor-operated RHR isolation valves meet the requirements of IEEE 279-1971. Power supplies for the RHR isolation valves, the pressurizer pressure sensors, and the RCS temperature sensors are designed so that no single failure in the electrical system or the loss of offsite power would isolate both
'f the RHR relief valves. Thus, we find the RHR relief valves, isolation valve circuitry and power supplies acceptable.
APC has conmitted to test the two RHR relief valves on an accelerated basis from that required by the ASME code.
Bench tests will be done at 18 month intervals n a rotating basis for at least one of the RHR relief valves to check the setpoint. The Technical Specifications surveillance has been nodified accordingly.
The RHR relief valves are certified by the manufacturer to be capable of withstanding an OBE with no degradation of perfornance.
The RHR system piping from the RCS hot leg to the RHR relief valves is quality group A per 10 CFR 50.55(a).
We find the valve testing and certified ability of the relief valves to operate during and after an OBE acceptable.
Changes to FNP Technical Specifications (Limiting Conditivas for Operation) submitted by APC (with ninor changes by our staff) wculd ensure proper operation of the system when required.
With the modifications we find these changes acceptable.
These modified changes are listed below:
1.
Specification 3.4.9.3:
The RHR suction valves must be open and the RHR relief valves must be operable or a RCS vent must be open v.aenever the RCS tenperature is less than or equal to 310 F.
2.
Specification 3.4.1 (Below P-7):
A reactor coolant pump cannot be started when the RCS temperature is below 310 F unless the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures, or the pressurizer water volume is less than 24 percent level.
APC has installed a permanent alarm in the control roon for Unit No.1 (and will install one prior to initial fue l loading for Unit No. 2) to alert the perator if the RHR isolation valves are not fully open when the RCS temperature is less than or equal to 300 F.
This s a seismic Category I alarm designed
- o the requirenents of IEEE-279 up to.he annunciator light.
The alarm has n unshared annunciator window.
APC hcs installed another alarn which would alert the operator to the existence of an overpressure event if the RCS pressure exceeds 450 psig.
781190
. Wc find the proposed overpressure mitigation system for the Joseph M.
Farley Nuclear Plant, Unit No. I to be acceptable because it satisfies our requirements noted above.
Although this evaluation has been prepared in connection with a pending licensing action on Farley Unit No.1, it applies to the identified OPMS being installed in Farley Unit No.
2.
This evaluation may be cited by the NRC as the basis for approval of the Unit No. 2 OPMS at a future date.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power 'evel and will not result in any significant environmental impact.
t.'ving made this determination, we have further concluded that the amenacent involves an action which is insignificant fran the standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does' not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the caamon defense and security or to the health and safety of the public.
Date:
July 31,1979 781131